04-02-2008 | On June 6, 2007, Turkey Point Unit 3 was operating at 100% power. At 0650, Operators observed that Rod Position Indication ( RPI) [IG] for control rod F-4 in Control Bank C [AA] began to oscillate above and below 218 steps. The RPI for rod M-6 in the same Control Bank C had been inoperable since September 1, 2006. At 0745, unable to comply with Technical Specification (TS) 3.1.3.2 Action a., i.e., maximum of one analog rod position indication per bank inoperable, TS 3.0.3 was entered to place the Unit 3 in Hot Standby.
At 0746, a plant shutdown was initiated. The cause of the failed RPIs is the incorrect application of the coil stack connector [IG:CON] insert material (neoprene) for the required environment. The root cause for this event is that the Integrated Head Assembly Vendor Quality Assurance Program implementation failed to ensure the proper connectors were used in the fabrication of the RPI pigtail assemblies supplied to FPL during the reactor vessel head replacement in 2004. Additionally, FPL's nuclear material management technical reviewer failed to review design basis docutnentation and
- to ensure proper engineering documentation was prepared for the new stock code part. The Unit 3 RPI coil connectors were removed, the RPI pigtail assembly wires were spliced to the RPI intermediate vessel head cables and upon completion of repairs Unit 3 was returned to service.
0 Unit 4 was proactively shutdown to replace the same type of connectors with qualified cable splices, similar to the Unit 3 repairs.
NRC FORM 366 (6-2004) PRINTED ON RECYCLED PAPER |
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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 3 05000250 NUMBERNUMBER
DESCRIPTION OF THE EVENT
On June 6, 2007, Turkey Point Unit 3 was operating at 100% power with no safety systems out of service.
At 0650, Operators observed that Rod Position Indication (RPI) for control rod F-4 in Control Bank C began to oscillate above and below 218 steps. At that time, Operations entered the off normal operating procedure for rod misalignment, declared control rod F-4 inoperable, and entered Technical Specification (TS) 3.1.3.1 Action d.1 to restore the inoperable rod to operable status. Operations confirmed that no control rod misalignment existed, and exited the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS action.
Subsequently, at 0745 Operations entered the off normal operating procedure for rod position indication malfunction, declared the RPI for rod F-4 in Control Bank C inoperable and reviewed TS 3.1.3.2 for compliance. TS 3.1.3.2 allows continued operation with a maximum of one inoperable rod position indication per Control Bank provided the rod position is verified indirectly by the moveable incore detectors (i.e., flux map) every eight hours or power must be reduced to 75%. If more than one RPI is inoperable in the same Control Bank, entry into TS 3.0.3 is required to initiate a unit shutdown.
Prior to this event, there had been three RPI failures since September of 2006. Flux mapping was being performed for two control rods G-5 in Control Bank A and E-5 in Shutdown Bank B due to these two RPI failures that had recently occurred on May 1, 2007 for rod G-5 and on June 2, 2007 for rod E-5. The third RPI for rod M-6 in control Bank C had occurred September 1, 2006. Flux mapping was not being performed for rod M-6, since a Technical Specification change was approved by the Nuclear Regulatory Commission (NRC), to allow an alternate method for monitoring the rod's position, i.e., by verifying gripper coil parameters of the control rod drive mechanism to determine it has not changed state.
At approximately 0745, with inoperable RPIs for rods F-4 and M-6 in the same control bank C, unable to comply with TS 3.1.3.2 Action a., i.e., maximum of one analog rod position indication per bank inoperable, Operations entered 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action in accordance with TS 3.0.3 to place Unit 3 in Hot Standby.
At approximately 0746, Operations initiated the Unit 3 shutdown. At 1152, reactor power was reduced below 25% and a manual reactor trip was performed in accordance with operating procedures. As such, at 1152, Operations exited TS 3.0.3. There were no abnormal indications observed during the duration of the Unit 3 shutdown. The RPI failures had no effect on the operation of any plant safety systems. There were no adverse effects on nuclear safety nor was the health and safety of the public compromised during this event.
This event was reported to the NRC on June 6, 2007 at 0932 pursuant to 10CFR50.72(b)(2)(i) due to initiation of a plant shutdown required by the plant Technical Specifications; and submitted event notification 43408. FPL condition report 2007-17324 was originated. This event is reportable pursuant to the requirements of 10CFR50.73(a)(2) (i)(A) due to a completion of a plant shutdown required by Technical Specifications. This LER is a supplement and supersedes the LER previously submitted to NRC by FPL letter L-2007-123 on August 6, 2007.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 3 05000250 NUMBER
BACKGROUND INFORMATION
The Turkey Point Unit 3 Reactor Vessel Closure Head (RVCH) was replaced during the Fall 2004 refueling outage. As part of the replacement effort, additional improvements were made to increase overall reactor vessel related system reliability, and enhance refueling/defueling operations. The RVCH was replaced with a new Integrated Head Assembly (IHA). This included all new IHA cable and connector assemblies for the Rod Position Indication, Control Rod Drive Mechanism, Core Exit Thermocouples, and Reactor Vessel Level Instrumentation System.
The RPI cable replacement included the following (Refer to Figure 1): a) replacement of cables surrounding the reactor cavity with new cable spliced into the existing cable, b) new bulkhead connector and panel on the refueling floor on the west end of the reactor cavity, c) new intermediate cables with connectors from the bulkhead connector panel to the RPI Seismic Plate Coil stack connectors. The existing Rod Position Indicator (RPI) coils were reused. The RPI coil stack seismic plate connectors were also replaced.
Figure 1: RPI Coil Stack Cable Connectors New Intermediate cable with Connectors Bulkhead (Angled) CRDMConnector Extension Tube Bulkhead Panel RPI Seismic Plate (Existing) RPI Coil Stack New Bulkhead � (Existing) Connector Pig-Tail (spliced to existing) New RPI Connector New RPI Pig-tails (spliced to coil leads) The rod position detector is a linear variable transformer consisting of primary and secondary coils alternately stacked on stainless steel support tube. The Rod Control Cluster Assembly (RCCA) drive rod serves as the "core" of the transformer. The vertical position of the drive rod changes the primary to secondary coupling and produces a unique A.C. analog secondary voltage. The output voltage is an analog signal directly proportional to the position of the control rod.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
EVENT ANALYSIS
There have been four Turkey Point Unit 3 RPI failures since September 1, 2006 resulting in inoperable RPIs for rods M-6 in control Bank C, G-5 in Control Bank A, E-5 in Shutdown Bank B, and F-4 in Control Bank C. The number and chronology of failures was considered highly abnormal and potentially indicative of common cause failure. The symptoms of failure were similar amongst the four RPI detectors consisting of sudden erratic rod position indication without intentional rod movement.
After testing the RPI cables and connectors it was determined that the cause of the erratic RPI indication was the insulation breakdown of the connector insert on the reactor head. Based on part numbers, vendor documentation, insert material color, and laboratory testing of the failed dielectric from the M-6 RPI coil stack connector, it was determined that the RPI coil stack seismic plate connectors utilized neoprene rubber inserts, instead of silicone rubber. The connector insert, which should provide insulation between connector pins and connector body, was found to be conductive across its exposed face. All the failures identified were at the RPI coil stack seismic plate connection with the seismic plate-mounted male connectors being the failure initiator/propagator. For these failed connectors, all tests showed evidence of a migration of neoprene material to the mating cable connectors with silicone rubber inserts.
The RPI coil stack seismic plate connectors with neoprene inserts are a subcomponent of the RPI coil stack pigtail assembly. These assemblies were fabricated and installed during the Fall 2004 outage for the replacement of the reactor vessel head. The RPI cables and connectors installed in the vicinity of the reactor vessel must be able to maintain their physical and electrical insulation properties over many years under high temperature environmental conditions. The degradation of the neoprene rubber insert from the pigtail assembly connector contaminated and permeated the silicon rubber insert of the intermediate cable mating connector. The contamination of the intermediate cable connector resulted in a breakdown of its silicon rubber insert and caused it to become conductive. This conductivity resulted in shorting of conductor dielectric thus causing the erratic RPI detector oscillations.
CAUSE OF THE EVENT
There are two root causes for this event. One root cause is that the IHA Vendor Quality Assurance Program implementation failed to ensure the proper connectors specified by the IHA vendor and FPL purchase order were used in the fabrication of the RPI pigtail assemblies supplied to FPL. The other is that the FPL nuclear material management technical reviewer failed to review design basis documentation and take action to ensure proper engineering documentation was prepared for the new stock code part per FPL quality instructions.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Turkey Point Unit 3 05000250
ANALYSIS OF SAFETY SIGNIFICANCE
Unit 3 was shutdown in accordance to plant procedures due to multiple RPI failures. The multiple RPI failures had no adverse impact on the ability of the operators to shutdown the reactor. There were no instances of rod misalignment.
ADDITIONAL INFORMATION: EXTENT OF CONDITION Based on the insulation resistance checks of all 45 RPI circuits, only the four RPIs identified previously were found with unacceptable resistance readings. The remaining 41 RPIs were capable of performing their function at the time of the Unit 3 shutdown. Only the RPI seismic plate stack coil pigtail assembly connectors had the neoprene inserts. The Unit 4 installed RPI seismic plate stack coil pigtail assembly connectors had neoprene inserts. Other head cable connector inserts were silicone rubber, which is the proper material for the application. The condition report for this event is 2007-17324.
CORRECTIVE ACTIONS
Turkey Point Unit 3 was shutdown and the RPI seismic plate coil stack connectors were removed.
Because of parts unavailability and the long lead time to procure new connectors, the RPI coil stack pigtail assembly wires have been spliced to the RPI intermediate head cables as a corrective action which eliminates the failure mode by eliminating the unqualified portion of the RPI system.
Turkey Point Unit 4 was proactively shutdown on July 22, 2007 to remove these connectors and splice the RPI coil stack pigtail assembly wires to the RPI intermediate head cables.
FPL personnel have investigated the IHA vendor's quality program. FPL Quality Assurance findings were incorporated into the IHA vendor's corrective action program for process resolution.
FPL nuclear material management technical reviewers were trained and qualified to FPL's Quality instructions.
SIMILAR EVENTS: There is no record of past occurrences of this type of event at Turkey Point.
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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 50.72(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 |
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