On March 18, 2007, while operating at 28% power, a chemistry excursion occurred shortly after Condensate Filter Demineralizer 1T-13D was placed in service. The magnitude of the chemistry excursion required operators to shutdown the reactor in accordance with abnormal operating procedures and plant chemistry procedures. As a result of inserting the manual scram, Primary Containment Isolation System ( PCIS) groups 2, 3, and 4 isolations occurred when reactor water level dropped below 170 inches. All isolations went to completion. The reactor water level decrease is normal following a scram from 28% power due to void collapse in the reactor vessel. Reactor water level was subsequently restored to normal and the PCIS group isolations were reset.
Troubleshooting subsequently determined that the chemistry excursion was the result of resin intrusion from Condensate Filter Demineralizers into the Condensate System.
There were no actual safety consequences and no effect on public health and safety as a result of this event.
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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 Duane Arnold Energy Center 05000331 NUMBER� NUMBER
I. Description of Event:
On March 18, 2007, while operating at 28% power, a chemistry excursion occurred shortly after Condensate Filter Demineralizer 1T-13D was placed in service.
A chemistry transient occurred on 3/18/07 as a result of an earlier transfer of resin from the influent to effluent piping of the Condensate Filter Demineralizer (F/D) System. Resin transferred from Condensate Filter Demineralizers that were left in HOLD after the Condensate System was secured during the first week of the plant refueling outage. With a F/D vessel pressurized and the Condensate System secured, resin was transferred from the F/D vessel into the Condensate System as follows:
- With the F/D pressurized and the Condensate System secured, influent header pressure equalizing valves SV-1715A (B, C, D, E) will lift off their seat allowing flow from the F/D vessel to the influent header and into the Condensate System.
- Leakage through Condensate F/D Influent Header Isolation Valves CV-1718A (B, C, D, E) allows flow from the F/D vessel to the influent header and into the Condensate System.
- During the backwash cycle, the F/D vessel is pressurized by service air and for a short period of time the hold pump is off. Leakage past the influent and pressure equalizing valve allowed resin with high concentrations of contaminants to pass through to the influent header.
- Leakage past the influent valves and influent pressure equalizing valves caused an undetected loss of suction on several operating Hold Pumps due to low level in the F/D vessel. This condition resulted in the release of resin from filter elements, accumulating in the bottom of the F/D vessel from which it leaked into the Condensate system via the flow paths noted above.
When the plant restored the Condensate System in preparation for plant startup, the bypass valve around the Condensate F/D system was open. After starting the first condensate pump on 3/13/07, flow moved resin that had previously leaked into the influent header to the effluent header through the bypass line. Resin subsequently accumulated in a dead leg section of piping on the effluent side of Condensate F/D Vessels D and E.
Condensate F/D's A, B, and C were placed in operation at various times after 3/13/07 to support long path cleanup and Condensate System startup. Condensate F/D A was removed from service on 3/14/07 due to high conductivity and placed in HOLD. On 3/18/07 with the plant operating at —30% power, conductivity was observed to be increasing and chemistry requested that operations place an additional F/D in service.
The Control Room Supervisor directed that F/D D be placed in operation. Immediately after placing F/D D in operation a significant increase in effluent conductivity occurred.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Duane Arnold Energy Center 05000331 YEAR NUMBER� NUMBER When F/D D was placed in service the resin that had accumulated in the dead leg of effluent piping was pumped forward into the reactor. A resin intrusion occurred, resulting in an increase in reactor water conductivity to 33 pS/cm and sulfate concentration to 4000 ppb (normal conductivity is -0.055 pS/cm and normal sulfate concentration is exceeded Action Level 3 of Plant Chemistry Procedure (PCP) 1.9, a reactor shutdown and cooldown of the RPV was performed in accordance with Abnormal Operating Procedure (AOP) 639 and PCP 1.9.
The magnitude of the chemistry excursion required operators to shutdown the reactor in accordance with abnormal operating procedures and plant chemistry procedures. As a result of the manual scram, PCIS groups 2, 3, and 4 isolations occurred when reactor water level dropped below 170 inches. All isolations went to completion. The reactor water level decrease is normal following a scram from 28% power due to void collapse in the reactor vessel. Reactor water level was subsequently restored to normal and the PCIS group isolations were reset.
Notifications were made under 10 CFR 50.72(b)(3)(iv)(A) and 50.72(b)(2)(iv)(B) on March 18, 2007 and are listed as event number 43247.
II. Assessment of Safety Consequences:
The function of the Condensate Demineralizer System is to remove soluble and particulate material from the condensate water in order to maintain required reactor water quality, including during minor condenser tube leakage. The system must also maintain condensate supply to the Feedwater System at required flow and pressure. The system does not have a safety-related function.
However, the system is important to power production.
The resin intrusion of 3/18/07 caused a significant degradation of reactor water chemistry parameters, exceeding Action Level 3 values. In accordance with industry guidelines and plant procedures, the plant was promptly shut down and reactor water temperature was reduced to less than 200°F to minimize the impact on fuel, RPV internals, and plant components. Plant demineralizer systems were subsequently operated to restore reactor water chemistry.
A review of the event was completed prior to plant restart. This review determined that, based on internal and external Operating Experience, no detrimental effects on reactor materials or fuel are expected.
Therefore, the plant shutdown did not result in any radiological or nuclear concern which would impact the health and safety of the public.
This event did not result in a Safety System Functional Failure.
Ill. Cause of Event:
An investigation into this event was completed under Root Cause Evaluation (RCE) 1064.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Overall RCE Conclusions It was concluded that keeping several FID beds in HOLD after the Condensate System was secured during the first week of the refueling outage created opportunities for resin to enter the influent headers of the condensate filter demineralizers due to existing equipment deficiencies. Resin that had accumulated in the influent header was subsequently transferred into the effluent header as the Condensate System was placed in operation during plant startup. Resin that had accumulated in a dead leg section of effluent piping was subsequently passed forward into the reactor after Condensate F/D D was placed in service causing a chemistry excursion.
Root Causes The RCE identified the following Root Causes (RC):
RC 1: The design of pressure equalization valves SV1715A (B, C, D, E) and leakage through Condensate Demineralizer influent valves CV1718A (B, C, D, E) allow resin to leak out of the Condensate F/Ds and enter the condensate system.
RC 2: Station procedures for operation of the Condensate and Condensate Demineralizer systems do not prevent consequences which are adverse to power production. Specifically, they do not provide clear direction to either secure and backwash F/Ds or operate them in FLOAT when the Condensate System is secured and do not ensure that all system piping is flushed prior to system restoration.
IV. Corrective Actions:
Immediate Actions to address the Condition In response to the resin intrusion of 3/18/07, the plant was promptly shutdown and reactor water temperature was reduced to less than 200°F, in accordance with industry guidelines and plant procedures. Plant demineratizer systems were operated to restore reactor water chemistry.
A formal troubleshooting plan was implemented to investigate the cause of the resin intrusion. A review of the event was also completed prior to plant restart. This review determined that, based on internal and external Operating Experience, no detrimental effects on reactor materials or fuel are expected. Affected piping systems were flushed prior to power operation.
Corrective Actions to Prevent Recurrence (CATPRs) CATPR 1-1 Establish a positive sealing method in both directions on the Pressure Equalizing Line for each Condensate F/D. This will require replacement with a different design, installation of check valves, or other action which accomplishes positive sealing.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Duane Arnold Energy Center 05000331 YEAR CATPR 1-2 Rebuild operator and valve internals on each Condensate Demineralizer Influent and Effluent Control Valves (CV1718A (B, C, D, E) and CV1719 A (B, C, D, E)) to ensure that the required isolation function of these valves is maintained.
CATPR 2-1 Revise Operating Instruction (01) 639 and 01 644 to provide clear direction that if a Condensate Demineralizer will not immediately be backwashed that it must be manually isolated and placed in FLOAT when the Condensate System is secured.
CATPR 2-2 Establish flushing criteria and methods for the Condensate, Feedwater, and Condensate Demineralizer Systems.
CATPR 2-3 Revise 01 639 to incorporate Condensate Demineralizer Effluent flushing criteria established in CATPR 2-2.
V. Additional Information:
Previous Similar Occurrences:
From LER review over the previous 10 years, the following two similar occurrences were identified in:
2003-001 - Punctured Main Condenser Tube Resulting in Rx Water Chemistry Excursion and Manual Rx Scram.
2003-005 - Unplanned Manual Reactor Scram due to High Reactor Coolant Conductivity.
EIIS System and Component Codes:
SF - Condensate Demineralizer System Reporting Requirements:
This report is being submitted pursuant to 10CFR50.73(a)(2)(iv)(A).
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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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