11-05-2012 | On September 6, 2012, at 1314 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.99977e-4 months <br />, the reactor building ventilation system was being restored to service following planned maintenance and surveillance activities. During the reactor building pressure transition when restoring reactor building ventilation (from the Standby Gas Treatment System), a worker entered the Unit 2 Reactor Building ( RB) 595 Interlock door (a secondary containment interlock) and found the Unit 2 RB Interlock (RBI) HRSS-Side door (which leads to the environment) had opened at the same time. The worker immediately secured the door and notified Operations personnel. A review of the door alarm history determined the door had been opened for approximately eight seconds, therefore secondary containment had been breeched for a short duration.
Since both Units 1 and 2 share a common reactor building, a breech occurred for eight seconds that impacted both Units 1 and 2 secondary containments when the Unit 2 RB 595 Interlock door and the Unit 2 RBI HRSS-Side door were opened at the same time.
The cause of the secondary containment breech event was that the Unit 2 RBI HRSS-side door latching mechanism was not fully engaged at the time that the Unit 2 RB 595 Interlock door opened.
Corrective actions included immediate closure of the Unit 2 RBI HRSS-side door (2-0020-153), confirmation of being fully latched, and removal of the Unit 1 and 2 RBI HRSS-side doors from service. Door latch repairs will be made, and the Unit 1 and 2 RBI HRSS-side doors will be removed by permanently enclosing the door openings.
The safety significance of this event was minimal. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), and (a)(2)(v)(D), which requires the reporting of any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:
control the release of radioactive material, and mitigate the consequences of an accident. |
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LER-2012-004, Breech in Secondary ContainmentDocket Number |
Event date: |
09-06-2012 |
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Report date: |
11-05-2012 |
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2542012004R00 - NRC Website |
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PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EllS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
A Secondary Containment Breech Occurred When Unit 2 Reactor Building 595 Interlock Door and the Unit 2 Reactor Building Interlock HRSS-Side Door Opened at the Same Time
A. CONDITION PRIOR TO EVENT
Unit: 2 Event Date: September 6, 2012EEvent Time: 1314 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.99977e-4 months <br /> Reactor Mode: 1 Mode Name: Power Operation Power Level: 100%
B. DESCRIPTION OF EVENT
On September 6, 2012, at approximately 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, the Reactor Building [NH] Ventilation system was secured for planned maintenance and surveillance activities. At approximately 1314 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.99977e-4 months <br />, the Reactor Building Ventilation system was being restored to service (from the Standby Gas Treatment System [BH]), following planned maintenance and surveillance activities. Operations restarted the Unit 1 Reactor Building Vent Fans [FAN] and Unit 2 Reactor Building Exhaust Fans in accordance with procedures. The Reactor Building to atmosphere differential pressure (d/p) rose above zero due to the slow response of the Reactor Building Ventilation (RBV) discharge modulating dampers [CDMP].
The Control Room [NA] panel alarm [PA] 912-5 C-1, Rx Bldg 1 Low DP, initiated with the associated Reactor Building to atmosphere d/p indicator, rising above zero (went positive). Technical Specification (TS) 3.6.4.1, Action A.1, was entered for the Reactor Building to atmosphere d/p rising above -0.1 inch water. The TS was exited when it was determined that the indication was anticipated due to the slow response of the RBV discharge modulating dampers, and that no emergency existed.
During the time that Operations was restoring the RBV system, also at approximately 1314 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.99977e-4 months <br />, a worker simultaneously entered the Unit 2 Reactor Building (to Turbine Building [NM]) 595 elevation Interlock [IEL] door [DR] from the Reactor Building. The Unit 2 RB Interlock High Radiation Sample Station [IK] (HRSS)-side door opened when Reactor Building d/p went positive. Prior to entry into the Unit 2 RB 595 Interlock area, no other interlock door was indicated as open, since there was no light indication [IL] present at the door that would indicate another interlock door was open. As the worker entered the interlock, the Unit 2 RBI HRSS-side door (2-0020-153), a secondary containment interlock door, was found open to the outside environment, and air was flowing from the Unit 2 RB.
The Unit 2 RBI HRSS-side door was immediately secured by the worker by pulling on the crash bar until the door was shut without the door re-opening; and air flow had stopped. The worker verified the Unit 2 RB 595 level interlocks were functional by checking that all the interlock lights would light when required, and the other two interlock doors functioned properly. The worker notified Operations of the event. The Field Supervisor was dispatched to verify that the Unit 1 and Unit 2 RB to Turbine Building 595 interlocks were functioning properly. A review of the door alarm history determined the door had been opened for approximately eight (8) seconds.
E■1111
- 11■ Since both Units 1 and 2 share a common reactor building, a breech occurred for eight seconds that impacted both Units 1 and 2 secondary containments when the Unit 2 RB 595 Interlock door and the Unit 2 RBI HRSS-Side door were opened at the same time. Given the impact on the secondary containment, this report is submitted (for Units 1 and 2) in accordance with the requirements of 10 CFR 50.73 (a)(2)(v)(C), and (a)(2)(v)(D), which requires the reporting of any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: control the release of radioactive material, and mitigate the consequences of an accident.
C. CAUSE OF EVENT
The apparent cause of the secondary containment breech event was that the Unit 2 RBI HRSS-side door latching mechanism was not fully engaged at the time that the Unit 2 RB 595 Interlock door opened. It is probable that the crash bar was bumped some time after the Security door test and before occurrence of this event.
Troubleshooting performed on September 7, 2012 revealed that the fingers on the Unit 2 RBI HRSS-side door crash bar were not re-engaging consistently after the crash bar was released. It was determined that use of the crash bar on the door will override the secondary containment interlock logic. The Unit 2 RBI HRSS-side door was inspected for any equipment reliability issues with the latching mechanism on the door. No damage to the door or the latching mechanism was found.
Due to the relative position of the strike plate on the door jam to the latching mechanism on the door upon closure, the door could only be latched completely if the door was closed with the aid of the closure device from approximately full open, or if extra pulling force was applied to the crash bar to extend the latch completely outward.
A contributing cause to this event was that the RBV discharge modulating dampers have been "tuned" to slow their response to d/p changes caused by atmospheric conditions outside of the RB. The RB to atmosphere d/p rose above zero (went positive) due to the slow response of the RBV discharge modulating dampers. During normal operations, the pressure that the RBI HRSS-side door in the RB to Turbine Building 595 interlock is exposed to is negative (pulling inward). Regardless of the status of the door latching mechanism, the normal pressure direction and magnitude would be enough to keep the RBI HRSS-side door closed if not acted upon by another force physically pushing it open.
A prior Engineering Change had installed a new control system to reduce or eliminate re-occurring failures of the old d/p control system. During the setup of the new control system, the new controller was tuned to react slower to minimize excessive swinging when the external pressure changes. If the Standby Gas Treatment System is operating, the amount of negative building pressure seen by the controller will cause the RBV discharge modulating dampers to go closed. When the RBV is started, the RBV discharge modulating dampers consistently react too slowly to prevent the RB d/p from going positive.
D. SAFETY ANALYSIS
System Design The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). In conjunction with operation of the Standby Gas Treatment System (SBGTS) and closure of certain valves [V] whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment, and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be operable, or that take place outside primary containment.
The SBGTS is designed to maintain the reactor building (reactor building is common to both Units 1 and 2) at a negative pressure and to filter the exhaust of radioactive matter from reactor building spaces to the environment (by particulate filtration and halogen adsorption) in the unlikely event of a DBA, including the Loss of Coolant Accident (LOCA) and the refueling accident. It is also instrumental in maintaining the integrity of secondary containment during a primary to secondary containment instrument line break. Two parallel trains are provided, each of which is capable of producing greater than 0.25 inches water negative pressure required in the reactor building while processing 4000 cubic ft /min of exhaust air.
Safety Impact Since both Units 1 and 2 share a common reactor building, a breech occurred for eight seconds that impacted both Units 1 and 2 secondary containments when the Unit 2 RB 595 Interlock door and the Unit 2 RBI HRSS-side door were opened at the same time.
TS 3.6.4.1, Action A.1, requires restoration of secondary containment to operable status within four hours. This four hour Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during Modes 1, 2, and 3, since the probability of an accident occurring during this short period where secondary containment is inoperable is minimal.
Risk Insights The plant Probabilistic Risk Assessment (PRA) model gives no credit to the secondary containment and does not include it in the model, hence the as-found conditions did not contribute to an increase in risk.
Although a secondary containment loss of function occurred momentarily when the Unit 2 RBI HRSS-side door was found open concurrently with the opening of the Unit 2 RB 595 Interlock door, there was no DBA condition in progress, and secondary containment was restored within eight (8) seconds when the Unit 2 RBI HRSS-side door was immediately reclosed.
In conclusion, the overall safety significance and impact on risk of this event were minimal.
E. CORRECTIVE ACTIONS
Immediate:
1. An initial inspection was performed on the latching mechanism for the Unit 2 RBI HRSS-side door.
2. The Unit 2 RBI HRSS-side door (2-0020-153) was closed and confirmed fully latched, which eliminated the immediate secondary containment integrity concerns.
3. The RBV was recovered and established to maintain the Reactor Building at a negative d/p compared to the outside atmosphere.
4. A solid robust barrier (block) was placed between the crash bar on the Unit 2 RBI HRSS-side door and the door itself to prevent inadvertent depressing of the crash bar.
5. Operations barricaded and removed the Unit 1 and Unit 2 RBI HRSS-side doors (1-0020-147 and 2-0020 153) from service.
Follow-up:
1. Unit 1 and Unit 2 RBI HRSS-side doors will be inspected and repairs will be made to return the crash bar and door latch assemblies to approved vendor specifications.
2. A determination will be made of the feasibility of installing a secondary latching device on the inside of the Unit 1 and Unit 2 RBI HRSS-side doors, capable of preventing the doors from inadvertent opening in the event the primary latch fails.
3. The Unit 1 and Unit 2 RBI HRSS-side doors will be removed by permanently enclosing the door opening.
4. The issue of re-occurring low d/p conditions during Reactor Building ventilation changes will be resolved.
F. PREVIOUS OCCURRENCES
The station events database, LERs, EPIX, and NPRDS were reviewed for similar events at Quad Cities Nuclear Power Station. This event was a secondary containment loss of function when a latch assembly failed to hold a door closed during a loss of d/p condition in the Reactor Building. No similar events were found.
G. COMPONENT FAILURE DATA
Door catch - Component Manufacturer: Folger Adam Co., Component Model Number: 310 — 2 3/4.
This event has been reported to ICES as Failure Report No. 300532.
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05000413/LER-2012-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000327/LER-2012-001 | Unanalyzed Condition Affecting Essential Raw Cooling Water System due to External Flooding | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000388/LER-2012-001 | Two Control Room Floor Cooling Systems Inoperable | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000395/LER-2012-001 | Vice President, Nuclear Operations 803.345.4342 August 3, 2012SCE~r .so A SCANA COMPANY Document Control Desk U. S. Nuclear Regulatory Commission
Washington, DC 20555
Dear Sir / Madam: Subject:VVIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSEE EVENT REPORT (LER 2012-001-01) CORE EXIT THERMOCOUPLES & REACTOR WATER LEVEL INDICATION SYSTEM INOPERABLE DUE TO AN INADEQUATE MAINTENANCE PROCEDURE Attached is Licensee Event Report (LER) No. 2012-001-01 for the Virgil C. Summer Nuclear Station Unit 1. This revised report describes a condition where Core Exit Thermocouples and Reactor Water Level Indication System would not be operable for accident monitoring. This report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B). This letter and attached LER contain no new commitments and no revisions to existing
commitments.
Should you have any questions, please call Bruce Thompson at (803) 931-5042. Very truly yours, Dal Thomas D. Gatlin TS/TDG/jw
Attachment
c: K. B. Marsh P. Ledbetter S. A. Byrne J. C. Mellette J. B. Archie EPIX Coordinator N. S. Carps K. M. Sutton J. H. Hamilton INPO Records Center R. J. White Marsh USA, Inc. W. M. Cherry R. J. Schwartz V. M. McCree NSRC R. E. Martin RTS (CR-11-01807) NRC Resident Inspector FileV(818.07) M. N. Browne PRSF (RC-12-0116) V fALVirgil C. Summer Station • Post Office Box 88 .Jenkinsville, SC • 29065 • F (803) 345-5209 1 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010) Estimated burden per response to comply with this mandatory collection request: 80 hours.0Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the FOIA/Privacy Section (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet e-mail to infocollects.resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management andLICENSEE EVENT REPORT (LER) 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 Virgil C. Summer Nuclear Station Unit 1 05000 395 1 OF 5 4. TITLE Core Exit Thermocouples & Reactor Water Level Indication System Inoperable due to Inadequate Maintenance Procedure | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000397/LER-2012-001 | DG-3 Inoperable for Longer than Allowed by TS Due to Failed Governor 05000 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000410/LER-2012-001 | Forced Shutdown Due to an Increase in Drywell Leakage in Excess of Technical Specifications Limit | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000289/LER-2012-001 | Single Condition Making Independent Trains Inoperable | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | 05000286/LER-2012-001 | Common Cause Inoperability of Both Trains of Motor Driven Auxiliary Feedwater (AFW) Pumps Due to Inability to Control AFW Regulating Valves After Isolation of Nitrogen Backup | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | 05000265/LER-2012-001 | Main Steam Isolation Valve Local Leak Rate Test Exceeds Technical Specifications Limit | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000456/LER-2012-001 | Two Main Steam Safety Valves Failed Pre-outage Setpoint Testing Due to Abnormal Spring Geometry | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000261/LER-2012-001 | Technical Specification Required Plant Shutdown Due to Missed Surveillance and Operation Prohibited by Technical Specifications | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000483/LER-2012-001 | Modification Implementation Error Adversely Impacted the Containment Cooling System | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 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-2012-001 | Containment Concrete Thickness At Spalled Patch Does Not Meet Technical Specification Design Value | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000414/LER-2012-001 | Carolinas Duke Energy Carolinas, LLC 4800 Concord Rd. York, SC 29745 803-701-4251 December 20, 2012 U.S. Nuclear Regulatory Commission
Attention: Document Control Desk
Washington, D.C. 20555
Subject:�Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station, Units 1 and 2
Docket Nos. 50-413 and 50-414
Licensee Event Report 414/2012-001
Pursuant to 10 CFR 50.73(a)(1) and (d), attached is Licensee Event Report 414/2012-001,
Revision 0 entitled, "Diesel Generator (DG) 2B Was Unknowingly inoperable from 09/28/12 to
10/23/12 Due to Failed Tachometer Relay Power Supply".
This report is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(B), 10 CFR
50.73(a)(2)(ii)(B), and 10 CFR 50.73(a)(2)(v)(A)-(D).
There are no regulatory commitments contained in this letter or its attachment. This event is considered to be of no significance with respect to the health and safety of the
public.
If there are any questions on this report, please contact L.J. Rudy at (803) 701-3084. Kelvin Henderson LJR/s Attachment www. duke-energy. corn Document Control Desk Page 2 December 20, 2012 xc (with attachment): V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 J.H. Thompson (addressee only) NRC Project Manager U.S. Nuclear Regulatory Commission Mail Stop 8-G9A 11555 Rockville Pike Rockville, MD 20852-2738 G.A. Hutto, III NRC Senior Resident Inspector Catawba Nuclear Station INPO Records Center 700 Galleria Place Atlanta, GA 30339-5957 Document Control Desk Page 3 December 20, 2012 bxc (electronic copy)(with attachment): INPO L.E. Harmon C.S. Kamilaris R.D. Hart G.Y. Helton S.F. Hatley (ICES) M.K. Green R.T. Simril, Jr. B.C. Carroll M.C. Nolan W.J. Pritchett, Jr. T.L. Patterson K.R. Alter H.D. Brewer R.E. Abbott, Jr. B.J. Horsley S.L. Western bxc (hard copy)(with attachment): D.B. Alexander L.S. Nichols L.J. Rudy ELL Master File CN-801.01 LER File RGC Date File NCMPA-1 NCEMC PMPA ICES Lee.Harmon@NRC.gov NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013
(10-2010) Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported
lessons learned are incorporated into the licensing process and fed back to industry. Send commentsLICENSEE EVENT REPORT (LER) regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to infocollects.resource@nrclov, 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(See reverse for required number of to impose an information collection does not display a currently valid OMB control number, the NRCdigits/characters for each block) may not conduct or sponsor, and a person is not required to respond to, the information collection. r1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE Catawba Nuclear Station, Unit 2 05000414 10OF •4. TITLE Diesel Generator (DG) 2B Was Unknowingly Inoperable from 09/28/12 to 10/23/12 Due to Failed Tachometer
Relay Power Supply | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 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 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | 05000247/LER-2012-001 | Technical Specification (TS) Prohibited Condition Caused by an Inoperable 23 Emergency Diesel Generator Fuel Oil Storage Tank Due to Fuel Oil Below TS Limit | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000269/LER-2012-001 | Unanalyzed Conditions Exist for Standby Shutdown Facility Mitizated Events | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000445/LER-2012-002 | COMANCHE PEAK 05000445 10OF06 | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System | 05000413/LER-2012-002 | Discovery of Inadequacy in Surveillance Testing of Solid State Protection System | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000261/LER-2012-002 | Unplanned LCO 3.5.4 Entry Due to RWST alignment to Purification | 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 | 05000410/LER-2012-002 | Loss of Isolation Function on RHR Shutdown Cooling Suction Line due to Breaker Trip | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000482/LER-2012-002 | . One Train of Automatic Safety Infection Blocked During Entry Into Mode 3 Due To Procedural Weakness | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000361/LER-2012-002 | Emergency Diesel Generator Vibration Trip Not Bypassed For Non-Accident Condition | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000346/LER-2012-002 | Leak from Reactor Coolant Pump Seal Piping Socket Weld due to High Cycle Fatigue | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000333/LER-2012-002 | High Pressure Coolant Injection Pressure Control Valve Failure | 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 | 05000306/LER-2012-002 | Unit 2 Emergency Diesel Generators Inoperable Due To Missing Flood Control Barrier Seal | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System | 05000250/LER-2012-002 | Non-compliance with TS 3.4.9.3 due to Manual Isolation Valve Found in Incorrect TS Configuration | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) | 05000278/LER-2012-002 | Failure of Primary Containment Isolation Valve due to Foreign Material Results in Condition Prohibited by TS | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000316/LER-2012-002 | Unit 2 Reactor Trip from Generator Trip Due to Incorrect Relay Setting | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000282/LER-2012-002 | Unplanned Actuation of 121 Motor Driven Cooling Water Pump | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000298/LER-2012-002 | Improper Rotor Installation Causes Failure of Diesel Generator to Start | 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-2012-002 | Technical Specification (TS) Prohibited Condition Caused by New Fuel Assemblies Stored in a Configuration Prohibited by the TS | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000237/LER-2012-002 | Inlet Steam Drain Pot Drain Line Leaks Result in HPCI Inoperabilities | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000298/LER-2012-003 | Reactor Building Doors Opened Simultaneously Causes Loss of Safety Function | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000410/LER-2012-003 | Suppression Pool Level Below Technical Specification Limit During Mode Change | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000261/LER-2012-003 | Plant Modification Interfered with the Operation of Containment Wide Range Level Indicator | | 05000298/LER-2012-004 | Cooper Nuclear Station 05000298 1 of 4 | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat | 05000410/LER-2012-004 | Manual Reactor Scram due to a Loss of Main Turbine Gland Sealing Steam Resulting in Lowering Condenser Vacuum | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000382/LER-2012-004 | Essential Chiller Oil Leak Creates Unanalyzed Past Operability Condition | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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 | 05000265/LER-2012-004 | Drywell Radiation Monitor Failed Downscale | | 05000261/LER-2012-004 | Reactor Tripped Due to a Turbine Trip Caused by a Feedwater Isolation Signal from Steam Generator 'B' High Level | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000454/LER-2012-004 | Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Repair Surface Indications | | 05000254/LER-2012-004 | Breech in Secondary Containment | | 05000482/LER-2012-004 | Two Charging Pumps Capable of Injecting into the RCS Due to Inadequate Definition of Centrifugal Charging Pump in LCO 3.4.12 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000410/LER-2012-005 | Automatic Diesel Actuation Due to the Loss of a 115 kV Offsite Power Source | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000298/LER-2012-005 | Cooper Nuclear Station 05000298 1 of 4 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000298/LER-2012-006 | Missing Vent Plug Results in Technical Specifications Prohibited Condition | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000220/LER-2012-007 | High Pressure Coolant Injection System Logic Actuation Following an Automatic Turbine Trip Signal Due to High Reactor Water Level | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
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