08-02-2012 | On June 4, 2012, at 0517, Nine Mile Point Unit 2 ( NMP 2) entered Mode 2 (startup) with suppression pool water level at 199.44 feet, below the minimum required level of 199.5 feet, per Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.2.2. Contrary to the requirements of LCO 3.0.1, the conditions for changing modes from Mode 4 (cold shutdown) to Mode 2 were not met when Mode 2 was entered. The low suppression pool level of 199.4 feet was discovered during shift checks on June 4, 2012 at 0846, when TS 3.6.2.2, Condition A was entered. Suppression pool water level was restored at 0926 and TS 3.6.2.2 Condition A was exited at 0933. The cause of this event is a failure to recognize abnormalities. The operators performing and verifying the Surveillance Requirements (SRs) and control room supervision reviewing the SRs did not recognize that little margin remained to the TS required lower level for suppression pool water level. Actions are being taken to communicate lessons learned from this event with operating crews for both units at Nine Mile Point Nuclear Station ( NMPNS) with an emphasis on operator fundamentals of plant parameter monitoring and control. This event was entered into the NMPNS corrective action program (Condition Report CR-2012-005507). |
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LER-2012-003, Suppression Pool Level Below Technical Specification Limit During Mode ChangeDocket Number |
Event date: |
06-04-2012 |
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Report date: |
08-02-2012 |
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Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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Initial Reporting |
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4102012003R00 - NRC Website |
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I. DESCRIPTION OF EVENT
A. PRE-EVENT PLANT CONDITIONS:
Prior to this event, Nine Mile Point Unit 2 (NMP2) was in Mode 4 (cold shutdown) with no inoperable systems affecting this event.
B. EVENT:
On June 4, 2012, at 0517, NMP2 entered Mode 2 (startup) with suppression pool water level at 199.44 feet, below the minimum required level of 199.5 feet, per Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.2.2, "Suppression Pool Water Level." Contrary to the requirements of LCO 3.0.1, the conditions for changing modes from cold shutdown to startup were not met when Mode 2 was entered.
There was no impact on Nine Mile Point Unit 1 (NMP1) from this event.
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE
EVENT:
There were no inoperable components or systems that contributed to this event.
D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES
5/31/2012; 0800 - Suppression pool level is 199.95 feet per computer point CMSLA02 (Suppression Pool Narrow Range Level). The control room log reading is 200 feet, taken from a separate level transmitter.
5/31 to 6/3/2012 - The shift checks log is completed between 0800 and 1000 for day shift and between 2000 and 2200 for night shift. The shift checks log documents completion of TS Surveillance Requirement (SR) 3.5.2.2.a. The shift checks log documents a lowering trend in suppression pool water level from 199.9 feet to 199.7 feet. SR 3.5.2.2.a verifies minimum suppression pool water level required for the High Pressure Core Spray (HPCS) System during Modes 4 and 5.
6/3/2012; 2000 - The shift checks log and daily checks log are started. Both of the logs record the suppression pool water level as 199.55 feet. The daily checks log documents completion of TS SR 3.6.2.2.1 to verify suppression pool water level is within limits during Modes 1, 2, and 3. The daily checks log has two notes; one noting that the surveillance was being performed in Mode 4, and the second noting that this surveillance is being performed in preparation for the mode change to Mode 2.
6/3/2012; 2152 - From Plant Information (PI) system data, the suppression pool water level, computer point CMSLA02, goes below 199.5 feet. This computer point is not the TS required shift checks log reading. NMP2 is in Mode 4 at this time. Per SR 3.5.2.2.a the required suppression pool water level for Mode 4 is greater than or equal to 195 feet. The higher level of greater than or equal to 199.5 feet is per SR 3.6.2.2.1, and required in Modes 1, 2 and 3.
6/4/2012; 0517 - Reactor mode switch change from Mode 4 to Mode 2.
6/4/2012; 0846 - The shift checks log identified the suppression pool water level is 199.4 feet. The suppression pool water level is declared not within limits and TS 3.6.2.2, Condition A, Required Action A.1 is entered to restore suppression pool water level to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
6/4/2012; 0906 - Started 2CSH*P1 (High Pressure Core Spray Pump) to restore suppression pool water level.
6/4/2012; 0926 - Secured 2CSH*P1 following suppression pool filling. Suppression pool water level is 199.9 feet.
6/4/2012; 0933 - Declared suppression pool water level within limits and exited TS 3.6.2.2, Condition A.
E. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:
None
F. METHOD OF DISCOVERY:
At the time of discovery, June 4, 2012, at 0846, NMP2 was in Mode 2 with the reactor critical, below the point of adding heat, when the shift checks log identified suppression pool water level at 199.4 feet. The low suppression pool water level resulted in an unplanned entry into TS 3.6.2.2 Condition A to restore level within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Based on a review of relevant information using control room logs and the PI system, there is firm evidence that the discrepancy existed before the time of discovery. PI data clearly shows a lowering trend of suppression pool water level for the preceding 4 days and indicates that the suppression pool water level was below 199.5 feet at the time the mode switch was changed from Mode 4 to Mode 2 on June 4, 2012, at 0517.
G. MAJOR OPERATOR ACTION:
Upon discovery of the condition, TS 3.6.2.2 Condition A was entered for suppression pool water level below 199.5 feet. Suppression pool water level was restored at 0926. TS 3.6.2.2 Condition A was exited at 0933.
H. SAFETY SYSTEM RESPONSES:
None. No operational conditions requiring the response of safety systems occurred as a result of this event.
II. CAUSE OF THE EVENT:
The cause of this event is a failure to recognize abnormalities. The surveillance performers, reviewers and verifiers failed to recognize that although suppression pool water level was within the required band, an abnormal trend existed based on data collected previously. The operators and control room supervision did not recognize that little margin remained to the TS required lower level for suppression pool water level for operation in Mode 2.
This event was entered into the Nine Mile Point Nuclear Station (NMPNS) corrective action program (CR-2012- 005507).
III. ANALYSIS OF THE EVENT:
This event is reportable in accordance with 10 CFR 50.73 (a)(2)(i)(B), as an operation or condition which was prohibited by the plant's Technical Specifications.
LCO 3.0.1 requires that LCOs shall be met during the MODES or other specified conditions in the Applicability.
Contrary to the requirements of LCO 3.0.1, LCO 3.6.2.2, was not met on June 4, 2012, at 0517, when NMP2 entered Mode 2 with suppression pool water level below the minimum required level of 199.5 feet.
At the time of discovery on June 4, 2012 at 0846, TS 3.6.2.2, Required Action A.1 was entered and suppression pool water level was restored at 0926, within the required action completion time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The suppression pool water level was not within limits for 40 minutes from the time of discovery and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 9 minutes from the time that Mode 2 was entered.
After the event, an evaluation of PI data and the shift checks log data was performed to determine suppression pool water level when NMP2 changed from Mode 4 to Mode 2 on June 4, 2012 at 0517. Minimum suppression pool water level for operation in Mode 2 is 199.5 feet. Evaluation of the PI data determined suppression pool water level was at 199.44 feet at the time of mode change. An evaluation of the shift checks log was performed and determined that suppression pool water level was at 199.47 feet at the time of mode change. Both sources of information provide firm evidence that suppression pool water level was below the required minimum limit at the time of mode change.
Per NUREG-1022, Revision 2, Section 3.2.2, an LER is required if a condition existed for a time longer than permitted by the technical specifications even if the condition was not discovered until after the allowable time had elapsed and the condition was rectified immediately upon discovery.
There were no actual safety consequences from this event. For the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 9 minutes that suppression pool water level was below 199.5 feet in Mode 2, reactor coolant system temperature was not above 212 degrees Fahrenheit; thus the pressure suppression capability of the suppression pool would not have been challenged during a Loss of Coolant Accident. Steam was not being generated and a safety relief valve opening would not have challenged the design limit of the suppression pool.
The most probable cause of suppression pool water level lowering from May 31, 2012 to June 4, 2012 is leakage from 2RHS*MOV1B, the Residual Heat Removal (RHR) pump B suction valve from the suppression pool. From May 31, 2012 to June 4, 2012, 2RHS*MOV1B was closed and the RHR B subsystem was operating in the shutdown cooling mode of operation. 2RHS*MOV1B is normally open during Mode 2.
This event does not affect the NRC Regulatory Oversight Process (ROP) Index items.
IV. CORRECTIVE ACTIONS:
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
1. Restored suppression pool water level to 199.9 feet.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
1. A communication will be developed for distribution to the NMP1 and NMP2 operating crews. The communication will discuss lessons learned from this event with an emphasis on operator fundamentals of plant parameter monitoring and control. The lessons learned will stress the importance of verifying margin to allowable limits when recording data, and monitoring for evidence of trends, so that preemptive action may be taken to prevent established limits from being exceeded.
2. Troubleshooting will be performed during the next outage to determine if leakage past 2RHS*MOV1B is present with the valve in the closed position. This troubleshooting is only able to be performed when the RHR B subsystem is in the shutdown cooling mode of operation.
V. ADDITIONAL INFORMATION:
A. FAILED COMPONENTS:
None
B. PREVIOUS LERs ON SIMILAR EVENTS:
None C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION
IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS
LER:
COMPONENT IEEE 803 � IEEE 805
COMPONENT IDENTIFIER SYSTEM IDENTIFICATION
None� NA� NA
None
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| | Reporting criterion |
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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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