At 12:37, on April 23, 2004, Progress Energy Florida, Inc., Crystal River Unit 3, was in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER. During performance of surveillance procedure SP-354A, "Monthly Functional Test of the Emergency Diesel Generator EGDG-1A," EGDG-1A did not start and achieve steady state voltage and frequency in less than or equal to 10 seconds from standby conditions upon receipt of a simulated Engineered Safeguards signal as required by Technical Specifications. The cause for this event was fuel oil header outlet check valve DFV-61 leaking by its seat due to foreign material lodged between the valve disk and seat. This condition allowed fuel oil to gradually drain back through the header outlet to the fuel oil tank, partially depleting the fuel header prime. The check valve was replaced. Based on engineering judgment, EGDG-1A was inoperable for a period of time longer than allowed by Technical Specifications. This condition is being reported under 10CFR50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications. This condition does not represent a reduction in the public health and safety. No previous similar occurrences have been reported to the NRC by CR-3. |
LER-2004-002, Emergency Diesel Generator Inoperable Due To Fuel Oil Header Outlet Check Valve Leaking Past SeatDocket Number |
Event date: |
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10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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3022004002R00 - NRC Website |
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At 12:37, on April 23, 2004, Progress Energy Florida, Inc., (PEF) Crystal River Unit 3 (CR-3) was in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER. During performance of surveillance procedure SP-354A, "Monthly Functional Test of the Emergency Diesel Generator EGDG-1A," EGDG-1A [EK, DG] did not start and achieve steady state voltage and frequency in less than or equal to 10 seconds from standby conditions upon receipt of a simulated Engineered Safeguards (ES) signal as required by Improved Technical Specification (ITS) Surveillance Requirement 3.8.1.6. The fast start time was 11.4 seconds (average of two stop watch measurements (11.5 seconds and 11.7 seconds) and the installed electronic timer (11.1 seconds)).
EGDG-1A was shut down in accordance with SP-354A.
The actions of ITS 3.8.1, Condition B, "One EDG inoperable," were previously entered for the performance of SP-907A, "Monthly Functional Test of 4160V ES Bus "A" Undervoltage and Degraded Grid Relaying." Those actions remained applicable following the unacceptable EGDG-1A fast start testing.
Troubleshooting the EGDG-1A start problem revealed fuel oil header outlet check valve DFV-61 [EK, V] leaking by its seat. This condition allowed fuel oil to gradually drain back through the header outlet to the fuel oil tank, partially depleting the fuel header prime (e.g., not maintaining EGDG fuel header full of fuel in standby conditions). DFV-61 had been replaced as part of a preventive maintenance activity during the most recent performance of refueling interval engine/generator inspections in accordance with maintenance procedure MP-499, "Emergency Diesel Generator Engine Inspection/Maintenance," in February 2004.
Evidence to establish a time and date for when EGDG-1A became unable to demonstrate a successful fast start test result is limited. However, the following information is known: (1) DFV-61 was replaced with a new component that had been bench tested successfully on receipt inspection and was placed in service for the first time on February 26, 2004; (2) the most recent successful completion of the fast start surveillance test was documented on February 28, 2004; (3) the most recent documented demonstration of adequate fuel oil header prime was upon shutdown of EGDG-1A from its most recent run on March 26, 2004; and, (4) the failure mode of the fuel oil header loosing prime is a standby phenomenon and could not have happened immediately upon the start demand on April 23, 2004. Therefore, loss of adequate fuel oil header prime occurred between March 26, 2004, and April 23, 2004.
EGDG-1A was returned to service at 0457 on April 25, 2004. In order for this event to be determined not reportable, EGDG-1A would have had to become inoperable on or after 0457 on April 22, 2004, the day before performing SP-354A. Based on engineering judgment, EGDG-1A became inoperable at some point prior to that time. Therefore, EGDG-1A is considered to have been inoperable for a period of time longer than the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by ITS 3.8.1, Condition B.
This condition is reportable under 10CFR50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
Two EGDGs automatically provide alternating current (AC) electrical power to 4160 volt (v) ES buses 3A and 3B [EB, BU] following a loss of off-site power or a degraded grid voltage condition.
Power to the 4160v ES buses supplies motive and control power to equipment required for safe shutdown of the plant and for mitigation and control of accidents.
EGDG-1A not starting and achieving steady state voltage and frequency in less than or equal to 10 seconds from standby conditions upon receipt of a simulated ES signal as required by Technical Specifications was reviewed to determine if the Emergency AC Power System safety function was lost at any time between February 28, 2004, and April 23, 2004. The review concluded that the safety function was never lost. Although the 11.4 second start time exceeded ITS requirements, the delay did not challenge the assumptions of the analysis documented in Calculation N04-0002, "CR-3 Loss of Coolant Accident (LOCA) Summary Report," Revision 0, dated January 28, 2004.
Calculation N04-0002, Table 4-1, "LOCA Inputs and Assumptions for CR-3," identifies ES Actuation System (ESAS) [JE] delay times after ESAS Low Reactor Coolant System Trips for Low Pressure Injection (LPI) [BP] and High Pressure Injection (HPI) [BO] for a Small Break LOCA (SBLOCA) and a Large Break LOCA (LBLOCA). The bounding figure (the minimum assumed delay) is the assumed delay of 35 seconds for LPI during a LBLOCA. (SBLOCA assumes a delay of 40 seconds for LPI and 67 seconds for HPI). The EGDG-1A start time of 11.4 seconds does not challenge that assumption. Adding the 11.4 seconds EGDG-1A start time to the LPI initiation time delay of 15 seconds and LPI pump start time of 5 seconds equals a value of 31.4 seconds. This value is less than the 35 second assumption.
Therefore, had the EGDG-1A slow fast start time condition existed at the same time EGDG-1B was voluntarily removed from service, or if EGDG-1B had become unavailable for any other reason, the safety function of the Emergency AC Power System would not have been lost. One train of the Emergency AC Power System would have been available at all times.
The impact of the identified condition is limited to EGDG-1A. EGDG-1B has demonstrated successful fast start test response times, and engine monitoring data supports the conclusion that the EGDG-1B fuel oil header is remaining primed in standby conditions.
Based on the above discussion, PEF concludes that the EGDG-1A slow fast start time did not represent a reduction in the public health and safety. This event does not meet the Nuclear Energy Institute definition of a Safety System Functional Failure (NEI 99-02, Revision 2).
CAUSE
The cause for this event was fuel oil header outlet check valve DFV-61 leaking by its seat. DFV-61 is a 1/8 inch check valve, Model No. 484-4B-28-2, manufactured by Parker Hannifin Corporation - Hydraulic Valve Division (formerly Teledyne - Republic). This check valve prevents back flow and maintains the pressure boundary of the fuel oil system for the EGDGs. Results of a laboratory analysis performed at the Harris Nuclear Plant Environmental and Engineering Laboratory indicate that the failure mechanism was valve leakage due to foreign material lodged between the valve disk and seat.
Laboratory analysis indicates that the primary material composition for each of the various particles falls into one of three categories: clear PolyVinylChloride (PVC); Teflon; and, Iron Oxide. Of the three identified types of foreign material, two are adequately sized and located to be disruptive to the operation of the valve. The clear PVC segment was found lying directly across the molded synthetic rubber seal. The shape of the material is irregular, but narrow and cylindrical, somewhat like the head of an irregular needle or barb. The PVC formed an obstruction that could have been a significant contributor to the valve's malfunction.
The Teflon material was found lying across a portion of the seat area. The Teflon material is a thin, elongated filament, similar to thread sealing tape. This item's size and location indicates that it may also have been interfering significantly with the seating of the valve.
A small red dot found near the PVC has been identified as a fragile Iron Oxide particle. The Iron Oxide particle size is sufficiently small that it was not a significant contributor to the event.
PEF concludes that unintentional introduction of foreign material into the fuel system components during replacement component manufacturing, during component receipt inspection and/or during maintenance activities on February 24, 2004, resulted in the blockage of DFV-61, culminating in the slow fast start of EGDG-1A on April 23, 2004. The most probable source of the clear PVC and the Iron Oxide particle is the transfer process used for the initial fill of the fuel filter cartridges, with the possibility existing that it could also have been introduced in trouble-shooting after the failure.
The most probable source of the Teflon material is concluded to be the manufacturing process, with the possibility existing that it could also have been introduced during receipt inspection.
CORRECTIVE ACTIONS
1. DFV-61 was replaced, SP-354A was performed successfully, and ITS 3.8.1, Condition B, actions were exited at 0457 on April 25, 2004.
2. A review of EGDG-1B fast start test response times and engine monitoring data supports the conclusion that the fuel oil header had been adequately primed in standby conditions.
3. Other actions associated with this event and probable sources of the foreign materials are being addressed in the CR-3 Corrective Action Program in Nuclear Condition Report 125149.
PREVIOUS SIMILAR EVENTS
No previous similar events involving loss of EGDG fuel oil header prime due to foreign material in a fuel oil header discharge check valve have been reported to the NRC by CR-3.
- notify the Supervisor, Licensing & Regulatory Programs of any questions regarding this document or any associated regulatory commitments.
RESPONSE COMMITMENT DUE DATE
SECTION
No regulatory commitments are being made in this submittal.
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05000348/LER-2004-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000333/LER-2004-001 | Inadvertent Actuation of ECCS and EDGs While in Refueling Mode | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000306/LER-2004-001 | | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000315/LER-2004-001 | Failure To Comply With Technical Specification 3 .7 .5 .1, Control Room Emergency Ventilation System | | 05000301/LER-2004-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000313/LER-2004-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000285/LER-2004-001 | Failure To Perform A Leakage Test Due To Lack Of Understanding of Penetration Design | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000271/LER-2004-001 | Main Steam Isolation Valve Leakage Exceeds a Technical Specification Leakage Rate Limit | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000266/LER-2004-001 | | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000318/LER-2004-001 | . Reactor Trip Due to Low Steam Generator Water Level After Feed Pump Trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000289/LER-2004-001 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000263/LER-2004-001 | | | 05000247/LER-2004-001 | Manual Reactor Trip Due to Oscillating Feedwater Flow and Steam Generator Level with Flow Perturbations Caused by a Degraded Feed Water Regulating Valve | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv), System Actuation | 05000244/LER-2004-001 | Gaps in the Control Room Emergency Zone Boundary | | 05000255/LER-2004-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000364/LER-2004-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000461/LER-2004-001 | Clinton Power Station 05000461 1 OF 4 | | 05000414/LER-2004-001 | relPowere Vice President A Duke Energy Company Duke Power Catawba Nuclear Station 4800 Concord Rd. / CNO1VP York, SC 29745-9635 803 831 4251 803 831 3221 fax November 9, 2004
U. S. Nuclear Regulatory Commission
ATTENTION: Document Control Desk
Washington, DC 20555-0001
SUBJECT: Duke Energy Corporation
Catawba Nuclear Station Unit 2
Docket No. 50-414
Licensee Event Report 414/04-001 Revision 0
Reactor Coolant System Pressure Boundary Leakage
Due to Small Cracks Found in Steam Generator
Channel Head Bowl Drain Line on 2C & 2D Steam
Generators
Attached please find Licensee Event Report 414/04-001
Revision 0, entitled "Reactor Coolant System Pressure
Boundary Leakage Due to Small Cracks Found in.Steam
Generator Channel Head Bowl Drain Line on 2C & 2D Steam
Generators."
This Licensee Event Report does not contain any regulatory
commitments. Questions regarding this Licensee Event Report
should be directed to R. D. Hart at (803) 831-3622.
Sincerely,
Dhiaa Jamil
Attachment
www.dukepower.corn 00- U.S. Nuclear Reguldhory Commission
November 9, 2004
Page 2
XC: W.D. Travers
U.S. Nuclear Regulatory Commission
Regional Administrator, Region II
Atlanta Federal Center
61 Forsyth St., SW, Suite 23T85
Atlanta, GA 30303
E.F. Guthrie
Senior Resident Inspector (CNS)
U.S. Nuclear Regulatory Commission
Catawba Nuclear Station
S.E. Peters (addressee only)
NRC Project Manager (CNS)
U.S. Nuclear Regulatory Commission
One White Flint North, Mail Stop 10-B3
11555 Rockville Pike
Rockville, MD 20852-2738
NRC FORM 366� U.S. NUCLEAR REGULATORY APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004)� COMMISSION 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. Nuclear Regulatory Commission, WashinMon, DC 2055
LICENSEE EN/ENT REPORT (LER) 0001, or by Internet e-mail to infocollects@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104). Office of Management(See reverse for required number of and Budget, Washington, DC 20503. If a means used to impose an Information col ectiond( inverse �for each block) does not display a currently valid OMB control number, the NRC may not conduct or sponsor. and a person Is not required to respond o. the Information collection. 1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE Catawba Nuclear Station, Unit 2 050- 00414 1 OF�6 4. TITLE Reactor Coolant System Pressure Boundary Leakage Due to Small Cracks Found in
Steam Generator Channel Head Bowl Drain Line on 2C & 2D Steam Generators | | 05000368/LER-2004-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000423/LER-2004-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000334/LER-2004-001 | 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 | 05000413/LER-2004-001 | Gas Accumulation in Centrifugal Charging Pump Suction Piping | 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 50.73(a)(2)(ii)(A), Seriously Degraded | 05000395/LER-2004-001 | Reactor Trip Due to Valve Failure During Forced Shutdown | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000390/LER-2004-001 | Automatic Reactor Trip Due to a Invalid Turbine Trip Signal (P-4) | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000369/LER-2004-001 | Auxiliary Feedwater System in prohibited condition due to inadequate procedure. | | 05000454/LER-2004-001 | Exelent
Exelon Generation Company, LLCRwww.exeloncorp.com NuclearByron Station 4450 North German Church Road Byron, IL 61010-9794 October 17, 2004 LTR: BYRON 2004-0111 File: 2.01.0700 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Subject:RLicensee Event Report (LER) 454-2004-001-00, "Reactor Containment Fan Coolers Flow Rates Below Technical Specification Requirements Due to Inaccurate Flow Indication" Byron Station, Unit 1
Facility Operating License No. NPF-37
NRC Docket No. STN 50-454
Enclosed is an LER involving the August 17, 2004, event involving low flow conditions discovered in Unit 1 Reactor Containment Fan Coolers for a time period longer than allowed by the Technical Specifications. This event is reportable to the NRC in accordance with 10CFR 50.73 (a)(2)(i)(B), as a condition prohibited by Technical Specifications. Should you have any questions concerning this matter, please contact Mr. William Grundmann, Regulatory Assurance. Manager, at (815) 234-5441, extension 2800. Respectfully, Stephen E. Kuczynski Site Vice President Byron Nuclear Generating Station Attachment LER 454-2004-001-00 cc:RRegional Administrator, Region III, NRC NRC Senior Resident Inspector— Byron Station NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004 (7.2001) COMMISSION Estimated burden per response to comply with this mandatory information 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 Management Branch (T.6 E6), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet e-mail toLICENSEE EVENT REPORT (LER) *I@ nrc.gov, and to the Desk Officer, Office of Informabon and Regulatory Affairs, NEOB:10202 (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose 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 rand ITV NAUP o natuerr An warn q par= . Byron Station, Unit 1 0500454 1 OF 5 4. Reactor Containment Fan Coolers Flow Rates Below Technical Specifications Requirements Due to Inaccurate Flow Indication | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000305/LER-2004-001 | | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | 05000336/LER-2004-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000364/LER-2004-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000423/LER-2004-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000237/LER-2004-002 | Dresden Nuclear Power Station Unit 2 05000237 1 of 5 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000261/LER-2004-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000244/LER-2004-002 | Consolidated Rod Storage Canister Placed in Incorrect Storage Location | | 05000530/LER-2004-002 | Main Turbine Control System Malfunction Results in Automatic Reactor Trip on Low DNBR | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000305/LER-2004-002 | | | 05000247/LER-2004-002 | Manual Reactor Trip Due to Decreasing 23 Steam Generator Level Caused by Feedwater Regulating Valve Closure Due to a De-energized Solenoid Operated Valve from Wiring Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000414/LER-2004-002 | Manual Reactor Trip Initiated Due to Control Rods from Shutdown Bank D Dropping into the Core | | 05000251/LER-2004-002 | AAA | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv)(b) | 05000397/LER-2004-002 | | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000346/LER-2004-002 | Reactor Trip During Reactor Trip Breaker Testing Due To Fuse Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000395/LER-2004-002 | Emergency Diesel Generator Start and Load Due to Momentary Fault on Incoming Feed | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000269/LER-2004-002 | eif Powere RON A. JONES Vice President A Duke Energy Company Oconee Nuclear Site Duke Power ONO1VP / 7800 Rochester Highway Seneca, SC 29672 864 885 3158 864 885 3564 fax September 9, 2004
U.S. Nuclear Regulatory Commission
Document Control Desk
Washington, D.C. 20555
Subject: Oconee Nuclear Station
Docket Nos. 50-269,-270, -287
Licensee Event Report 269/2004-02, Revision 1
Problem Investigation Process No.: 0-04-2808
Gentlemen:
Pursuant to 10 CFR 50.73 Sections (a)(1) and (d), attached
is Licensee Event Report 269/2004-02, Revision 1, regarding
a Main Steam Line Break mitigation design/analysis
deficiency which could result in the main and startup
feedwater control valves being technically inoperable for
mitigation of some steam line break scenarios.
This report is being submitted to supplement Revision 0
submitted July 6, 2004. At that time the root cause
investigation and an analysis of the consequences of
potentially exceeding the Environment Qualification (EQ)
envelope curve were still in progress.
This event is being reported in accordance with 10 CFR
50.73 (a)(2)(i)(B) as a condition prohibited by Technical
Specifications, 50.73(a)(2)(ii)(B) as an Unanalyzed
Condition, and 50.73(a)(2)(V)(D) as a potential loss of
safety function for Accident Mitigation. This event is
considered to be of no significance with respect to the
health and safety of the public.
www.dukepower.corn Document Control Desk
Date: September 9, 2004
Page 2
Attachment: Licensee Event Report 269/2004-02, Revision 1
cc: Mr. William D. Travers
Administrator, Region II
U.S. Nuclear Regulatory Commission
61 Forsyth Street, S. W., Suite 23T85
Atlanta, GA 30303
Mr. L. N. Olshan
Project Manager
U.S. Nuclear Regulatory Commission
Office of Nuclear Reactor Regulation
Washington, D.C. 20555
Mr. M. C. Shannon
NRC Senior Resident Inspector
Oconee Nuclear Station
INPO (via E-mail)
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004 (7-2001) COMMISSION Estimated burden per response to comply with this mandatory information 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 Management Branch (T-6 E6), U.S. Nuclear Regulatory Commission, Washington. DCLICENSEE EVENT REPORT (LER) 20555-0001, or by Internet e-mail to bpi @nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0104), Office of Management and(See reverse for required number of Budget, Washington, DC 20503. If a means used to impose information collection doesdigits/characters for each block) not display a currently valid OMB control number, the NRC may not conduct or sponsor, and,1 nnmnn lc not rent Owl to tocnnni-I to the intnrmatinn rntlentinn 1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE Oconee Nuclear Station, Unit 1 050-81 OF 0269 11 4. TITLE Main Steam Line Break Mitigation Design/Analysis Deficiency | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000313/LER-2004-002 | | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000271/LER-2004-002 | Special Nuclear Material Inventory Location Discrepancy | | 05000285/LER-2004-002 | Inoperable Diesel Generator for 28 Days Due to Blown Fuse During Shutdown | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000336/LER-2004-002 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000311/LER-2004-002 | Failure To Comply With Technical Specifications During Reactor Protection Instrument Calibration . | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000254/LER-2004-002 | Quad Cities Nuclear Power Station Unit 1 05000254 1 of 4 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000263/LER-2004-002 | | | 05000302/LER-2004-002 | Emergency Diesel Generator Inoperable Due To Fuel Oil Header Outlet Check Valve Leaking Past Seat | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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