10-12-2004 | On 8/14/2004 at 0740 hours0.00856 days <br />0.206 hours <br />0.00122 weeks <br />2.8157e-4 months <br /> during performance of the normally scheduled Control Rod Assembly Partial Movement Surveillance Test at Beaver Valley Power Station ( BVPS) Unit 1, the operator performing the test observed an unexpected condition during the insertion of control rod Shutdown Bank "A" which showed Group II rods at 223 steps withdrawn, two steps higher than expected. When the expected Technical Specification (TS) 3.1.3.5 requires shutdown bank control rods to be withdrawn at least 225 steps except during surveillance testing. There are no applicable action requirements for a whole bank not being withdrawn. At 0810 hours0.00938 days <br />0.225 hours <br />0.00134 weeks <br />3.08205e-4 months <br />, TS 3.1.3.5 and 3.0.3 were entered. At 11:07 hours, the Shutdown Bank "A" rods were withdrawn to 226 steps on the Group II demand counter using the normal rod withdrawal process, exiting TS 3.0.3 and TS 3.1.3.5.
The initiation of preparation for shutdown was completed within one hour of entering TS 3.0.3. Thus, BV1 initiated actions to shutdown the Unit, even though actual power level was not decreased.
Therefore, this event is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by plant Technical Specifications as per NUREG-1022, Rev. 2, page 36. The cause of the Group II Shutdown Bank "A" control rods to fail to move as demanded was a degraded Slave Cycler Logic card inside the Rod Control Logic cabinet. The safety significance of this event was low. All control rods, including the Shutdown Banks, remained trippable throughout the event. |
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LER-2004-001, Control Rod Shutdown Bank Anomaly Causes Entry into Technical Specification 3.0.3Docket Number |
Event date: |
08-14-2004 |
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Report date: |
10-12-2004 |
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Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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3342004001R00 - NRC Website |
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PLANT AND SYSTEM IDENTIFICATION
Westinghouse-Pressurized Water Reactor {PWR} Control Rod Drive System {AA}
CONDITIONS PRIOR TO OCCURRENCE
Unit 1: Mode 1 at 100 percent power There were no systems, structures, or components that were inoperable at the start of the event that contributed to the event other than as described below.
DESCRIPTION OF EVENT
On 8/14/2004, at 0740 hours0.00856 days <br />0.206 hours <br />0.00122 weeks <br />2.8157e-4 months <br />, during performance of the normally scheduled Control Rod Assembly Partial Movement Surveillance Test at Beaver Valley Power Station (BVPS) Unit 1, the operator performing the test observed an unexpected condition during the insertion of control rod Shutdown Bank "A". This test is performed to satisfy BVPS Unit 1 Technical Specification (TS) Surveillance Requirement (SR) 4.1.3.1.1, which requires that each shutdown and control rod not fully inserted in the core to be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days. The surveillance first tests Shutdown Bank "A" by inserting the bank 10 steps. Rod positions at the start of the test were at the all-rods-out position for the current cycle, 226 steps withdrawn. Each of the two control rod Groups within Shutdown Bank "A" normally alternates stepping in(out) one step during normal continuous rod motion. The first insertion of the Shutdown Bank "A" rods was 3 1/2 steps, with the group demand counters for Shutdown Bank "A" showing Group I at 222 steps withdrawn and Group II at 223, as expected. In the next rod move, Group I moved in a step instead of the expected Group II, now showing Group I at 221 and Group II at 223 steps withdrawn. This resulted in having a two step difference between the two control rod Groups which was not expected, and, therefore, rod motion was immediately stopped. No rod control alarms were received.
When the expected response to the rod motion was not received, the test was stopped.
Therefore, the TS surveillance exception was no longer applicable. Further rod motion was not attempted due to conservative decision considerations concerning potential adverse control rod consequences, awaiting evaluation by station Instrument and Control technicians.
BVPS Unit 1 TS 3.1.3.5 requires that all shutdown rods meet the insertion limits specified in the Core Operating Limits Report (COLR) in Modes 1 and 2. [The COLR insertion limit is shutdown rods withdrawn at least 225 steps.] TS 3.1.3.5 also requires that with a maximum of one shutdown rod inserted beyond the insertion limit, except for surveillance testing pursuant to SR 4.1.3.1.1, the rod shall be restored to within the limit within one hour or declare the rod inoperable and apply TS 3.1.3.1. The Action requirements of TS 3.1.3.5 allows for only one rod to be inoperable. There are no applicable action requirements for a whole bank not being appropriately withdrawn. Thus, at 0810 hours0.00938 days <br />0.225 hours <br />0.00134 weeks <br />3.08205e-4 months <br />, TS 3.1.3.5 and 3.0.3 were entered, which requires restoration within one hour or place the unit in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
TS 3.1.3.2 was also entered since the operation of the control rod group demand counters in the control rod position indication system were believed to be potentially inoperable. TS 3.1.3.2 requires control rod group demand counter rod position indication be operable. With one group demand position indicator inoperable, TS 3.1.3.2 Action b.1 requires all rod position indicators for the affected bank be verified operable and that the rods are within 12 steps of each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
At 0910 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.46255e-4 months <br />, Shutdown Bank "A" had not been restored to meet the COLR requirements.
Action was initiated to commence plant shutdown per TS 3.0.3 with the requirement that the plant must be in Hot Standby by 15:10 hours. Power reduction was scheduled to commence at 11:10 hours.
Since no anomaly was identified within the control rod power cabinet and logic cabinet, at 11:07 hours, the Shutdown Bank "A" rods were withdrawn to 225 steps on the Group I demand counter and to 226 steps on the Group II demand counter using the normal rod withdrawal process. With all Control Rods being returned to acceptable Rod Insertion Limit positions, TS 3.0.3 and TS 3.1.3.5 were exited. TS 3.1.3.2 remained in effect as the group demand counter was the initial suspected cause of the event.
On 8/16/2004, the group demand counter was determined to be functioning properly, but the Unit conservatively remained in TS 3.1.3.2 Action b.1.
On 8/17/2004, additional testing indicated a continued rod movement problem, which was determined to be due to a lack of demand signal being continuously generated somewhere within the rod control system. Although subsequently shown not to be required, BVPS Unit 1 conservatively entered TS 3.1.3.1 Action d at 1408 hours0.0163 days <br />0.391 hours <br />0.00233 weeks <br />5.35744e-4 months <br /> due to declaring more than one control rod trippable but inoperable due to the inability to complete the surveillance at that time. TS 3.1.3.1 Action d states: With more than one rod trippable but inoperable, power operation may continue provided that the remainder of the rods in the bank(s) are aligned to within 12 steps while maintaining the rod sequence and insertion limits provided within the COLR and the inoperable rods are restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
On 8/19/04, following replacement of the Slave Cycler logic card, rod control pulse voltage returned to the expected value. At 0425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br />, following successful completion of the surveillance test pursuant to SR 4.1.3.1.1, TS 3.1.3.1 and 3.1.3.2 were exited.
REPORTABILITY
Pursuant to NUREG-1022, Rev. 2, page 36: "Entry into STS 3.0.3 is not necessarily reportable under 10 CFR 50.73(a)(2)(i)(B). However, it should be considered reportable under this criterion if the condition is not corrected within an hour, such that it is necessary to initiate actions to shutdown, cooldown, etc." Based upon this NUREG-1022 criteria, any event where Tech Spec 3.0.3 is applied longer than one hour would be reportable under 10 CFR 50.73.
As noted in the description above, BVPS Unit 1 entered TS 3.0.3 and remained in TS 3.0.3 for more than one hour. Therefore, this event is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by plant Technical Specifications because TS 3.0.3 had been entered for longer than one hour, even though actual power level was not decreased from 100 percent.
CAUSE OF EVENT
The cause of the Group II Shutdown Bank "A" control rods to not move as demanded was the result of a degraded Slave Cycler Logic card inside the Rod Control Logic cabinet.
ANALYSIS OF EVENT
The Slave Cycler Logic card was determined to be degraded. The as found pulse voltage to the Slave Cycler Logic card for the associated AC power cabinet was 9 volts or less (6.5 volts to 8.5 volts is a voltage transition region which may deter rod stepping). The correct voltage should be 14 volts. This degraded condition affected the ability of the Group II Shutdown Bank "A" control rods to move under manual or automatic control since movement demand was intermittent due to the degraded voltage. The ability to trip these control rods was not affected.
SAFETY IMPLICATIONS
The control rods associated with Group II Shutdown Bank "A" remained trippable. The degradation caused intermittent Group II movement due to the intermittent logic gate safety analyses in the BVPS Unit 1 Updated Final Safety Analysis Report (UFSAR). The UFSAR only credits the ability of the control rods to trip in the Design Basis Accident safety UFSAR had remained valid.
The plant risk associated with the BVPS Unit 1 Shutdown Bank "A" Group 2 position anomaly that occurred on August 14, 2004 is considered to be low. This is based on the incremental core damage probability for the event when considering manual rod insertion to be unavailable during the time period.
Based on the above, the safety significance of the Shutdown Bank "A" Group 2 position anomaly on August 14, 2004 was low.
CORRECTIVE ACTIONS
1. An initial investigation determined that Shutdown Bank "A" Group I and II control rods had remained trippable.
2. A subsequent investigation identified that a failure in a Slave Cycler Logic card caused intermittent failure to move upon demand of Shutdown Bank "A", Group II. Following replacement of the failed Slave Cycler Logic card and the Master Cycler Selector Card, the surveillance test for SR 4.1.3.1.1 was performed successfully.
3. The Rod Control Preventive Maintenance frequency for both BVPS Units is being evaluated for optimal reliability.
4. Additional actions are being evaluated to address replacing obsolete Rod Control System cards with upgraded cards (as they become available) to help prevent age related card failures.
Completion of the above and other corrective actions are being tracked through the BVPS corrective action program.
PREVIOUS SIMILAR EVENTS
A review found no Beaver Valley Power Station Licensee Event Reports within the last three years involving control rod manual movement or control rod position indication.
COMMITMENTS
There are no new commitment made by FirstEnergy Nuclear Operating Company (FENOC) for BVPS Unit No. 1 in this document.
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| | Reporting criterion |
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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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