On May 28, 2005 at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> during Operator walkdown of the main control board, the oncoming nightshift crew discovered the "A" Motor Driven Emergency Feedwater Pump (MDEFP) switch was in Pull To-Lock ( PTL) with the plant in Mode 3. South Carolina Electric & Gas Company (SCE&G) was performing start-up of V. C. Summer Nuclear Station ( VCSNS) following the recent refueling outage. The plant entered Mode 3 on 5/28/05 at 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br />. In Mode 3 the Emergency Feedwater ( EFW) pumps are required to be Operable. The "A" MDEFP switch was apparently placed in Pull-To-Lock ( PTL) following extensive retesting of the EFW pumps over the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The last test occurred at approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on 5/28/05. With the "A" MDEFP inoperable, the plant was in a condition prohibited by VCSNS Technical Specification (TS) 3.7.1.2.
Upon discovery, the "A" MDEFP switch was placed in Normal-After-Stop and the pump was declared Operable.
This event is attributed to procedural weaknesses in conducting multiple tasks concurrently for plant startup influenced by insufficient administrative controls. Also, an additional contributing factor was that the Bypass Inoperable Status Indication (BISI) program did not provide a back up indication that the EFW system was out of alignment due to the Integrated Plant Computer System ( IPCS) computer being upgraded and not fully functional at the time of the event. |
LER-2005-002, Vice President, Nuclear Cperations
803.345.4214
U®
�
A SCANA COMPANY April 6, 2006
Document Control Desk
U. S. Nuclear Regulatory Commission
Washington, DC 20555
Dear Sir or Madam:
Subject VIRGIL C. SUMMER NUCLEAR STATION
DOCKET NO. 50-395
OPERATING LICENSE NO. NPF-12
LICENSEE EVENT REPORT (LER 2005-002-01)
MODE 3 ENTRY WITH AN INOPERABLE EMERGENCY FEEDWATER PUMP
Attached is supplemental Licensee Event Report (LER) No. 2005-002-01, for the Virgil C.
Summer Nuclear Station (VCSNS). This report identifies a plant condition prohibited by
VCSN S Technical Specifications and is being submitted in accordance with 10 CFR
73(a)(2)(i)(B). Revisions are identified by vertical bars in the right side margin of the
attached.
Should you have any questions, please call Mr. Robert G. Sweet at (803) 345-4080.
Very truly yours,
ffrey B. Arc le
JT/JBA/dr
Attachment
c: N. 0. Lorick D. L. Abstance
S. A. Byrne K. M. Sutton
N. S. Cams EPIX Coordinator
J. H. Hamilton (w/o attachment) INPO Records Center
R. J. White J&H Marsh & McLennan
W. D. Travers NSRC
R. E. Martin RTS (C-05-2300)
NRC Resident Inspector File (818.07)
P. Ledbetter DMS (RC-06-0069)
SCE&G I Virgil C. Summer Nuclear Station • P. 0. Box 88 • Jenkinsville, South Carolina 29065 • T (803) 3453209 • www.sccna.com
APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004NRC FORM 366 U.S. NUCLEAR REGULATORY
(7-2001) COMMISSION Estimated burden per response to comply with this mandatory informat on 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
LICENSEE EVENT REPORT (LER) Management Branch (T-6 E6), U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, or by Internet e-mail to bjsl@nrc.gov, and to the Desk Officer, Office of
(Soe reverse for required number of Information and Regulatory Affairs, NEOB-10202 (3150-0104), Office of Managementdigits/characters for each block) and Budget, Washington, DC 20503.
1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE
Virgil C. Summer Nuclear Station 0 5 0 0 0 3 9 5 1 OF
4. TITLE
Mode 3 Entry Wi.:h An Inoperable Emergency Feedwater PumpDocket Number |
Event date: |
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10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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3952005002R01 - NRC Website |
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PLANT IDENTIFICATION
Westinghouse - Pressurized Water Reactor
EOUIPMENT IDENTIFICATION
N/A
IDENTIFICATION OF EVENT
During start-up from Refueling Outage (RF) 15, VCSNS was escalated in power from Mode 4 to Mode 3 on May 28, 2005 at 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br />. At 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />, during Operator walkdown of the Main Control Board (MCB), the oncoming nightshift crew discovered the "A" Motor Driven Emergency Feedwater Pump (MDEFP) switch was in Pull-To-Lock (PTL). VCSNS Technical Specifications (TS) 3.7.1.2 requires that all Emergency Feedwater (EFW) Pumps be operable in Modes 1, 2, and 3.
EVENT DATE
May 28, 2005 Condition Evaluation Report CER 05-2300 was generated to address investigation, cause, and corrective actions associated with this event.
REPORT DATE
July 27, 2005 Original submittal April 6, 2006 Revision 1
CONDITIONS PRIOR TO EVENT
Mode 3, 0% Power — Refueling Outage 15
DESCRIPTION OF EVENT
On May 28, 2005 at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> during Operator walkdown of the MCB, the oncoming nightshift crew discovered the "A" Motor Driven Emergency Feedwater Pump (MDEFP) switch was in Pull-To-Lock (PTL) with the plant in Mode 3.
South Carolina Electric & Gas Company (SCE&G) was performing start-up of V. C. Summer Nuclear Station (VCSNS) following the recent refueling outage. The plant escalated to Mode 3 on 5/28/05 at 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br />. In Mode 3 the Emergency Feedwater (EFW) pumps are required to be Operable. The "A" MDEFP switch was apparently placed in Pull-To-Lock (PTL) following extensive retesting of the EFW pumps over the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The last test occurred at approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on 5/28/05. At 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br />, entry into Mode 3 was declared. With the "A" MIDEFP inoperable, the plant was in a condition prohibited by VCSNS Technical Specification (TS) 3.7.1.2 and subject to report in accordance with 10CFR50.73(a)(2)(i)(B).
CAUSE OF EVENT
This event is attributed to procedural weaknesses in conducting multiple tasks concurrently for plant startup influenced by insufficient administrative controls. Also, an additional contributing factor was that the Bypass Inopera'ple Status Indication (BISI) program did not provide a back up indication that the EFW system was out of alignment due to the Integrated Plant Computer System (IPCS) computer being upgraded and not fully functional at the time of the event.
ANALYSIS OF EVENT
The plant escalated to Mode 3 on 5/28/05 at 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br />. In Mode 3 the EFW pumps are required to be Operable.
The switch for "A" MDEFP was apparently placed in Pull-To-Lock (PTL) following extensive retesting of the EFW pumps over the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The last test occurred at approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on 5/28/05. At 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br />, entry into Mode 3 was declared. With the "A" MDEFP inoperable, the plant was in a condition prohibited by VCSNS Technical Specification (TS) 3.7.1.2.
Upon discovery at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />, the "A" MDEFP switch was placed in Normal-After-Stop and the pump was declared Operable. There was no adverse impact from this event. The system remained functional at all times. Had the "A" MDEFP been required to start, numerous Main Control Board (MCB) alarms would have alerted the operators to the need for the pump and the fact that the pump was not in operation.
The "A" MDEFP had been tested many times over the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and left in PTL. Since testing occurred many times, the pump was left in PTL to facilitate the next test. Due to the multiple tasks being conducted concurrently for plant startup, the "A" MDEFP was not restored from PTL when the mode escalation occurred.
An additional contributing factor to the human error that failed to recognize the PTL on the "A" MDEFP was that the Integrated Plant Computer System (IPCS) computer was replaced this outage. Since the IPCS was not fully functional during this phase of startup the Bypass Inoperable Status Indication (BISI) program did not provide a back up indication that the EFW system was out of alignment.
Condition Evaluation Report (CER) 05-2620 is incorporating this and 2 other human performance CERs which were of high visibility intc a common cause evaluation for problem resolution.
CORRECTIVE ACTIONS
Immediate corrective action was taken to place the "A" MDEFP switch in Normal After Stop. This returned the pump to Operable status.
VCSNS has initiated the following programmatic improvements to install barriers and controls on mode escalation:
- General Operating Procedures (GOPs) are being revised to reference the Surveillance Test Procedure (STP) that needs to be performed prior to applicable mode changes.
- The STP has been revised to restore/ensure EFW pumps and valves are in the required operable position.
- A crew mode change sheet was developed and will be implemented prior to mode changes.
In addition to the above program changes: compensatory actions are being developed to monitor safeguards system status should the BISI system become inoperable.
All corrective actions will be completed prior to the next VCSNS refueling outage (RF-16) currently scheduled to begin in October 2006
PRIOR OCCUR ENCES
No prior occurrences were identified in which an EFW pump was in PTL during Mode 3 entry.
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05000328/LER-2005-001 | Unit 2 Reactor Trip Following Closure of Main Feedwater Upon Inadvertent Opening of Control Breakers | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000388/LER-2005-001 | DDegradation of Primary Coolant Pressure Boundary due to Recirculation Pump Discharge Valve Bonnet Vent Connection Weld Flaw | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000423/LER-2005-001 | | 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 | 05000455/LER-2005-001 | Unit 2 Automatic Reactor Trip Due to Low Steam Generator Level resulting from a Software Fault on the Turbine Control Power Runback Feature | | 05000370/LER-2005-001 | Automatic Actuation of Motor Driven Auxiliary Feedwater Pumps During Outage | | 05000244/LER-2005-001 | Failure of ADFCS Power Supplies Results in Plant Trip | | 05000247/LER-2005-001 | 0Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for One Inoperable Train of ECCS Caused by an Inoperable Auxiliary Component Cooling Water Check Valve | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000529/LER-2005-001 | REACTOR HEAD VENT AXIAL INDICATIONS CAUSED BY DEGRADED ALLOY 600 COMPONENT | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000336/LER-2005-001 | | | 05000266/LER-2005-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000269/LER-2005-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000289/LER-2005-001 | | | 05000293/LER-2005-001 | | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000298/LER-2005-001 | Reactor Scram due to Reactor Level Transient and Inadvertent Rendering of High Pressure Coolant Injection Inoperable | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000331/LER-2005-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000315/LER-2005-001 | Reactor Trip Following Intermediate Range High Flux Signal | | 05000316/LER-2005-001 | Reactor Trip from RCP Bus Undervoltage Signal Complicated by Diesel Generator Output Breaker Failure | 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)(i)(B), Prohibited by Technical Specifications | 05000317/LER-2005-001 | Main Feedwater Isolation Valve Inoperability Due to Handswitch Wiring | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000323/LER-2005-001 | TS 3.4.10 Not Met During Pressurizer Safety Valve Surveillance Testing Due to Random Lift Spread | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000333/LER-2005-001 | Inoperable Offsite Circuit In Excess of Technical Specifications Allowed Out of Service Time | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000352/LER-2005-001 | Loss Of Licensed Material In The Form Of A Radiation Detector Calibration Source | | 05000353/LER-2005-001 | Core Alterations Performed With Source Range Monitor Alarm Horn Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000306/LER-2005-001 | | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000361/LER-2005-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000362/LER-2005-001 | Emergency Diesel Generator (EDG) 3G003 Declared Inoperable Due to Loose Wiring Connection on Emergency Supply Fan | | 05000263/LER-2005-001 | | | 05000456/LER-2005-001 | Potential Technical Specification (TS) 3.9.4 Violation Due to Imprecise Original TS and TS Bases Wording | | 05000454/LER-2005-001 | Failed Technical Specification Ventilation Surveillance Requirements During Surveillance Requirement 3.0.3 Delay Period | | 05000282/LER-2005-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000286/LER-2005-001 | Plant in a Condition Prohibited by Technical Specifications due to Error Making Control Room Ventilation System Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000400/LER-2005-001 | Reactor Auxiliary Building Emergency Exhaust System Single Failure Vulnerability | | 05000395/LER-2005-001 | Emergency Diesel Generator Start and Load Due To A Loss Of Vital Bus | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000382/LER-2005-001 | RCS Pressure Boundary Leakage Due to Primary Water Stress Corrosion Cracking of a Pressurizer Heater Sleeve | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000305/LER-2005-001 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000369/LER-2005-001 | Reactor Coolant System Leakage Detection Instrumentation Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000266/LER-2005-002 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000255/LER-2005-002 | | 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 | 05000361/LER-2005-002 | Missing Taper Pins on CCW Valve Cause Technical Specification Required Shutdown | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000370/LER-2005-002 | Ice Condenser Lower Inlet Door Failed Surveillance Testing | | 05000353/LER-2005-002 | High Pressure Coolant Injection System Inoperable due to a Degraded Control Power Fuse Clip | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000263/LER-2005-002 | | | 05000454/LER-2005-002 | One of Two Trains of Hydrogen Recombiners Inoperable Longer Than Allowed by Technical Specifications Due to Inadequate Procedure | | 05000244/LER-2005-002 | Emergency Diesel Generator Start Resulting From Loss of Off-Site Power Circuit 751 | | 05000362/LER-2005-002 | Emergency Containment Cooling Inoperable for Longer than Allowed by Technical Specifications | 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 | 05000247/LER-2005-002 | DTechnical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for One Inoperable Train of ECCS Caused by Gas Intrusion from a Leaking Check Valve | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000306/LER-2005-002 | | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000265/LER-2005-002 | Main Steam Relief Valve Actuator Degradation Due to Failure to Correct Vibration Levels Exceeding Equipment Design Capacities | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000286/LER-2005-002 | • Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249Entergy Buchanan. NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Vice President Administration July 5, 2005 Indian Point Unit No. 3 Docket Nos. 50-286 N L-05-078 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001 Subject:L Licensee Event Report # 2005-002-00, "Automatic Reactor Trip Due to 32 Steam Generator Steam Flow/Feedwater Flow Mismatch Caused by Low Feedwater Flow Due to Inadvertent Condensate Polisher Post Filter Bypass Valve Closure." Dear Sir: The attached Licensee Event Report (LER) 2005-002-00 is the follow-up written report submitted in accordance with 10 CFR 50.73. This event is of the type defined in 10 CFR 50.73(a)(2)(iv)(A) for an event recorded in the Entergy corrective action process as Condition Report CR-IP3-2005-02478. There are no commitments contained in this letter. Should you or your staff have any questions regarding this matter, please contact Mr. Patric W. Conroy, Manager, Licensing, Indian Point Energy Center at (914) 734-6668. Sincerely, 4F-/t R. Dacimo Vice President Indian Point Energy Center Docket No. 50-286 NL-05-078 Page 2 of 2 Attachment: LER-2005-002-00 CC: Mr. Samuel J. Collins Regional Administrator — Region I U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident Inspector's Office Resident Inspector Indian Point Unit 3 Mr. Paul Eddy State of New York Public Service Commission INPO Record Center NRC FORM 3660 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004) Estimated burden per response to comply with this mandatory collection request 50 hours.RReported lessons teamed are incorporated into the licensing process and fed back to Industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 29555-0001, or by InternetLICENSEE EVENT REPORT (LER) e-mail to Infocoilectsenrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-l0202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person Is not required to respond to, the Information collection. 1. FACIUTY NAME 2. DOCKET NUMBER 3. PAGE INDIAN POINT 3 05000-286 10OF06 4. TITLE Automatic Reactor Trip Due to 32 Steam Generator Steam Flow/Feedwater Flow Mismatch Caused by Low Feedwater Flow Due to Inadvertent Condensate Polisher Post Filter Bypass Valve Closure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000287/LER-2005-002 | Unit 3 trip with ES actuation due to CRD Modification Deficiencies | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000336/LER-2005-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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