On April 8, 2005, NMC staff identified a potential deficiency in the Point Beach Appendix R Safe Shutdown Strategy. A review of the physical routing of Unit 1 Charging Pump 2C ( 1P-2C) control cables indicated that those control cables located within Fire Area A06 would be damaged and prevent remote operation from the control room following postulated fire damage to these control cables. Fire Organizational Plan (FOP) 1.2 was charging pump capability and coolant pump seal cooling.
On June 7, 2005, a continuing review determined that this condition affected available charging pump capability. The condition could have resulted in losing the availability of all but a single charging pump operating at slow speed, which would not provide sufficient seal cooling to the reactor coolant pump (RCP), thereby degrading the level of plant safety.
On April 28, 2005, a similar issue was documented for the Unit 2 charging pumps. Specifically, neither FOP 1.2, nor the safe shutdown Analysis Compliance Strategies provide adequate guidance for manual actions necessary to restart the charging pumps. This is necessary to ensure adequate RCP seal cooling is established or maintained. FOP 1.2 was revised on May 9, 2005, to ensure adequate manual actions are in place to establish and maintain Unit 2 charging pump capability and reactor coolant seal cooling. |
FACILITY NAME (1_1 DOCKET NUMBER (2) I YEAR I SEQUENTIAL REVISION LER NUMBER (6 PAGE (3)
Event Description:
On April 8, 2005, NMC staff identified a potential deficiency in the Point Beach Appendix R Safe Shutdown Charging Pumps' for a fire in Fire Area A06, 1B32 motor control center (MCC2) area.
- A review of the control cables' associated with charging pump 1P-2C indicated that cables were located within Fire Area A06. If a fire occurred in this area and damaged the control cables, the ability to start the pump from the control room would be lost.
- Fire Organization Plan, FOP 1.2 (Revision 6) for fire area A06 stated the strategy that should be used if reactor coolant pump4 (RCP) seal injection flow is lost for 10 minutes. Specifically, it states to ensure seal injection remains isolated. This is in conflict with the Safe Shutdown Analysis, which states that RCP seal cooling must be restored in a timely manner. If a prolonged loss of seal cooling occurs, the RCP seals could fail, resulted in leakage in excess of makeup capability.
charging pump capability and coolant pump seal cooling.
On April 28, 2005, a similar issue was documented for the Unit 2 charging pumps. Specifically, neither FOP 1.2, nor the Safe Shutdown Analysis Compliance Strategies, provided adequate guidance for manual actions necessary to restart these pumps to ensure adequate RCP seal cooling is established.
FOP 1.2 was revised on May 9, 2005 to revision 8, to ensure adequate manual actions are in place to maintain and establish Unit 2 charging pump capability and reactor coolant pump seal cooling.
On June 7, 2005, a continuing review of further information related to this condition, determined that the available charging pump capability was affected. The system would not provide sufficient reactor coolant pump seal cooling, thereby degrading the level of plant safety.
Event Analysis:
Unit 1:
The Safe Shutdown Analysis for this area discusses a problem where control and power cables for all three (3) Unit 1 charging pumps are routed within the fire area. An exemption was provided by the NRC, based upon adequate separation of the power cables between the A train charging pumps (1P-2A and 1P-2B) and the B train pump (1P-2C). For a fire in the vicinity of motor control center 1B32, charging pump 1P-2C is credited, since its associated power cables have adequate separation. However, control cables could be damaged resulting in a loss of speed control. With the loss of control cables, the pump will continue to run at minimum speed, which would be adequate for normal make-up when the RCP seals are not challenged.
The Safe Shutdown Analysis stresses the importance of ensuring continuity of charging, such that the seals are protected from sustained loss of cooling. The coping strategy, contained within the Safe Shutdown Analysis indicates that no manual actions are required outside the control room if control power is lost.
(EIIS Component Identifier: P) 2 (EIIS Component Identifier: MCC) (EIIS Component Identifier: CBL) 4 (EIIS Component Identifier: P) FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) A review of the control cables associated with 1P-2C, indicates that cables are located within Fire Area A06.
If damaged by fire, the pump would not be able to be started from the control room. The Safe Shutdown Analysis did not identify the control cables as required for safe shutdown, and the circuit analysis did not consider the impact of fire damage on the associated control circuit. There are three reasons that the pump may not be running and require a start. First, the pump could be in standby as part of normal pump line-up.
Second, the pump could be tripped on under voltage. Third, the pump could trip due to intra-cable or inter- cable shorts in the control cable. A loss of off-site power is assumed for this scenario. A review of power cables in this fire area confirm that off-site power to Unit 1 could be lost if a fire occurred.
FOP 1.2 (Revision 6) for fire area A06 stated the strategy that should be used if RCP seal injection flow is lost for 10 minutes. Specifically, it requires that seal injection remain isolated. This is in conflict with the Safe Shutdown Analysis, which states that RCP seal cooling must be restored in a timely manner to ensure make-up capability would not be lost as a result of failure of the RCP seals.
Unit 2:
charging pumps. Specifically, a fire in the east side of this Fire Area credits the use of either charging pump 2P-2A or 2P-2B. In this portion of the fire area, power and control power to charging pump 2P-2C is affected and the pump is not credited. The power cables for 2P-2A and 2P-2B are routed through the western portion of the fire area (in the vicinity of 2B-32). An exemption is provided for this separation to ensure availability of the 2P-2A and/or 2P-2B in the event of a fire in the eastern portion of the room. It is also recognized that the fire would affect the speed control of 2P-2A and 2P-2B. For this reason it is important to minimize the interruption of charging flow to RCP seal cooling. If seal cooling would be lost for too long a duration, the resulting leakage could exceed the capability of charging pump(s) operating at minimum flow.
Through a physical walk down of cable routing, it was confirmed that at least one of the control power circuit cables for each of the Unit 2 'A' Train Charging Pumps is routed through the eastern portion of the fire area.
Fire damage to the control cables in this area could result in a total loss of the control circuit.
Neither FOP 1.2 nor the Safe Shutdown Analysis Compliance Strategies provide adequate guidance of manual actions necessary to restart these pumps as necessary to ensure adequate RCP cooling.
Safety Significance:
The plant maintains administrative controls for the introduction of combustible materials and ignition sources to limit the probability of a fire. These areas are also provided with automatic fire detection and suppression systems. The safety significance of this condition continues to be evaluated. Based on these administrative controls and fire protection features, NMC considers the safety significance of this issue to be very low.
NMC will provide a supplement to this LER if the result of the final evaluation indicates more than very low safety significance.
Point Beach Nuclear Plant Unit 1 05000266 i YEAR 1 SEQUENTIAL REVISION LER NUMBER (6 PAGE (3)
Cause:
The cause of this condition was a result of only considering the impact of control circuit damage on the speed control of an operating charging pump, and not the manual actions required to operate the pump should the pump be in standby or tripped due to under voltage or short circuit.
Corrective Action:
Completed:
- Implemented twice per shift compensatory fire watches for Fire Area A06 until resolution of manual actions is complete.
- Revised FOP 1.2 with selected method of recovery and remove option to isolate RCP seal cooling from FOP 1.2, Fire Area A-06.
- Implemented a Mode 4 hold during the Unit 2 outage until FOP 1.2 was revised.
- Revised FOP 1.2 with selected method of recovery and removed options to isolate seal cooling from FOP 1.2 for Fire Area A-15.
Planned:
- Evaluate potential manual actions to resolve this issue.
- Update the Safe Shutdown Analysis with selected method of recovery.
Previous Similar Events:
A review of recent LERs (past three years) identified one event that involved improper routing of electrical cables resulting in inadequate train separation.
LER Number Title 266-2002-002-00 Unit 1 A-Train Reactor Protection Cable Routed in B-Train Cable Trays
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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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