01-08-2007 | On November 8, 2006, at 2254 hours0.0261 days <br />0.626 hours <br />0.00373 weeks <br />8.57647e-4 months <br /> ( CST), with Unit 2 at approximately 85 percent power and Unit 3 in a refuel outage, Dresden Nuclear Power Station control room personnel identified that the Control Room Emergency Ventilation Air Conditioning System failed to maintain proper control room temperature.
Technical Specification 3.7.5, "Control Room Emergency Ventilation Air Conditioning (AC) System," was entered. The system was restored to operable status on November 9, 2006 at 0555 hours0.00642 days <br />0.154 hours <br />9.176587e-4 weeks <br />2.111775e-4 months <br /> (CST). This event is being reported in accordance with 10 CFR 50.73(a)(2)(v)(D), "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident," as the Control Room Emergency Ventilation Air Conditioning System is a single train system.
The cause of the event was attributed to a loss of process fluid from the Control Room Emergency Ventilation Air Conditioning System's Refrigeration Condensing Unit due to loose instrument tube fittings caused by vibration. The corrective action to address this event is to enhance existing Control Room Emergency Ventilation Air Conditioning System inspections by establishing a new maintenance activity to perform a yearly walk down to monitor for refrigerant leaks on the Control Room Emergency Ventilation Air Conditioning System. |
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LER-2006-005, Units 2 and 3 Control Room Emergency Ventilation Air Conditioning System Inoperable Due To Leaking FittingsDocket Number |
Event date: |
11-08-2006 |
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Report date: |
01-08-2007 |
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Reporting criterion: |
10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
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2372006005R00 - NRC Website |
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Dresden Nuclear Power Station (DNPS) Units 2 and 3 are General Electric Company Boiling Water Reactors with a licensed maximum power level of 2957 megawatts thermal. The Energy Industry Identification System codes used in the text are identified as [XX].
A. Plant Conditions Prior to Event:
Unit: 02 Event Date: 11-8-2006 Reactor Mode: 1 Mode Name: Power Operation Power Level: 85 percent Reactor Coolant System Pressure: 1000 psig
B. Description of Event:
On November 8, 2006, at 2254 hours0.0261 days <br />0.626 hours <br />0.00373 weeks <br />8.57647e-4 months <br /> (CST), with Unit 2 at approximately 85 percent power and Unit 3 in a refuel outage, DNPS control room personnel identified that the Control Room Emergency Ventilation Air Conditioning System (CREVS) [ACU] failed to maintain proper control room temperature. Technical Specification 3.7.5, "Control Room Emergency Ventilation Air Conditioning (AC) System," was entered. The system was restored to operable status on November 9, 2006 at 0555 hours0.00642 days <br />0.154 hours <br />9.176587e-4 weeks <br />2.111775e-4 months <br /> (CST).
An Emergency Notification System (ENS) call was made on November 9, 2006, at 0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> (CST) for the above-described event. The assigned ENS event number was 42976.
This event is being reported in accordance with 10 CFR 50.73(a)(2)(v)(D), "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident," as the CREVS is a single train system.
C. Cause of Event:
The cause of the event was attributed to a loss of process fluid from the CREVS's Refrigeration Condensing Unit (RCU) due to vibration causing the instrument tube fittings to become loose.
The Control Room Heating, Ventilation and Air Conditioning (HVAC) System is designed with two (2) HVAC Systems. One system is a non-safety related "A" Train System with an air handling unit, 2/3 5731, a return air fan, 2/3-5728, and two (2) 50% RCUs. The second of the two HVAC systems is the safety related "B" Train HVAC System (i.e. CREVS) comprised of an air handling unit, 2/3-9400-100, an air filtration unit, 2/3-9400-101, and a RCU, 2/3-9400-102. This system is designed in accordance with General Design Criterion 19, "Control Room Habitability," to maintain the control room habitability in the event of a Design Basis Accident. The system is designed to maintain the Control Room between 70 and 80 °F.
On November 8, 2006, at approximately 2145 hours0.0248 days <br />0.596 hours <br />0.00355 weeks <br />8.161725e-4 months <br />, CREVS was operating and the control room area temperature had increased to 78 °F, continuing to rise slowly. Operations personnel investigated the event and discovered the CREVS RCU compressor was short cycling rapidly. The compressor would stop momentarily and start again within a two-minute period. During the evaluation of the compressor, operations personnel identified that a small leak at the compressor low-pressure switch was spraying refrigerant and oil at the base of the RCU skid. The instrument tubing is not rigidly supported and has a tendency to transmit the motor/compressor vibrations along the tubing to the various CREVS instruments and equipment. The vibrations caused the instrument tubing fittings to loosen and resulted in the Freon leaks. The cause of the high control room temperature was the failure of the CREVS RCU to provide adequate cooling because of excessive leaking of oil and refrigerant.
The CREVS was declared inoperable and TS 3.7.5 Actions A.1, C.1 and C.2 were entered. The CREVS RCU was repaired and leak checked during RCU operation. On November 9, 2006 at 0555, CREVS was declared operable and TS 3.7.5 Actions A.1, C.1 and C.2 were exited.
A review of CREVS operating history did not identify any documented leaks of Freon from the system since installation in 1984. An Apparent Cause Evaluation (ACE) of this event identified three possible actions to address the CREVS inoperability: (1) eliminate the tubing compression fitting, (2) installation of tubing supports, or (3) a non-hardware solution.
The ACE evaluated the tubing compression fittings to determine if they were being used in the correct application. An industry search for compression fitting failures identified that there were fewer compression fitting failure events in this type of application than other connection options (i.e., rigid solder connections).
The ACE also evaluated the installation of instrument-tubing supports for the various tubing configurations. This would entail the removal of the current tubing to configure the tubing in a uniform layout through out the CREVS RCU compressor skid. The RCU skid is currently extremely compressed with the instruments and equipment installed on the skid. The addition of these supports would increase the complexity of the layout of skid components.
The ACE evaluated actions to mitigate the recurrence of this event without hardware changes. A review of the Exelon Reciprocating Chiller PCM Template guidance identified a recommendation to conduct periodic refrigerant leak walk downs. This guidance was originally not implemented because no documented history of CREVS refrigerant leaks existed at DNPS. The implementation of the PCM Template refrigerant leak walk downs guidance would have likely identified the leaking fittings prior to CREVS inoperability.
The ACE concluded that the best corrective action to address this event is to enhance existing CREVS inspections by establishing a new maintenance activity to perform a yearly walk down of the CREVS to monitor for refrigerant leaks.
D.� Safety Analysis:
The safety significance of the event is minimal. During this event, DNPS continued to operate within the requirements of the TS. Additionally, the CREVS during this event maintained the capability to isolate and pressurize the control room envelope in the event of a postulated radiological or toxic gas event. Therefore, the consequences of this event had minimal impact on the health and safety of the public and reactor safety.
E. Corrective Actions:
The CREVS RCU was repaired and leak checked during RCU operation. The system was restored to operable status on November 9, 2006 at 0555 hours0.00642 days <br />0.154 hours <br />9.176587e-4 weeks <br />2.111775e-4 months <br /> (CST).
DNPS will enhance existing CREVS inspections by establishing a new maintenance activity to perform a yearly walk down of the CREVS to monitor for refrigerant leaks.
F. Previous Occurrences:
A review of DNPS Licensee Event Reports (LERs) for the last three years did not identified any LERs associated with CREVS inoperability due to Freon leaks.
G. Component Failure Data:
NA
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05000305/LER-2006-010 | | | 05000456/LER-2006-001 | Unit 1 Reactor Coolant System Pressure Boundary Leakage Due To Inter-Granular Stress Corrosion Cracking of a Pressurizer Heater Sleeve | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000454/LER-2006-001 | Technical Specification Required Action Completion Time Exceeded for Inoperable Containment Isolation Valves Due to Untimely Operability Determination | | 05000423/LER-2006-001 | Loss Of Safety Function Of The Control Room Emergency Ventilation System | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000369/LER-2006-001 | Ice Condenser and Floor Cooling System Containment Isolation Valve inoperable longer than allowed by Technical Specification 3.6.3. | | 05000353/LER-2006-001 | HPCI Ramp Generator Signal Converter 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)(v), Loss of Safety Function | 05000352/LER-2006-001 | Loss Of One Offsite Circuit Due To Invalid Actuation Of Fire Suppression System | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000336/LER-2006-001 | | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor | 05000316/LER-2006-001 | Failure to Comply with Technical Specification 3.6.2, Containment Air Locks | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000315/LER-2006-001 | Plant Shutdown Required by Technical Specification Action 3.6.5.B.1 | | 05000293/LER-2006-001 | | 10 CFR 50.73(a)(2)(iv), System Actuation | 05000289/LER-2006-001 | | | 05000287/LER-2006-001 | Actuation of Emergency Generator due to Spurious Transformer Lockout | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000251/LER-2006-001 | Turkey Point Unit 4 05000251 1 OF 6 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000247/LER-2006-001 | Manual Reactor Trip Due to Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to Personnel Error | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000440/LER-2006-001 | Incorrect Wiring in the Remote Shutdown Panel Results in a Fire Protection Program Violation | | 05000413/LER-2006-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000368/LER-2006-001 | Completion of a Plant Shutdown Required by Technical Specifications Due to Loss of Motive Power to Certain Containment Isolation Valves as a Result of a Phase to Ground Short Circuit in a Motor Control Cubicle | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000306/LER-2006-001 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000298/LER-2006-001 | Cooper Nuclear Station 05000298 1 of 4 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000286/LER-2006-001 | I | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000282/LER-2006-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000266/LER-2006-001 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000261/LER-2006-001 | Manual Reactor Trip Due to Failure of a Turbine Governor Valve Electro-Hydraulic Control Card | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000255/LER-2006-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000461/LER-2006-002 | Turbine Bypass Function Lost Due to Circuit Card Maintenance Frequency | | 05000458/LER-2006-002 | Loss of Safety Function of High Pressure Core Spray Due to Manual Deactivation | | 05000456/LER-2006-002 | Units 1 and 2 Entry into Limiting Condition for Operation 3.0.3 due to Main Control Room Ventilation Envelope Low Pressure | | 05000443/LER-2006-002 | Noncompliance with the Requirements of Technical Specification 6.8.1.2.a | | 05000387/LER-2006-002 | DMissed Technical Specification surveillance requirement | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000362/LER-2006-002 | Unit 3 Shutdown to Inspect Safety Injection Tank Spiral Wound Gaskets | | 05000336/LER-2006-002 | Manual Reactor Trip Due To Trip Of Both Feed Pumps Following A Loss Of Instrument Air | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000316/LER-2006-002 | | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000315/LER-2006-002 | Failure to Comply with Technical Specification Requirement 3.6.13, Divider Barrier Integrity | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000293/LER-2006-002 | | | 05000289/LER-2006-002 | | | 05000251/LER-2006-002 | Intermediate Range High Flux Trip Setpoint Exceeded Technical Specification Allowable Value | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000440/LER-2006-002 | Scaffold Built in the Containment Pool Swell Region | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000413/LER-2006-002 | Safe Shutdown Potentially Challenged by an External Flooding Event and Inadequate Design and Configuration Control | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000388/LER-2006-002 | Missed Technical Specification LCO 3.8.1 Entry for Unit 2 During Unit 1 ESS Bus Testing | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000348/LER-2006-002 | Main Steam Isolation Valve Failure to Close | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000305/LER-2006-002 | | 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 | 05000301/LER-2006-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2006-002 | 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249Entergy Tel (914) 734-6700 Fred Dacimo Site Vice President Administration September 13, 2006 Indian Point Unit No. 3 Docket No. 50-286 N L-06-084 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001 Subject:L Licensee Event Report # 2006-002-00, "Manual Reactor Trip as a Result of Arcing Under the Main Generator Between Scaffolding and Phase A&B of the Isophase Bus Housing" Dear Sir: The attached Licensee Event Report (LER) 2006-002-00 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-IP3-2006-02255. 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. Fred R. Dacimo Site Vice President Indian Point Energy Center Docket No. 50-286 NL-06-084 Page 2 of 2 Attachment: LER-2006-002-00 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 366 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.DReported 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, Washington, DC 20555-0001, or by internetLICENSEE EVENT REPORT (LER) e-mail to infocollects@nrc.gov, 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 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. FACILITY NAME 2. DOCKET NUMBER I 3. PAGE
INDIAN POINT 3 05000-286 1 OF 6
4.TITLE: Manual Reactor Trip as a Result of Arcing Under the Main Generator Between
Scaffolding and Phase A&B of the Iso-phase Bus Housing | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000282/LER-2006-002 | | 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-2006-002 | High Energy Line Breaks Outside Licensing Basis May Result in Loss of Safety Function | | 05000263/LER-2006-002 | | | 05000255/LER-2006-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000254/LER-2006-002 | Quad Cities Nuclear Power Station Unit 1 05000254 1 of 3 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000483/LER-2006-003 | Unexpected Inoperability of the Emergency Exhaust System due to Inoperable Pressure Boundary | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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