05000302/LER-2004-001, Regarding Actuation of the Reactor Protection System and Emergency Feedwater System Caused by a Failed Circuit Board within the Main Feedwater Integrated Control System on March 24, 2004
| ML041470297 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/18/2004 |
| From: | Franke J Progress Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 3F0504-02 LER 04-001-00 | |
| Download: ML041470297 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3022004001R00 - NRC Website | |
text
as Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 50.73 May 18, 2004 3F0504-02 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
LICENSEE EVENT REPORT 50-302/04-001-00
Dear Sir:
Please find enclosed Licensee Event Report (LER) 50-302/04-001-00.
The LER discusses actuation of the Reactor Protection System and Emergency Feedwater System caused by a failed circuit board within the Main Feedwater Integrated Control System on March 24, 2004.
This report is being submitted pursuant to 1 OCFR50.73(a)(2)(iv)(A).
No new regulatory commitments are made in this letter.
If you have any questions regarding this submittal, please contact Mr. Sid Powell, Supervisor, Licensing and Regulatory Programs at (352) 563-4883.
Sil Jon) Franke Pijt General Manager 0 rystal River Nuclear Plant JAF/dwh Enclosure xc:
Regional Administrator, Region II Senior Resident Inspector NRR Project Manager Progress Energy Florida, Inc.
Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428
, e. 2-
Abstract
At 03:31, on March 24, 2004, Progress Energy Florida, Inc., (PEF) Crystal River Unit 3 (CR-3) was in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER. While in steady state operation, a large decrease in the reactor demand and Main Feedwater flow demand signals occurred within the Integrated Control System (ICS). The Main Feedwater pumps ran back until discharge pressure was less than steam generator pressure, at which time the Main Feedwater flow reduced to zero. Upon sensing less than 17 percent Main Feedwater flow, the Anticipated Transient Without Scram (ATWS) Mitigation System Actuation Circuitry (AMSAC) initiated a turbine trip and actuated the Emergency Feedwater System (EFW). The Reactor Protection System (RPS) sensed the turbine trip above 45 percent reactor power and initiated an anticipatory reactor trip. The cause for this event was age-related failure of the zener diodes in the +15 volt regulator circuit for a Bailey 820 Control Module in the Main Feedwater Integrated Control System.
The circuit board was replaced. RPS and EFW valid actuations are reportable under 10CFR50.73(a)(2)(iv)(A). 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.
NRC FORM 366 17-2001)U.S. NUCLEAR REGULATORY COMMISSION 1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE CRYSTAL RIVER UNIT 3 05000302 YEAR U
NUENTAL REVSON OF6 04
- - 001 00
- 17. TEXT (If more space is required, use additional copies of NRC Form 366A)
EVENT DESCRIPTION
At 03:31, on March 24, 2004, Progress Energy Florida, Inc., (PEF) Crystal River Unit 3 (CR-3) was in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER. While in steady state operation, a large decrease in the reactor demand and Main Feedwater [SJ] flow demand signals occurred within the Integrated Control System (ICS) [JA]. As expected, Main Feedwater flow decreased faster than reactor power. The Main Feedwater pumps [SJ, P] ran back until discharge pressure was less than steam generator [AB, SG] pressure, at which time the Main Feedwater flow reduced to zero percent. Upon sensing less than 17 percent Main Feedwater flow, the Anticipated Transient Without Scram (ATWS) Mitigation System Actuation Circuitry (AMSAC) [JE] initiated a turbine trip and actuated the Emergency Feedwater System (EFW) [BA]. The Reactor Protection System (RPS) [JC] sensed the turbine trip above 45 percent reactor power and initiated an anticipatory reactor trip.
No structures, systems or components were inoperable at the start of the event that contributed to the event. Plant safety systems responded as expected during the reactor trip with the following exceptions:
Auxiliary Steam Valve ASV-26 [SA, FCV] did not control auxiliary steam pressure in automatic and was placed in manual. The low Auxiliary Steam demand during this transient is outside the normal control range for ASV-26.
Heater Drain Valve HDV-83 [SJ, V] did not adequately control the Deaerator Feed Tank (DFT) [SJ, DEA] level. This may have been caused by mis-calibration of level switch FW LS [SJ, LS]. Condensate pumps CDP-1A and CDP-1 B [SD, P] tripped on high DFT level.
CDP-1 B was restarted.
RPS and EFW valid actuations are reportable to the NRC. At 04:39, on March 24, 2004, a non-emergency four-hour notification and a non-emergency eight-hour notification were made to the NRC Operations Center (Event Number 40608) in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A), respectively. This report is being submitted pursuant to 1 OCFR50.73(a)(2)(iv)(A).
SAFETY CONSEQUENCES
Based on the loss of Main Feedwater, valid actuation of AMSAC, EFW and RPS occurred as expected to shut down the reactor and maintain adequate steam generator levels. When the Main Feedwater flow decreased, AMSAC initiated a turbine trip and actuated EFW upon sensing less than 17 percent Main Feedwater flow. The RPS sensed the turbine trip above 45 percent reactor power and initiated a reactor trip. Reactor operators properly executed the Emergency Operating Procedures for plant shutdown.
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run I.o 1-:
IJU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE I
SEQUENTIAL I REVISION CRYSTAL RIVER UNIT 3 05000302 YEAR NUMBER i NUMBER 3 OF 6 04 001-00
- 17. TEXT (If more space is required, use additional copies of NRC Form 366AJ Based on the above discussion, PEF concludes that actuation of the RPS and EFW 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 age-related equipment failure. Module 3-8-4 [JA, IMOD] in the Main Feedwater ICS is a Bailey 820 Control Module. The +15 volt regulator circuit on the Module 3-8-4 circuit board contained four failed devices. The five elements of this circuit are three 5 volt zener diodes connected in series to provide a regulated 15 volt supply, a capacitor to reduce ripple, and a current limiting resistor. The first failure was one of the 5 volt zener diodes. This failed to a shorted state. This failure forced each of the other devices to increase its load current, dissipating more heat. The failure of this first device may have occurred months before the next step occurred. The increased heat dissipation led to subsequent failure of the remaining circuit components. The cause for this event was failure of the first 5 volt zener diode due to end of life.
The Progress Generation Nuclear Electronic Service Center recommends replacing zener diodes, carbon resistors and operating-amplifiers when service has reached 25 to 30 years. This is in line with the guidance from Electric Power Research Institute (EPRI) Technical Report 1007916, "Printed Circuit Board Maintenance, Repair and Testing Guide."
CORRECTIVE ACTIONS
- 1.
CR-3 Administrative Instruction Al-704, "Reactor Trip Review and Analysis," was performed.
- 2.
ICS Module 3-8-4 was replaced with a bench-tested, calibrated module.
- 3.
Other suspected ICS modules were pulled, inspected for visible damage and calibration checked on the test rack. No anomalies were identified.
- 4.
Other actions associated with this event are being addressed in the CR-3 Corrective Action Program in Nuclear Condition Report 122486.
PREVIOUS SIMILAR EVENTS
No previous similar events involving failure of the Bailey 820 Multiplier Control Module 15 volt regulator circuit have been reported to the NRC by CR-3.
NHRC FHM Jb3A (1-0UU1)U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE CRYSTAL RIVER UNIT 3 05000302 YEAR NQUMNAL REVSON 4OF6 04 001 00
- 17. TEXT (If more space is required, use additional copies of NRC Form 366A)
ATTACHMENTS - Abbreviations, Definitions, and Acronyms - List of Commitments NHU FURM 366A (1-2001)U.S. NUCLEAR REGULATORY COMMISSION (1 -2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE CRYSTAL RIVER UNIT 3 l
05000302 YEAR NUMBER I NUMRERS5O F6 l
04 001-00
- 17. TEXT (If more space is required, use additional copies of NRC Form 366AJ ATTACHMENT 1 ABBREVIATIONS, DEFINITIONS AND ACRONYMS Al Administrative Instruction ATWS Anticipated Transient Without Scram AMSAC ATWS Mitigation System Actuation Circuitry ASV Auxiliary Steam Valve CDP Condensate Pump CFR Code of Federal Regulations CR-3 Crystal River Unit 3 DFT Deaerator Feed Tank EFW Emergency Feedwater EPRI Electric Power Research Institute HDV Heater Drain Valve ICS Integrated Control System NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission PEF Progress Energy Florida, Inc.
RPS Reactor Protection System NOTES:
Improved Technical Specifications defined terms appear capitalized in LER text
{e.g., MODE 1}
Defined terms/acronyms/abbreviations appear in parenthesis when first used {e.g.,
Reactor Building (RB)).
EIIS codes appear in square brackets {e.g., reactor building penetration [NH, PEN]).U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE ISEQUENTIAL IREVISION CRYSTAL RIVER UNIT 3 05000302 YEAR NUMBE NUMBER 6 OF 6 04
- - 001-00
- 17. TEXT (If more space is required, use additional copies of NRC Form 366A)
ATTACHMENT 2 LIST OF COMMITMENTS The following table identifies those actions committed to by PEF in this document. Any other actions discussed in the submittal represent intended or planned actions by PEF. They are described to the NRC for the NRC's information and are not regulatory commitments. Please 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.
NKU f-OHM.366A (1-2001)