05000302/LER-2008-002, Emergency Feedwater Actuation on Low Steam Generator Level Due to Feedwater Pump Speed Tuning
| ML081220246 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/29/2008 |
| From: | Annacone M Progress Energy Florida |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 3F0408-09 LER-08-002-00 | |
| Download: ML081220246 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3022008002R00 - NRC Website | |
text
Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 50.73 April 29, 2008 3F0408-09 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
CRYSTAL RIVER UNIT 3 - LICENSEE EVENT REPORT 50-302/2008-002-00
Dear Sir:
Florida Power Corporation, currently doing business as Progress Energy Florida, Inc., hereby submits Licensee Event Report (LER) 50-302/2008-002-00. The LER discusses an Emergency Feedwater Initiation and Control System actuation on low Steam Generator level due to Main Feedwater pump oscillations caused by changes made to a calibration data sheet. This report is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A).
No new regulatory commitments are made in this letter.
If you have any questions regarding this submittal, please contact Mr. Dennis W. Herrin, Acting Supervisor, Licensing and Regulatory Programs at (352) 563-4633.
Sincer Mic I J. Annacone Plant General Manager Crystal River Nuclear Plant MJA/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
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE CRYSTAL RIVER UNIT 3 05000302 1 of 6
- 4. TITLE Emergency Feedwater Actuation On Low Steam Generator Level Due To Feedwater Pump Speed Tuning
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONT DAY YEASEQUENTIAL REV I MOH DFACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER REV MONTH DAY YEAR 05000 Si iFACILITY NAME DOCKET NUMBER 03 01 2008 2008 - 002 -
00 04 29 2008 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
[1 20.2201(b)
[I 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
E] 50.73(a)(2)(vii)
[ El 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[I 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
[1 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL [I 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
[
50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
[3 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
E] 73.71(a)(4)
El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5)
- - 20%
El 20.2203(a)(2)(v)
[I 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
[I OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Dennis W. Herrin, Lead Enginee r (Licensing and Regulatory Programs) 352-563-4633SYSTEM MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
COMPONENT FACTURER TO EPIX CUFACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION [E YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 9 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
At 04:27 on March 1, 2008, while operating in MODE 1 (POWER OPERATION) at approximately 20 percent RATED THERMAL POWER, Progress Energy Florida, Inc., Crystal River Unit 3 (CR-3) was being shut down to replace a reactor coolant pump seal when an Emergency Feedwater Initiation and Control (EFIC) System actuation occurred. Feedwater (FW) flow to both Once Through Steam Generators (OTSGs) began oscillating and an EFIC actuation was received on low level in both OTSGs as the Steam Generator/Reactor Demand was being lowered to zero in accordance with OP-209A, "Plant Shutdown and Cooldown." EFIC actuated as expected by automatically starting Emergency Feedwater pumps EFP-2 and EFP-3 and aligning Emergency Feedwater to both OTSGs. Required equipment operated as designed. The root cause for this event was a programmatic weakness in that the impact on other systems was not understood when the Feedwater pump FWP-2A Integrated Control System (ICS) Bailey module IC-5714-FW calibration data sheet was revised prior to Refueling Outage 15. ICS Bailey module IC-5714-FW was re-calibrated on March 21, 2008. This condition does not represent a reduction in the public health and safety. Loss of FW is an event analyzed and bounded by the CR-3 Final Safety Analysis Report accident analysis. A previous similar occurrence has not been reported to the NRC.
NRC FORM 366 (9-2007)
PRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL I REVISION YA NUMBER NUMBER CRYSTAL RIVER UNIT 3 05000-302 2008 002 00 2
OF 6
EVENT DISCRIPTION At 04:27 on March 1, 2008, while operating in MODE 1 (POWER OPERATION) at approximately 20 percent RATED THERMAL POWER, Progress Energy Florida, Inc. (PEF), Crystal River Unit 3 (CR-3) was being shut down to replace a Reactor Coolant System pump [AB, P] seal [AB, SEAL) when an Emergency Feedwater Initiation and Control (EFIC) System [JB] actuation occurred due a low level in both Once-Through Steam Generators (OTSGs) [AB, SG].
Main Feedwater System (FW) [SJ] flow to both OTSGs began oscillating and Integrated Control System (ICS) [JAI stations were taken to manual in an attempt to stabilize FW flow. An EFIC actuation occurred on low level in both OTSGs as the Steam Generator/Reactor Demand was being lowered to zero in accordance with Operating Procedure OP-209A, "Plant Shutdown and Cooldown." Steam Driven Emergency Feedwater Pump EFP-2 [BA, P] and Diesel Driven Emergency Feedwater Pump EFP-3 automatically started and were aligned to feed both OTSGs.
Computer data indicates Emergency Feedwater (EFW) [BA] flow to the OTSGs occurred for a short duration. The control stations for the EFW control valves to the "A" OTSG (EFV-56 and EFV-
- 58) [BA, FCV] were taken to manual and closed to secure EFW flow to the OTSGs since FW flow existed and was being controlled. FW flow to both OTSGs was quickly recovered and supplied inventory to the OTSGs with levels stabilized at the required low level limit setpoint.
At 05:27 on March 1, 2008, the Main Turbine [TA, TG] was tripped per OP-209A.
At 06:01 on March 1, 2008, CR-3 entered MODE 2.
At 06:19 on March 1, 2008, EFP-2 and EFP-3 were shutdown and both trains of EFIC were reset.
At 09:52 on March 1, 2008, an 8-hour notification to the NRC Operations Center was made in accordance with 1 OCFR50.72(b)(3)(iv)(A) for automatic actuation of the EFW System. This condition is also reportable as a 60-day Licensee Event Report under 10 CFR 50.73(a)(2)(iv)(A) for automatic actuation of the EFW System.
SAFETY CONSEQUENCES
EFIC performs the following safety functions : (1) initiates EFW when required by plant conditions; (2) controls EFW to establish/maintain required levels in the OTSGs; (3) controls the rate of OTSG level increase to minimize overcooling the primary coolant system; (4) isolates the Main Steam [SB] and FW lines of a depressurized OTSG; (5) selects and supplies EFW to the appropriate OTSG(s) in the event of a steam or feedwater line rupture; (6) terminates EFW to an OTSG that approaches an overfill condition; and, (7) controls the atmospheric dump valves [SB, RV] to maintain steam pressure at a predetermined setpoint.
Each of the four (4) EFIC channels receives analog inputs from dedicated level and pressure instruments from each OTSG. The instruments have no other plant control functions. Each EFIC channel will actuate if any of six (6) conditions are met. One of those conditions is a low level (-< 0 inches indicated) in either OTSG for >2 seconds.PRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR ISEQUENTIAL I REVISION YEAR NUMBER NUMBER CRYSTAL RIVER UNIT 3 05000-302 2008 002 00 3
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Upon receipt of a valid low level limit signal on both OTSGs, EFIC actuated as expected and initiated EFW. EFP-2 and EFP-3 automatically started and were aligned to feed both OTSGs.
Required equipment operated as designed.
This event did not result in the release of radioactive material. Design safety limits were not exceeded. No fission product barriers or components were damaged. Loss of FW is an event analyzed'and bounded by the CR-3 Final Safety Analysis Report accident analysis.
Based on the above discussion, PEF concludes that actuation of EFW did not represent a reduction in the public health and safety.
This event is not reportable under 10 CFR 50.73(a)(2)(v) and does not represent a condition that would have prevented the fulfillment of a safety function. Therefore, this event does not meet the Nuclear Energy Institute (NEI) definition of a Safety System Functional Failure (Reference: NEI 99-02, Revision 5).
CAUSE
The root cause for this event was a programmatic weakness in that the impact on other systems was not understood when the Feedwater pump FWP-2A [SJ, P] Integrated Control System (ICS)
Bailey module IC-5714-FW [JA, IMOD] calibration data sheet was revised prior to Refueling Outage 15. The calibration data sheet revision required electronic calibration of the proportional gain instead of a visual check of the proportional gain potentiometer dial position. The ICS Bailey module IC-5714-FW proportional gain potentiometer dial position of 0.2 does not correspond to an electronically calibrated proportional gain of 0.2.
The desired response from a properly tuned system is that pump speed will oscillate in reducing magnitude levels while recovering after a perturbation. With an improperly tuned system the pump speed will oscillate following a perturbation, but the magnitude of each oscillation grows larger.
The latter occurred on March 1, 2008.
FWP-2A and FWP-2B each have an analog speed governor (Woodward Model 2301) [SJ, 65] that receives an input from an ICS Bailey module (IC-5714-FW for FWP-2A and IC-5814-FW for FWP-2B). Both the Woodward governor and the ICS Bailey module have proportional and integral gain settings that can be adjusted during tuning activities. The ICS system has two FWP control modes, flow control above 50% power and differential pressure control below 50% power. Both of these modes use different modules that are switched at approximately 50% power when the main block valve [SJ, FVC] cycles open or closed. If either the ICS Bailey module or the Woodward governor tuning is inadequate, there is the possibility for pump speed instability.
FWP-2A ICS Bailey module IC-5714-FW is calibrated in the Preventive Maintenance Program (PMID 00029642-01) on a 4 year frequency. This module was calibrated in November 2007 during Refueling Outage 15 (R15) under Work Order 801516 using a calibration data sheet that had been revised shortly after R13 to include an electronic proportional gain check due to module calibration challenges. R1 5 was the first time that the revised calibration data sheet was used.PRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE I I SEQUENTIAL I REVISION YEAR NUMBER NUMBER CRYSTAL RIVER UNIT 3 05000-302 2008 002 00 4
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In the past, the calibration data sheet required only a visual check of the proportional gain potentiometer dial position. The revised calibration data sheet has the proportional gain electronically calibrated by using the voltage at pin 7. For a gain setting of 0.2 specified in the calibration data sheet, an output voltage of -2.0 volts direct current (vdc) is expected at pin 7 with an input voltage of 10 vdc at pin 11. The as-found output voltage in R1 5 was -0.415 vdc. The proportional gain potentiometer dial position was adjusted to reach an output voltage of -2.002 vdc to satisfy the value specified in the calibration data sheet.
The proportional gain setting was increased enough to cause adverse FWP-2A response speed stability. Following this R1 5 adjustment, FWP-2A was not stable in automatic and tuning activities were implemented which focused on the Woodward governor adjustments. The post-R15 tuning activities were adequate to achieve automatic FWP-2A control at higher flow rates but not adequate to assure stable operation for lower flow rates.
Corrective Actions
- 1.
The proportional gain setting for IC-5714-FW was adjusted in a direction that set it closer to the pre-R15 value which greatly improved FWP-2A stability and the ability to tune the Woodward governor. This was accomplished in accordance with Work Order 1305996 on March 21, 2008.
- 2.
Additional corrective actions are identified in Nuclear Condition Report 268400.
PREVIOUS SIMILAR EVENTS
Valid actuations of EFIC have occurred in the past at CR-3. However, none of those actuations occurred as the result of FWP speed tuning activities or incorrect information in the calibration data sheet for an ICS Bailey module.
ATTACHMENTS - Abbreviations, Definitions, and Acronyms - List of CommitmentsPRINTED ON RECYCLED PAPERPRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL I REVISION YEAR NUMBER NUMBER CRYSTAL RIVER UNIT 3 105000-302 2008 002 00 5
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Abbreviations, Definitions, and Acronyms CFR CR-3 EFIC EFP EFV EFW FW FWP FWV ICS NEI NRC OP OTSG PEF R15 vdc NOTES:
{e.g., MODE 1 Code of Federal Regulations Crystal River Unit 3 Emergency Feedwater Initiation and Control System Emergency Feedwater Pump Emergency Feedwater Valve Emergency Feedwater System Main Feedwater System Main Feedwater Pump Main Feedwater Valve Integrated Control System Nuclear Energy Institute Nuclear Regulatory Commission Operating Procedure Once-Through Steam Generator Progress Energy Florida, Inc.
Refueling Outage 15 Volts Direct Current Improved Technical Specification Defined terms appear capitalized in LER text I.
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]}PRINTED ON RECYCLED PAPERPRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME-
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUEENTIAL RVSO NUMBER INUMBER CRYSTAL RIVER UNIT 3 05000-302 2008 002 00 6
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LIST OF COMMITMENTS The following table identifies those actions committed 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 Acting Supervisor, Licensing and Regulatory Programs of any questions regarding this document or any associated regulatory commitments.PRINTED ON RECYCLED PAPERPRINTED ON RECYCLED PAPER