05000293/LER-1993-001, :on 930126,high Pressure Coolant Injection Sys Declared Inoperable.Caused by Blown Power Fuse Located within Flow Controller.Blown Fuse Replaced
| ML20058D453 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/24/1993 |
| From: | Boulette E, Gay R BOSTON EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BECO-LTR-93-147, LER-93-001, LER-93-1, NUDOCS 9312030158 | |
| Download: ML20058D453 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(x) |
| 2931993001R00 - NRC Website | |
text
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ff 10 CFR 50.73 BOSTON EDISON Pilg tm Nuclear Power Station Rocky Hill Road P!yrnouth, Massachusetts 02360 November 24, 1993 BECo Ltr. 93 147 E. T. Boulette, PhD senior Vice Presdent-Nuclear U.S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C. 20555 Docket No. 50-293 License No. DPR-35
Dear Sir:
The enclosed supplemental Licensee Event Report (LER) 93-001-01, "High Pressure Coolant Injection System Declared Inoperable Due to No Flow Indication During Surveillance", is submitted in accordance with 10 CFR Part 50.73.
Please do not hesitate to contact me if there are any questions regarding this report.
A D
C E. T. Boulette, PhD RAG /bal
Enclosure:
LER 93-001-01 cc:
Mr. Thomas T. Martin Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Rd.
King of Prussia, PA 19406 Mr. R. B. Eaton Div. of Reactor Projects I/II Office of NRR - USNRC One White Flint North - Mail Stop 14D1 11555 Rockville Pike Rockville, MD 20852 Sr. NRC Resident Inspector - Pilgrim Station i
Standard BECo LER Distribution 1
3000.'i 9312030158 931124
}ihI PDR ADDCK 05000293 S
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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROUED BY OMB NO. 3150-0104 e sm; -
EXPIRES 3/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY W.TH THfS
- LICENSEE EVENT REPORT (LER)
- l2"Z L ' %'&" "R 5 % J f0""llNJ,RIL's AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U S. NUCLEAR I
REGULATORY COMMISSION, WASHINGTON, DC 20555-0301, AND TO THE PAPERWORK REDUCTION PROJECT p1SO.0104), OFFICE i
(See reme tot number of dqts tcharacte+s for each bir,cN OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
1 FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
PILGRIM NUCLEAR POWER STATION 05000 -293 1 of5 TITLE (4)
High Pressure Coolant Injection Systern Declared Inoperable Due to No Flow Indication During Surveillance EVENT DATE (5)
LER NUMBER (6 i REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
SEQUENT:AL REASON F ACLU 1Y NAME DOCKET NUMBER MoNm oAv vEAR YEAR NUMBER NUMBER MONTH DAV YEAR N/A 05000 FACluTY NAME DOCKET NUMBER 01 26 93 93 001 01 11 24 93 N/A 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 6: (Check one or more)(11)
ODE (9)
N 20.402(b) 20.405(c) 50.73(a)(2)0v) 73 71(b) 20.405(aH1)0) 50.36ic)(1) 50.73(a)(2)(v) 73.71(c)
LEVE 10) 100 20 405(a)(110i) 50.36(c)(2) 50.73(a)(2)(vii)
X OTHER 20 405(a)(1)0ii) 50.73(aH2)C) 50.73(a)(2)(viii)(A) 20 405(a)(1Hiv) 50.73(aH210i) 50.73(aH2)(viii)(B) tgg 20.405(a)(1)h) 50 73(a)(2)(iii) 50.73(a)(2)(x)
Fgy,gAJ LICENSEE CONTACT FOR THIS LER (12) twAME TELEPHONE NUMBER {lmcluoe Area Cooet Robert A. Gay - Senior Compliance Engineer (508) 747-8047 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT MANUF ACTURER S
CAUSE
SYSEM COMPONENT MANUTACTURER D
SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED
'^#
YES NO SUBMISSION m yn compwe EXPECTED SUUM:SSON DATO y
DATE (15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewntten lines)(16)
On January 26, 1993, at 1004 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.82022e-4 months <br />, the High Pressure Coolant injection (HPCI) System was declared inoperable and a seven day Technical Specification (3.5.C.2) Limiting Condition for Operation was entered.
The HPCI System was declared inoperable because no flow was indicated at the HPCI pump flow indicating controller during a scheduled monthly operability surveillance test. All other parameters registered normal during the test.
The HPCI System was returned to normal standby service during subsequent investigative and corrective action activities, which identified a blown fuse within the HPCI flow controller.
An operability evaluation completed on October 4,1993, concluded the HPCI System was capable of its intended design function even with the fuse blown.
The cause was a blown power supply fuse located within the flow controller. The cause of the blown fuse could not be determined.
Corrective action taken included replacing the blown fuse.
The HPCI System was tested with satisfactory results and declared operable on January 26, 1993, at 1620 hours0.0188 days <br />0.45 hours <br />0.00268 weeks <br />6.1641e-4 months <br />. While the HPCI System was declared inoperable, the applicable systems were verified operable as required by Technical Specifications. An engineering evaluation of the blown fuse resulted in planned changes in the fuse rating and addition of alarms to the plant information computer.
The event occurred during power operation while at 100 percent reactor power.
The reactor mode selector switch was in the RUN position.
The Reactor Vessel (RV) pressure was 1025 psig with the RV water temperature at 545 degrees Fahrenheit.
This report is submitted voluntarily.
The event posed no threat to the public health and safety.
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j NRC FOBM 3MA U.S. NUCLEAR REGULATORY COMMISSION APPROtfED BY OMB NO. 3150-0104 a s_%;
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DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR B
B PILGRIM NUCLEAR POWER STATION 05000-293 U
93
- - 001 -
01 TEXT Of more space is reqwed, use 6dd!tional copies of NRC Form 3%A)(17)
BEASON FOR SUPPLEMENT This supplement is submitted voluntarily to reflect the completion of an operability evaluation and corrective actions as described in the initial report submitted on January i
26, 1993.
l t
An evaluation of the blown fuse relative to HPCI System operability was completed on October 4, 1993. The evaluation concluded the HPCI System was capable of providing its i
design function while the fuse was blown.
The HPCI System would have started automatically or manually upon receipt of an initiation signal or on demand and brought the system to its design flow rate within 90 seconds.
The blown fuse had/has no affect on the equipment or instruments necessary for the system to operate automatically or manually from the Control Room or Alternate Shutdown Panel, i
BACKGROUND i
Plant Design Change 83-11 replaced GEMAC transr.i+ters with Rosemount transmitters in September of 1983.
Field Revision Notice 83-11-;.3 was written to lower the protective fuse amperage rating to 1/20 amp. due to zener diodes added to power supply and controller output by PDC 83-11.
In March of 1984, FRN 83-11-30 was written to replace the fast blow r
fuses with slow blow fuses (Bussman MDI 1/32 amp.) due to fast blow fuses failing on
[
circuit high in-rush current.
Maintenance Request 89-23-93 was written in October of 1989 to replace a blown fuse (Bussman MDL 1/32 amp.) located inside Flow Indicating Controller FIC-2340-1. The Flow Indicating Controller unit FIC-2340-1 was replaced in November of 1990 via Maintenance Request 90-23-92.
Engineering personnel involved with the Rosemount transmitter installation indicate at the l
time the slow blow fuse selection was made, it was done on a conservative judgment basis i
that focused on a design requirement to protect zener diodes rated for 50 mA steady state cerrent.
Based on the availability of the fuse, a 32mA fuse was selected to protect the zener diode.
t The flow indicating controller responds to a pump flow demand signal and adjusts steam flow to accommodate varying reactor pressure. The controller can be operated in either manual or automatic modes.
The flow indicating controller is normally in AUTO with its l
i tape set at 4250 gpm.
The operators utilize the flow indication from FIC-2340-1 as part
)
of the pump and valve operability test to ensure the HPCI pump flow rate is in compliance with Technical Specifications.
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N4C FORM MA p 92) i
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104a s-en EXPIRES 5/31/95
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ESTIMATED BU5OEN PER RESPONSE TO COMPtY WrrH THIS
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FACILITY NAME (1)
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PAGE (3)
VEAR UMB R M
PILGRIM NUCLEAR POWER STATION 05000-293 93
--001 -
01 TEXT pf more space is required, use additional copies of NRC Form 366A)(17)
EVENT DESCRIPTION
On January 26,1993, at 1004 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.82022e-4 months <br />, the HPCI System was declared inoperable and a seven day Technical Specification (3.5.C.2) Limiting Condition for Operation (LCO) was entered.
The system was declared inoperable because no pump flow was indicated at the flow indicating controller, FIC-2340-1, during a scheduled monthly operability surveillance test conducted in accordance with Procedure 8.5 4.1 (Rev. 41), "HPCI System Pump and Valve Monthly / Quarterly Operability".
Even though no flow was detected on the flow controller, all other parameters registered as normal during the test. The HPCI System was returned to normal standby status for subsequent investigative and corrective action activities.
Following immediate investigation, Maintenance Request 19300305 was issued. The fuse (Bussman MDL 1/32 amp.) was replaced and the HPCI System was returned to standby service on January 26, 1993, at 1620 hours0.0188 days <br />0.45 hours <br />0.00268 weeks <br />6.1641e-4 months <br /> after satisfactory post work testing.
Problem report 93.9020 was written to document the event.
The NRC Operations Center was notified in accordance with 10 CFR 50.72 on January 26, 1993, at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />. The Reactor Core Isolation Cooling (RCIC) System was verified operable in accordance with Technical Specification 3.5.C.2.
The event occurred during power operation while at 100 percent reactor power with the reactor mode selector switch in the RUN position.
The Reactor Vessel (RV) pressure was 1025 psig with the RV water temperature at 545 degrees Fahrenheit.
CAUSE
The cause of the no flow indication at the flow indicating controller FIC-2340-1 was due to a blown power supply fuse (Bussman MDL 1/32 amp.) located within the flow controller that is part of the circuitry related to the HPCI pump flo;. transmitter FT-2358. The root cause of the blown fuse could not be determined, but potential factors such as normal operating characteristics and ambient temperature in which it functions possibly caused the fuse failure. The fuse either failed due to an overcurrent surge condition or from a premature manufacturing deficiency failure. The methodology used to perform calibration, surveillance, troanicshooting and the uniqueness of using the controller's internal power supply for instrw.:ent loop power versus the majority of other instrument loops, suggested current surges and possible elevated temperature of the fuse were the key factors contributing to the blown fuse.
The fuse manufacturer (Bussman Division - Cooper Industries) indicated the b!nwn fuse is typical of an overcurrent surge condition versus a dead short condition. The meutacturer noted the existing fuse is designed to handle two times rated current for about 12 seconds at 77 degrees Fahrenheit. The manufacturer indicated the fuse would degrade over time if surge currents in excess of specifications occurred.
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YEAR N BER PILGRIM NUCLEAR POWER STATION 05000-293 93
--001 -
01 TEMT (It more space is required, use aud:tional copies of NRC Form 366A)(17)
Surge currents in this circuit have been documented in excess of 50 mA, and the rating of the fuse in this applica. ion should be derated due to elevated ambient temperature.
Because the flow controller requires removal for calibration, it is evident surge currents are generated. The HPCI controller operates at full output and a higher temperature during standby service while the RCIC controller operates at zero output during standby service.
The Emergency and Plant Information Computer historical file for the HPCI pump flow transmitter FT-2358 data point HPC010 indicates the HPCI System would have exhibited no-indicated flow if initiated from January 15, 1993, to the time of fuse replacement on January 26, 1993.
This blown fuse did not inhibit automatic initiation or operation of the system but affects the ability to measure pump flow.
CORRECTIVE ACTION
Immediate corrective action taken included replacement of the blown fuse via Maintenance Raquest 19300305.
The HPCI System was subsequently tested in accordance with Procedure 8.5.4.1 (Rev. 41), "HPCI System Pump and Valve Monthly / Quarterly Operability", with satisfactory results.
The HPCI System was declared operable and the seven day LC0 was terminated on January 26, 1993, at 1620 hours0.0188 days <br />0.45 hours <br />0.00268 weeks <br />6.1641e-4 months <br />.
a Engineering has evaluated the fuse failure and provided recommendations to preclude recurrence.
The HPCI and RCIC flow measuring loops can fail without easily being detected if the instrument loop fuse fails.
This condition does not inhibit automatic initiation or operation of the system but affects the ability to measure pump flow.
Engineering has performed an evaluation to add the HPCI pump flow transmitter computer point (HPC010) and l,
RCIC pump flow transmitter competer point (RC1010) on the computer alarm typer in the Main Control Room. This will provide prompt identification of future fuse failures.
Field Revision Notice 93-04-22 was issued on October 21, 1993. This FRN will change the rating of the installed power supply fuse from a Bussman MDL 1/32 ampere to a Bussman MDL 1/16 ampere time-delay fuse, for HPCI flow indicating controllers FIC-2340-1 (Control Room) and -2 (Alternate Shutdown Panel) and the RCIC flow indicating controllers FIC-1340-l 1 (Control Room) and -2 (Alternate Shutdown Panel). The new fuse size should significantly reduce the likelihood of spurious fuse blowing without undue risk to other j
circuit components.
in addition, this FRN will add low s-ignal alarms to computer input points HPC010 and RCIO10 for the HPCI and RCIC pump flow transmitters, respectively. These alarms will appear on the control room point alarm log screen and will be printed on the control room
)
alarm typer.
Installation of the new fuse size (1/16 amp.) and software changes for computer input points HPC010 and RCIO10 will be conducted via Maintenance Requests 19303690 (HPCI) and 19303689 (RCIC).
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NRC FQRM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-oto4
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YEAR N P U H PILGRIM NUCLEAR POWER STATION 05000-293 93
--001 -
01 T EXT (tt more space as requned, use additional copies of NRC Form 356A)(17)
The Nuclear Plant Reliability Data System (NPRDS) was reviewed for trending of fuse failures. NPRDS does not trend the failure of fuses.
SAFETY CONSEOUENCES l
This event posed no threat to the public health and safety.
The blown fuse was not a component failure.
The Core Standby Cooling System (CSCS) consists of the HPCI System, Automatic Depressurization System (ADS), Core Spray System, and Residual Heat Removal System / Low Pressure Coolant Injection (LPCI) mode.
Although not part of the CSCS, the RCIC System is capable of providing water to the Reactor Vessel for core cooling, similar to the HPCI System. During the period the HPCI System was declared inoperable, the RCIC, ADS, Core Spray, and RHR System /LPCI mode were verified operable in accordance with Technical Specification 3.5.C.2.
An operability evaluation concluded the HPCI System was capable of performing its intended design function even with the fuse blown and therefore, the event is not reportable pursuant to 10 CFR 50.73. Consequently, this report is submitted voluntarily.
SIMILARITY TO PREVIOUS FVENTS A review was conducted of Pilgrim Station Licensee Event Reports (LERs) submitted since January 1984.
The review focused on LERs submitted in accordance with 10 CFR 50.73(a)(2)(v) that involved the HPCI System. This review did not identify any similar events.
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES The EIIS codes for this report are as follows:
COMPONENTS CODES Control, Indicating, Flow (FIC-2340-1)
FIC Fuse FU SYSTEMS High Pressure Coolant Injection (HPCI) System BJ NRC f ORM 3%A f5 92)