05000293/LER-1994-001, :on 940104,HPCIS Became Inoperable & Seven Day TS 3.5.C.2 LCO Entered Due to Unplanned Isolation During Surveillance Testing.Caused by False Low Rv Pressure Signal. Calibr Unit Replaced & Hpcis Declared Operable
| ML20063E480 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/03/1994 |
| From: | Boulette E, Ellis D BOSTON EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 94-014, 94-14, LER-94-001, LER-94-1, NUDOCS 9402100028 | |
| Download: ML20063E480 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) |
| 2931994001R00 - NRC Website | |
text
gOSTON EDISON Pdgrim Ncclear Power Stanon Rocky Hill Road Plymouth, Massachusetts 02360 E. T. Boulette. PhD senior vice Pressfent-Nuclear February 3,1994 BECo Ltr. 94- 014 U.S. Nuclear Regulatory Commission
. Attn: Document Control Desk Washington, D.C.
20555 P
Docket No. 50-293 License No. DPR-35 The enclosed Licensee Event Report (LER) 94-001-00, "High Pressure Coolant Injection System Inoperable Due to Unplanned Isolation During Surveillance Testing", 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, b Y DLL
(
E.T. Boulette, PhD DWE/bal/94011
Enclosure:
LER 94-001-00 cc:
Mr. Thomas T. Martin Regional Administrator, Region I U.S. Nuclear Regulatory Commicsion 475 Allendale Rd.
King of Prussia, PA 19406 l
Sr. NRC Resident Inspector - Pilgrim Station l
Standard BECo LER Distribution L
I 080073 MR2tggg ;gggs j S
NRC FORM 356 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 a.en EXPlRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THtS E
,e LICENSEE EVENT REPORT (LER) lL""'2"EG #Su"RE s" LEO TLdRMA3 R
AND RECORDS MAN AGEMENT BRANCH (MNB9 7714). U,S. NUCLEAR REGLA.ATORY COMM4SSION, WASHINGTON, DC 205554001, AND TO THE PAPERWORK REDUCTON PROJECT {31504104), OFFICE (See reverse tot number of digi 1m/charactem for each tdock)
OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
{
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
PILGRIM NUCLEAR POWER STATION 05000 -293 1 of 5 TITLE (4)
High Pressure Coolant injection System inoperable Due to Unplanned Isolation During Surveillance Testing EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8) i SEQUENTW.
REVSON FACIUTY NAME DOCKET NUMBER MONN DAY YEAR YEAR NUMBER NUMBER MONTH DAY YEAR N/A 05000 FACILfrY NAME DOCKET NUMBER 01 04 94 94 001 00 02 03 94 N/A 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 6: (Check one or more)(11)
MODE M N
l 20.402(b) 20.405(c)
X 50.73(a)(2)(iv) 73.71(b) 20MaM 50mm X m aH2 m 73m fLEJE 10) 100 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii)
OTHER 20.405(a)(1)(iii) 50.731a)(2)(i) 50.73(a)(2)(viii)(A) 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) gag g i
20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)
Form 366Al LICENSEE CONTACT FOR THIS LER (12)
TotEPHcd nuuBER tinciuoe A,e. Coo.>
NAME Douglas W. Ellis. Senior Cornpliance Engineer (508) 830-8160 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT MANUFACTURER
CAUSE
SYSTEM COMPONENT MANUFACTURER F
=
SUPPLEMENTAL REPORT EXPECTED (14)
M Nm DAY YEAR EXPECTED YES NO SUBMtSSION (K yes. complete EXPEC J SUBMISStON DATE)
X DATE (15)
ABSTRACT (Limit to Spaces, i.e., approximately 15 single-spaced typewntten lines)(16)
On January 4, 094, at 1834 hours0.0212 days <br />0.509 hours <br />0.00303 weeks <br />6.97837e-4 months <br />, the High Pressure Coolant Injection (HPCI) System became inoperable and a seven day Technical Specification (3.5.C.2) Limiting Condition for-Operation (LCO) was entered. The HPCI System became inoperable because the HPCI turbine steam supply isolation valves closed during a scheduled calibration test of related Analog Trip System (ATS) units.
The HPCI System was declared operable and returned to normal standby service at 0326 hours0.00377 days <br />0.0906 hours <br />5.390212e-4 weeks <br />1.24043e-4 months <br /> on January 5, 1994. Applicable systems were verified operable during the period the HPCI System was inoperable.
_j The cause was a false low Reactor Vessel pressure signal. The cause of the signal was l
investigated but coulu not be determined with certainty. The investigation concluded the most probable cause was human performance.
Corrective action taken included feedback to i
Instrumentation & Control (I&C) technicians and supervisors regarding this event.
Corrective action planned includes procedure improvement and I&C technician training.
1 The event occurred while at 100 percent reactor power. The reactor mode selector switch was in the RUN position.
The Reactor Vessel (RV) pressure was 1029 psig with the RV water temperature at 545 degrees Fahrenheit. The event posed no threat to the public health and safety.
NRO F0FOA 366A (5-92)U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 a s-m EXPlRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTTH THIS LICENSEE EVENT REPORT (LER)
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TEXT CONTINUATION E7AWC DC W SS G
TO THE PAPERWORK REDUCTION PROJECT p1504104L OFFICE OF MANAGEMENT AND SUDGET, WASHINGTON, DC 70503.
FACILITY NAME (1)
DOCKcT NUMBER (2)
LER NUMBER (6)
PAGE (3)
W YEAR B
PILGRIM NUCLEAR POWER STATION 05000-293 U
94
-- 001 -
00 TEXY pt more space is required, use additional copies of NRC Form 366AJ (17)
BACKGROUND The Reactor Vessel (RV) instrumentation includes RV water level, RV coolant flow rates and differential pressure, and RV internal pressure.
The Drywell instrumentation includes Drywell pressure.
By design, this safety-related instrumentation is arranged into separate, redundant Channels 'A' and
'B'.
Each instrument channel is connected to piping extending into primary containment. The Analog Trip System (ATS) is part of the instrumentation monitoring RV and Drywell parameters.
The ATS consists of transmitters, master trip units, slave trip units, and trip relays. The transmitters are housed on l
instrument racks located outside primary containment. The transmitters convert the parameter being monitored into signals.
The signals are converted into indications and/or trip. functions to related systems. The systems include the Reactor Protection System (RPS), Anticipated Transient Without Scram (ATWS) System, Primary Containment Isolation Control System (PCIS)/ Reactor Building Isolation Control System (RBIS), Core Standby Cooling Systems (CSCS), and Reactor Core Isolation Cooling (RCIC) System.
R The ATS cabinets contain the master trip units, slave trip units, trip relays, and power supplies. Typically, each transmitter provides signals to a master trip unit.
The signal is proportional to the parameter being monitored.
The master trip unit converts the signals into an indication and/or trip function.
Slave trip units provide functions similar to the master trip units.
Slave trip units receive signals from the transmitter via the related master trip unit.
The setpoints of the master trip units and slave trip units are individually calibrated.
The CSCS includes the High Pressure Coolant Injection (HPCI) System that is designed to automatically initiate if a low-low RV water level condition or high Drywell pressure condition occurs.
The system is also designed to automatically isolate if certain conditions occur, including low RV pressure. The low RV pressure conditior, is signalled to the HPCI System control circuitry via channel
'A' relays 23A-51A and 23A-51C.
Relays 23A-K51A and 23A-K51C are controlled by ATS slave trip units PS-263-50A-3 and PS-263-52A-2, rvspectively. These relays and trip units are located in ATS cabinet C2233A section
'B'.
The channel 'B' relays are similarly controlled by separate trip units and relays located in ATS cabinet C22338 section
'B'.
By design, both normally de-energized relays in instrument Channel ' A' and/or Channel 'B' must be in an energized state for the low RV pressure isolation function to occur.
EVENT DESCRIPTION
On January 4,1994, at 1834 hours0.0212 days <br />0.509 hours <br />0.00303 weeks <br />6.97837e-4 months <br />, an unplanned automatic actuation of the Group 4 (four) i portion of the Primary Containment Isolation Control System (PCIS) occurred during a scheduled surveillance test.
The actuation resulted in the automatic closing of the HPCI turbine steam supply isolation valves M0-2301-4 and -5 and caused the HPCI System to become inoperable.
I The event occurred during a calibration check of analog trip units that provide low RV steam pressure trip signals to the circuitry that was actuated.
The event occurred at step [8](f)(2) of Procedure 8.M.!-32.6 (Rev. 9) Attachment 1, "ECCS Analog Trip Cabinet C2233A Section B - Top Card File Calibration".
NFIC F ORM 366A (192)
NRC FOBM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150,o104 g
EXP!RES 5/31/95 t s.m Y
ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS LICENSEE EVENT REPORT (LER) 1%ISiMJ& "ZiMObl%l72 TEXT CONTINUATION EE*c"o*u"ISSICIN
?C 0X TO THE PAPERWORK REDUCTION PRCUECT Q1500104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
(EAR M
B PILGRIM NUCLEAR POWER STATION 05000 293 0
94
- - 001 -
00 TEXT (if more space is required, use adational copies of NRC Form 366A)(17) l The HPCI System was left in the isolated configuration and the I&C technicians who were l
performing the activity backed out of the surveillance test.
The Reactor Core Isolation Cooling System, Automatic Depressurization System, Residual Heat Removal System / Low Pressure Coolant Injection (LPCI) mode, and Core Spray System were verified operable in accordance with Technical Specification 3.5.C.2.
Problem Report 94.9004 was written to document the vent. The NRC Operations Center was notified in accordance with 10 CFR 50.72 at 2158 ars on January 4, 1994. A critique was convened at 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br /> on January 4, 1994. The
,tique was attended by applicable personnel including the I&C technicians who were performing the surveillance. After initial investigation and corrective action, the HPCI System was declared operable and
)
returned to normal standby service at 0326 hours0.00377 days <br />0.0906 hours <br />5.390212e-4 weeks <br />1.24043e-4 months <br /> on January 5,1994.
The event occurred during power operation while at 100 percent reactor power. The reactor mode selector switch was in the RUN position.
The RV pressure was approximately 1029 psig with the RV water temperature at 545 F.
CAUSE
The cause of the HPCI System becoming inoperable was the unplanned actuation of the Group 4 (four) circuitry.
The actuation caused the HPCI turbine steam supply isolation valves i
M0-2301-4 and M0-2301-5 to close automatically be to a low RV pressure signal.
I The cause of the low RV pressure signal was investigated but could not be determined with certainty. The investigation revealed the signal occurred approximately six seconds after the ATS cabinet C2233A calibration unit knob was depressed at procedure step [8](f)(2).
By depressing the knob, low RV pressure signals were generated as expected from master trip unit PIS-263-50A and related slave trip units including PS-263-50A-3 (controls relay 23A-K51A). Alone, this action should not have caused the event because the in-series, normally de-energized relay 23A-K51C (controlled by PS-263-52A-2) would have to be in an energized state for the isolation function to occur.
Based on circuit design, troubleshooting, and interviews with the I&C technicians who were conducting the activity, the investigation identified three possible causes of the event:
the cabinet C2233A calibration unit; a malfunction of relay 23A-K51C; and human performance. Of the possible causes, the most probable cause is believed to be human performance i.e., a continuity check of relay 23A-K51C instead of relay 10A-K32A at step [8](f)(3).
- Human performance was investigated. The investigation concluded human performance was the most probable cause of the event. The state of relays (e.g., 23A-K51C) is verified in several procedural steps. For a step, the relay state is determined by a continuity check of relay contacts (e.g., M1-TI).
The continuity check is accomplished using a test device (Fluke) at selected terminals. The investigation revealed the procedure was not specific regarding how the continuity check was to be accomplished.
Interviews and demonstration revealed differences in the method used for a continuity check.
Some checks are accomrlished using the voltmeter function while other checks are accomplished using the ohmmeter function of the test device.
The investigation concluded some checks should be accomplished using the voltmeter function while other checks should be accomplished using the ohmmeter function of the test device.
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YEAR BE JMBER PILGRIM NUCLEAR POWER STATION 05000-293 4
94
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00 TEXT pf more space is requned, use additional copies of NRC Form 366A)(17)
Moreover, verbal communication is necessary for performing parts of the surveillance procedure. The ATS cabinet C2233A is located approximately 40 feet from the ECCS panels including Panel C932. The RHR System channel ' A' relay 10A-K32A is located at Panel C932. Normally de-energized relay 10A-K32A (controlled by master trip unit PIS-263-50A) becomes energized as a consequence of depressing the calibration knob at step [8](f)(2).
The energized state of relay 10A-K32A is checked at step [8](f)(3).
The investigation concluded a continuity check of relay 23A-K51C contacts M1-Tl was performed instead of checking relay 10A-K32A at step [8](f)(3) due to less than clear communication. This action would have allowed current to flow through the test device in parallel to relay 23A-K51C contacts M1-Tl and, in conjunction with the energized state of relay 23A-K51A (controlled by slave trip unit PS-263-50A-2), resulted in the t
event six seconds after depressing the calibration knob at step [8](f)(2). The investigation concluded human performan'ce was the most probable cause of the event.
- The ATS cabinet C2233A calibration unit was investigated. Troubleshooting revealed 3-5 volt spikes when the calibration knob was depressed. This could have caused slave trip unit PS-263-52A-2 to provide a momentary trip signal causing relay 23A-K51C to momentarily energize. Slave trip unit PS-263-52A-2 is related to master trip unit PIS-263-52A. Neither PIS-263-52A nor slave trip units including PS-263-52A-2 were being checked at step [8](f)(2).
The..sS cabinet C2233A calibration unit is common to all trip units in C2233A.
Therefore, it is possible relay 23A-K51C energized because i
of a momentary trip signal from PS-263-52A-2 due to the calibration unit.
Prior to returning the HPCI System to standby service on January 5,1994, the calibration unit was checked in-situ. The testing included depressing the calibration knob and did not i
cause relay 23A-K51C to energize. After the calibration unit was replaced, the replaced unit was bench tested.
The testing indicated the voltage spikes would not have caused relay 23A-K51C to energize. The investigation eliminated the calibration i
unit (Rosemount model 710DU) as the probable cau;.e.
)
- A possible malfunction of normally de-energized relay 23A-K51C (Agastat model EGP-B-002) was investigated.
The relay's normally open contacts M1-Tl were verified closed at step [8M e)(84) and trip unit PS-263-52A-2 (controls 23A-K51C) was restored to normal service at step [8](e)(94). The procedure does not include a check of the state of relay 23A-K51C after step [8](e)(84) and prior to step [8](f)(2).
Contacts M1-Tl could have remained closed after PS-263-52A-2 was. returned to normal service.
i Prior to returning the HPCI System to standby service on January 5,1994, ' relay 23A-K51C was visually inspected and electrically tested with satisfactory results. The investigation eliminated relay 23A-K51C as the probable cause.
CORRECTIVE ACTION
On January 25 and 27, 1994, I&C technicians and supervisors received feedback regarding this event, continuity checks and the potential for parallel current paths during a continuity check, and the importance of communication.
The calibration unit was replaced as a preventive action and the HPCI System was declared operable and returned to standby service at 0326 hours0.00377 days <br />0.0906 hours <br />5.390212e-4 weeks <br />1.24043e-4 months <br /> on January 5,1994.
NRC FOl3M 366A -
U.S. NUCLEAR REGULATORY COMMisslON APPROVED BY OMB NO. 3150 o104 mm-y EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS
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YEAR NUMBER NUMBER PILGRIM NUCLEAR POWER STATION 05000-293 94
- - 001 -
00 TEXT (tf more space is required, use additional copies of NRC Form 366A)(17)
The HPCI System was removed from service for planned maintenance on January 11, 1994, at.
1140 hours0.0132 days <br />0.317 hours <br />0.00188 weeks <br />4.3377e-4 months <br />.
The NRC Operations Center was notified in accordance with 10 CFR 50.72 at 1204 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.58122e-4 months <br />. The system was removed from service to replace relay 23A-K51C as a preventive-action. After the replacement and satisfactory post work testing, the HPCI System was declared operable and returned to standby service at 1427 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.429735e-4 months <br /> on January 11, 1994.
Subsequent visual inspection and electrical tests of the formerly installed relay 23A-K51C were satisfactory.
Corrective action planned includes revision of Procedure 8.M.1-32.6 and companion ATS trip unit calibration procedures (8.M.1-32.5, 8.M.1-32.7, and 8.M.1-32.8) regarding continuity checks.
I&C personnel will be trained on the revised procedures as part of the routine training program.
SAFETY CONSEQUENC'tS This event posed no threat to the public health and safety.
The Reactor Core Isolation Cooling System, Automatic Depressurization System, Core Spray and RHR (LPCI mode) Systems were verified operable as specified by Technical Specification 3.5.C.2 during the period the HPCI System was inoperable.
This report is submitted in accordance with 10 CFR 50.73(a)(2)(iv) because the actuation, although a designed response to a sensed low RV pressure condition, was not planned.
This report is also submitted in accordance with subpart (a)(2)(v)(D) because the HPCI System was inoperable during the period the HPCI turbine steam isolation valves were closed while the related circuitry was not reset.
SIMILARITY TO PREVIOUS EVENTS A review was conducted of Pilgrim Station LERs submitted since January 1984.
The review focused on LERs submitted in accordance with 10 CFR 50.73 subpar +s (a)(2)(iv) or (a)(2)(v) involving a similar event or cause.
The review, although identifying previous unplanned Group 4 isolation events, revealed no similar event or cause.
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES The EIIS codes for this report are as follows:
COMPONENTS CODES Relay (23A-K51C)
RLY Valve, Electrically Operated (M0-2301-4 & 5) 20 s
SYSTEMS High Pressure Coolant Injection (HPCI) System BJ