ML20213E343
| ML20213E343 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Salem |
| Issue date: | 03/04/1983 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20213E339 | List: |
| References | |
| FOIA-87-152 NUDOCS 8303110612 | |
| Download: ML20213E343 (45) | |
Text
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MEMORANDUM FOR:
G. Lainas, Assistant Director for Operating Reactors Division of Licensing FROM:
R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration
SUBJECT:
SALEM 1 RESTART SER, DEFINITION OF SAFETY-RELATED, DSI(ICSB) INPUT The enclosure provides the DSI (ICSB) input for the subject SER. This input addresses the definition of " safety-related" used by the staff in its safety reviews, and the licensee's classification of the Reactor Trip System relative to this definition. We have concluded that the licensee has classified the Reactor Trip System including the trip breakers and their undervoltage trip mechanism as safety-related in accordance with the staff definition.
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R. L'ayne Houston, Assistant Director for Reactor Safety Division of Systems Integration cc:
R. Mattson DISTRIBUTION:
V. Hoonan Docket File ICSB Reading File R. Stevens (PF)(2)
Salem Subject File
Contact:
l R. Stevens, ICSB X29456 l
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o ENCLOSURE DSI/ICSB INPUT TO SALEM 1 RESTART SER The definition of safety related systems or equipment used by the staff in performance of safety reviews is derived (originally) f rom 10 CFR 100, Appendix A-Sections III.(C),
VI. a. (1), and VI.b.(3).
(See the enclosure to the memorandum from H.R.
Denton to AlL NRR Personnel dated November 20, 1981, Standard Definiti,ons For Commonly-Used Safety Classification Terms).
This definition is incorporated in the recently adopted regulations on environmental qualification, 1,0 CFR 50.49 (b) (1), and reads as fotLows:
" Safety-related electric equipment:
This equipment is that relied upon to remain functional during and following design basis
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events to ensure (i) the integrity of the reactor coolant pressure boundary, (ii) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (iii) the capability to prevent or mitigate the con-sequences of accidents that could' result in potential offsite exposures comparable to the 10 CFR Part 100 guidelines.
Design basis events are d' fined as conditions of normal e
l operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be designed to ensure functions (i) through (iii) of this paragraph."
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The Reactor Trip System including the trip breakers and their undervoltage t rip coil and the associated mechanical trip-Linkage clearly fall within this definition.
The undervoltage t rip feature is also required for both automatic and manual trip in order to meet the fait safe requirements of GDC-23.
The shunt trip feature f or the trip breakers is not required by pres 6nt regulations and, although it is provided to perform the manual t rip function, no credit is taken for this design feature in the safety analysis.
The manual trip function does provide undervoltage trip of the trip breakers (as discussed above) in addition to shunt trip.
The Reactor Protection System is Listed as item A.42 in Table 17.2-1, Salem Q-List, of the Salem UFSAR Rev.
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dated July 22, 1982.
The Q-List identifies those activities, services, structures, components and systems to which the operational Quality Assurance Program applies.
This system is described in Section 7.2.1.1 of the UFSAR as: "AtL equipment from sensors to trip breakers or initiation circuits of Engineered Saf ety i
Features are part of the Reactor Trip System."
The scope of l
l the syste= is f urther defined ac including the trip breakers and undervoltage t rip in the Ref erence 3 (WCAP-7488-L, Solid State Logic Protection System Description) cited in,Section l
7.2.1.1, and in t he,desc ription inc luded in Section 7.2.2.5.
Therefore, it is clear that the Reactor Protection System, including the trip breakers and their undervoltage trip mechanism, has been identified in the UFSAR as being safety-related in accordance with the staff definition of this term.
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ISSUES RELATED TO 5AlFM' RESTART
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AGENDA 1.
DESCRIPTION OF BREAKER 2.
LICENSEE CORRECTIVE ACTIONS 3.
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PUBLIC SERVIG ELECTRIC AND GAS CO.
REACTOR TRI.P 3REAKER PROBLEMS SALEM NO 1 AND 2 UNITS INTRODUCTION (RAU)
HISTORY OF SALEM NO,1 AND 2 UNIT OPERATION DECEMBER 1982 THROUGH PEBRUARY 25, 1983 (HJM)
DESCRIPTION OF EVENTS OF. FEBRUARY 22'AND FEBRUARY 25,1983 (JDD)
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EXPLANATION OF SOLID STATE PROTECTION SYSTEM (AO)
EXPl.ANATION OF DB-50 CIRCU,IT BREAKER OPERATION (W)
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HISTORY OF REACTOR TRIP BREAKER PROBLEMS AT SALEM (JG)
CORRECTIVE ACTION (JG., AD)
CONCLUSION (RAU) 6 O
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1, PSE8G VER]FIED SALEM SURVEIL. LANCE TESTItiG MEETS TECH SPEC REQUIREMENTS.
t 2.
NAINTENANCE PROCEDURES FOR THE UV TRIP DEVICES WILL BE DEVELOPED BASED ON NSD 74-02 AND NCD-ELEC-18.
3.
1[ WILL INSTALL NEW UNIT 1 UV ATTACHMENTS, i
u, PROPER OPERATION OF THE BREAKERS WILL BE VERIFIED BY PSEAG AND 1[ - PROGRAM BEING DEVELOPED.
5, il WILL VERIFY THAT THE UV ATTACHMENTS MEET SAFETY CLASSIFICATION SPECIFICATIONS FOR THE ORIGINAL RX SWITCHGEAR.
([)SURVEILLANCEOFBREAKEROPERATIONWILLBEINCREASEDTO MONTHLY INTERVAL,
(}[)
PROCEDURES.WILL BE REVISED TO REQUIRE THE OPERATOR TO ACTUATE THE REACTOR MANUAL TRIP SWITCH FOL-LOWING AN AUTOMATIC REACTOR TRIP, i,'DEVELOPAFORMALIZEDPOSTTRIPsEVIEWP.}0CEDURE.
9.
l! WILL SEND COMPILATION OF ALL TECH BULLETINS, MANUAL.S PERTAININGTOliEQUIPMENTATSALEM-SALEMWILLREVIEW A.ND INCOP. POP. ATE AS NECESSARY INic STATION DOCUMENTS, 4
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E IS CONDUCTING AN INTERNAL REVIEW 0F THElR PROCEDilRES F DISSENINATION OF TECHNICAL INFORMATION TO UTILITIES, PSEaG HAS IDENTIFIED THE DESIRED DISTRIBUTION OF THIS IN 0RMATION AS PART OF RECENT'IMPRnVEMENT IN THEIR HANDLING OF TECHNICAL DOCUMENTS..
11.
A REVIEW IS IN PROGRESS AT SALEM OF PAST EQUIPMENT FAILURES DOCUMENTED IN LER's., CR.
A PREVENTIVE MAINTENANCE PROGRAM WILL BE IMPLEMENTED BASED UPON RESULTS OF REVIEW, 4
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AUGUST 20, 1562, UNIT #2 "B" TRIP BREAKER FAILED TO'OPEN 05 SOLID ST; ? FROTECTION SYSTEN SIGNAL DURING SURVEILLANCE TESTING.
2.
. JANUARY 6, 1983, UNIT #2 "A" TRIP BREAKER FAILED TO OPEN ON SOLID STATE PROTECTION SYSTEM SIGNAL DURING A UNIT TRIP, 3.
FEBRUARY 22/.1983, UNIT #1 "B" TRIP BREAKER FAILED TO CLOSE DURING UNIT START-UP.
4.
FEBRUARY 22 AND 25, 1983, UNIT #1 "A" AND "B" TRIP BREAKERS FAILED TO OPEN DN A UNIT TRIP.
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SALEM RE;CTCR TRIP BREAKER FAILURE AUGUST 20, 1982 - UNIT #2 1.THE B REACTOR TRIP BREAKER WOULD NOT OPEN ON A SIGNAL FROM THE SOLID. STATE PROTECTION SYSTEM DURING lac SURVEILLANCE TESTING (WORK ORDER #90110),
2.
BREAKER WAS CAPABLE OF BEING TRIPPED MANUAL FROM CONTROL ROOM.
3.
PROBLEM WAS FOUND TO BE BINDING OF THE.UNDERVOLTAGE TRIP ATTACHMENT LATCH MECHANISM.
4.
CHANGED C0ll AND CLEANED TRIP Coll LATCH MECHANISM.
BREAKER TESTED SATISFACTORILY.
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, ll2 1, REACTOR TRIP BREAKER A FAILED TO T?.!3 DN SOLID STATE P SYSTEM SIGNAL.
2.
REFLACED REACTOR TRIP BREAKER FROM UNIT #2 WITH BRE UNIT #1 BREAKER TESTED SATISFACTORY.
3.
THE FOLLOWING REMEDIAL ACTION WAS TAKEN:
A.
WORK ORDERS WERE WRITTEN FOR UNIT #1 AND #2 REACTO BREAKERS TO PERFORM INSPECTION, CLEANING AND TESTING OF THE BREAKERS, 3.
PURCHASE ORDERS WERE WRITTEN TO WESTINGHOUSE TO PROV A SERVICE ENGINEER FROM DB-50 BREAKER INSPECTION AND MAINTENANCE ON #1 AND',#2 UNIT, 4.
BETWEEN JANUARY 13 AND 18, 1983s A SERVICE REPRESENTATIVE 0F WESTINGHOUSE WAS ON SITE TO PROVIDE TECHNICAL ASSIST IN THE OVERHAUL AND INSPECTION OF THE REACTOR. TRIP BREAKERS, 5.
THE FIRST BREAKER fb BE ADDRESSED t'AS THE EREAKER. T REMOVED FROM UNIT G THE FROBLEM FOUND WAS A BINDING OF-THE UNDERVOLTAGE Trip' ATTACHiENT, 6.
WESTINGHOUSE RECOMMENLED iHAT Cttsioim miuluBRICATION Or THE MECHANISM WITH CRC-2-26 SHOULD CORRECT THE PROBLEM, THIS WAS SATISFACTORILY f.CCO.".'LIC"::.
~7.
WESTINGHOUSE ASSISTED US DIFECTL 0:: T;;C U:::T #1A AND B REACTOR TRIP BREAKER OVERHAuts Aius int nuiu i+esuu< A AND 8 FxEAKERS, THE A AND B BYPASS BREAKERS WERF rn"of FTrn P.Y.CTaTinn preenimgt, i
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WESTINGHOUSE STATED THAT OVERALL THE BREAKERS WERE It! GOOD CONDITION COM?ARED TO SOME BREAKERS HE HAD 1.'3,-;ED ON AT OTHER LOCATIONS, 9.
WESTINGHOUSE REJECTED A SUGGESTION BY THE MAINTENANCE SUPERVISOR TO USE ANDER0L LUBRICANT ON THE UNDERVOLTAGE ATTACHMENT AND OTHER MOVING PARTS TO ELIMINATE A POSSIBLE FRICTION PROBLEM.
THE SERVICE ENGINEER STATED THAT THE
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GREASE WOULD ACT IN SUCH A MANNER AS TO COLLECT DUST AND DIRT, WHICH WOULD PROBABLY LEAD TO A FAILURE, AtiDER0L WAS NOT USED AND THE DEVICES WERE LUBRICATED WITH CRC-2-26, 10.
AT THE CONCLUSION OF THE INS?ECTION AND LUBRICATION, ALL BREAKERS WERE TESTED SATISFACTORILY, e
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DURING PLANT START-UP DN FEBRUARY 22,1983, "B" REACT 0:; TalP BREAKER F. AILED TO CLOSE PROPERLY.DUE TO THE BREAKER MECHANISM DUST COVER COMING LODSE AND JAMMING THE BREAKER.
2.
THE BREAKER WAS INSPECTED AND WAS FOUND TO BE UNDAMAGED.
3.
THE DUST COVER WAS REMOVED.
II,
THE BREAKER WAS TESTED SATISFACTORILY AND RETURNED TO SERVICE, 5.
ALL DUST COVERS HAVE BEEN REMOVED FROM NO. 1 UNIT BREAKERS.
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i SALEM REACTOR TRIP EF.dA;:'d FAILURE FEERVARY 22 AND 25, 1953 - UNIT #1 s.
1.
DURING PLANT START-UP DN FEBRUARY 22 AND 25, 1983, BOTH "A"
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AND "B" REACTOR TRIP BREAKERS FAILED TO TRIP WHEN A SOL STATE PROTECTION SYSTEM SIGNAL WAS RECEIVED, 2,
WESTINGHOUSE AND PSEEG MAINTENANCE AND ENGINEERING PE HAVE INSPECTED AND TESTED THE UNDERVOLTAGE' TRIP ATTACHMENis, 3,
THE PROBLEM IS BINDING OF THE UNDERVOLTAGE TRIP ATTACHMENT, CAUSED BY A LACK OF LUBRICATION TO THE TRIP LATCH, 4.
AREVISEDMAINTEN5NCEAPPROACHHASEEENDEVELOPED.,ASFOLLOWS:
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Ud5TINGH00SE D5-50) 1.
LIGHT DUTY BREAKER WITH NO THERMAL OR MAGNETIC OVERLOADS.
2, NO LUBRICATION REQUIRED BY WESTINGHOUSE TECHNICAL MANUAL (I.B. 33-850-3D, PG, 5),
3.
BREAKERS WERE TRIP TESTED'USING THE UNDERVOLTAGE COIL EVERY SIXTY (60) DAYS AS PART OF 18C DEPARTMENT SURVEILLANCE.
- TESTING, 4.
NO KNOWN PROBLEMS EXISTED WITH THE BREAKERS.
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7, SALEM REf.CT0n in1." :n-AKER FAILURES SHORT-TERM MAINTENANCE APPROACH 1.
MAINTEN4NCE PROCEDURES FOR THE UNDERVOLTAGE TRIP ATTACHMENTS WILL BE DEVELOPED BASED ON NSD74-02'AND NCD-ELEC-18.
2, WESTINGHOUSE WILL PROVIDE FOUR NEW UNDERVOLTAGE TRIP ATTACH-MENTS FOR NO. 1 UNIT.
3.
PROPER OPERATION OF THE BREAKERS WILL THEN BE VERIFIED BY PSE8G AND WESTINGHOUSE.
4.
THE lsC DEPARTMENT WILL TEST THE REACTOR TRIP BREAKERS AFTER MAINTENA' ICE IS PERFORMED.
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I COMMISSION BRIEFING i
SALEM EVENT l
OF FEBRUARY 25, 1983 i
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L CONTROL ROOM INDICATION
- REACTOR TRIP -
i POSITIVE
- 1. REACTOR TRIP BREAKER "0 PEN" SSPS DISPLAY l
BREAKER CONTROL PUSHBUTTON'
- 2. R0D POSITION INDICATORS
- 3. R0D BOTTOM LIGHTS
- 4. NUCLEAR INSTRUMENTATION j
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- 5. PLANT COMPUTER i
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CONTROL ROOM IllDICATOR
- REACTOR TRIP -
FEEDBACK 1.
SSPS LOGIC DISPLAY 2.
SECONDARY REACTOR TRIP ALARMS o
LOW-LOW LEVELS IN SGs o
REACTOR TRIP / TURBINE TRIP 3.
GENERATOR BREAKER OPEN o
STOP VALVES AND GOVERNOR VALVES CLOSE
' URBINE SPEED LESS THAN 1800 RPM AND DECREASING T
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FEBRUARY 22 EVENT SEQUENCE
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TIME 2155
- REACTOR BUS TRANSFER FROM 0FF-SITE TO ON-SITE IN PROGRESS AT 20% POWER.
DURING TRANSFER LOSS OF #13 RCP AND #12 MAIN FEED PUMP (MFP) OCCURS DUE TO LOSS OF CONTROL POWER (#12 MFP ONLY OPERATING MFP) 2156 :54
- REACTOR TRIP SIGNAL FROM LOW LOW LEVEL
- 13 S/G
- AUXILIARY FEEDWATER (AFW) PUMPS START 2156 :58
- MANUAL REACTOR TRIP DilE TO DEGRADING CONDITIONS
- TURBINE TRIP: REACTOR TRIP BREAKERS OPEN 2204
- SAFETY INJECTION (SI) DUE TO 100 PSI DP BETWEEN #13 MAIN STEAM LINE AND OTHER STEAM LINES 2206
- OPERATOR NOTED #11 RCP HAD TRIPPED (WITH BOTH #11 AND #13 RCPs LOST, NO PRESSURIZER SPRAY TO CONTROL PRESSURE)
- BOTH PORVs LIFT FROM PRESSURE INCREASE DUE TO SI FLOW AND LOSS OF SPRAY FLOW
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2211
- SI TERMINATED BY OPERATORS
- BOTH PORVs CLOSE
- PLANT STABILIZED IN MODE 3 2346 a - NRC NOTIFIED VIA ENS
- 100 PSI DEVELOPED BECAUSE #13 SG SUPPLYING TURBINE AFW PUMP AND #13 RCP NOT RUNNING "NRC WAS INFORMED THAT THE SG LOW LOW LEVEL TRIPPED THE Rx A THAT THE MANUAL TRIP INITIATED NEARLY SIMULTANE0USLY
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4 TIME FEBRUARY 23 BLOCK VALVE FOR PORV PR-2 CLOSED BECAUSE 0628 OF PORY SEAT LEAKAGE e
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i-SEQUENCE OF EVENT FOR FEBRUARY 25, 1983 EVENT INITIAL CONDICTIONS - REACTOR POWER 12% TURBINE ON LINE AND GENERATOR SYNCHRONIZED WITH GRID: FEEDWATER SYSTEM IN MANUAL CONTROL TIME 0021
- - LOW LOW WATER LEVEL #12 STEAM GENERATOR
- REACTOR TRIP SIGNAL GENERATOR BY SSPS AND INDICATED IN CONTROL ROOM
- PLANT PARAMETERS NOT CONSISTENT WITH SCRAM 0021 :30 (APPROX)
- REACTOR MANUAL SCRAM FROM CONTROL ROOM
- PLANT PARAMETERS INDICATE SCRAM i
0048 - 0115
- EACH BREAXER TESTED VIA SSPS 5 TIMES -
"B" TRIP BREAKER FAILED 5 TIMES, A TRIP BREAKER FAILED 3 TIMES 0130
- ALERT DECLARED 0146
- ENS NOTIFICATION MADE 0200
- ALERT TERMINATED
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6 37 38 PR RC L0 FLOW REACTOR STM GEN II SIM GEN 11 STM DIFF P CONDENSER SIM GEN II i
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HIGil FLUX OPEN & P-B' ilIGH PRESS LEVEL LO LVLAFLO SI LOW LEVEL REAC TRIP REAC TRIP REAC TRIP REAC TRIP REAC TRIP REAC TRIP TURB TRIP TURB TRIP i
PR RC LO FLOW REACTOR STM GEN 12 SIM GEN 12 STM DIFF P TURBINE STM GEN 12 LOW RANGE OR RCP BKR COOLANT LOW-LOW FEEDWATER LOW P2 BEARING HIGH-HIGH I
7 HIGH FLUX OPEN & P-7 LOW PRESS LEVEL LOW LVL&FLt SI LOW OIL LEVEL
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REAC TRIP REAC TRIP REAC TRIP REAC TRIP REAC TRIP REAC TRIP TURB TRIP TURB TRIP f
IR 4KV PRESSURIZR STM GEN 13 STM GEN 13 STM DIFF P TURBINE STM GEN 13 l
HIGH FLUX GRP BUSES HIGH LEVEL LOW-LOW FEEDWATER LOW P3 TilRUST INC HIGH-HIGH 13 UNDERFREQ LEVEL LO LVL&FLO SI BRG FAIL LEVEL REAC TRIP REAC TRIP REAC TRIP REAC TRIP REAC TRIP REAC TRIP TURB TRIP TURB TRIP l
SR 4KV PRESSURIZR STM GEN 14 STM GEN 14 STM DIFF P TURBINE STM GEN 14 i
llIGH FLUX LOW PRESS LOW-LOW FEEDWATER LOW P4 OVERSPEED HIGH-HIGH I
19 REAC TRIP GRP BUSES 5.I. &
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OVERPOWER OVERTEMP MANUAL MANUAL GENERATOR MANUAL l
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ACTUATION j
31 REAC TRIP REAC TRIP.
REAC TRIP REAC TRIP TURB TRIP TURB TRIP j
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.RK HISTORY OF'PWR SCRAM BREAKER' FAILURES Ys/ss SINCE 1973 THERE HAVE BEEN APPR0XIMATELY 340 PWR REACTOR YEARS OF OPERATION:
o 220 WESTINGHOUSE 70 B&W 50 CE DURING THIS PERIOD OF TIME THERE HAVE BEEN 35 KNOWN SCRAM BREAKER FAILURES:
o 21 WESTINGHOUSE 13 B&W 4
1 CE THE AVERAGE NUMBER OF SCRAM BREAKER FAILURES PER REACTOR YEAR BY VENDOR IS:
o 0,095 WESTINGHOUSE 0,19 B&W 0.021 CE THE CORRESP0 DING AVERAGE NUMBER OF REACTOR YEARS BETWEEN BREAKER FAILURES o
- 11..
i 5.3 B&W i
48 CE I
THE CORRESPONDING EXPECTED NUMBER OF SCRAM BREAKER FAILURES IN A CALENDAR o
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YEAR BY VENDOR IS:
i 2,9 WESTINGHOUSE 1.3 B&W Foz # /f a.
0.15 CE D/4
i F ILURF MECHANISMS'FOR 5 CRAM BREAKERS 35 SCRAM DREAKER FAILURES SINCE 1973 25 UUE TO UNDERVOLTAGE COIL MECHANISM FAILURE OR BINDING o
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y n/n SALEM I RESTART REPORT I.
IDENTIFICATION OF CAUSE OF FAILURE
SUMMARY
AND INITIAL FINDINGS Initial inspection of the under-voltage trip unit indicates a possibility of multiple contributing causes of failure.
Possible contributors are (1) dust and dirt; (2) lack of lubrication; (3) wear; (4) more frequent
-operation than intended by design; (5) corrosion from improper lubrication in January,1983; and (6) nicking of latch surfaces caused by vibration from repeated operation of the breaker. The contributors appear to be cumulative with no one main cause. The initial investigation does indicate that the failure is age related and that a new device would perform properly.
Many surfaces of the latch mechanism are worn and the additional friction
" tends to prevent proper operation. Proper lubrication throughout the life of the device may have prevented the wear that can be seen on the sample.
The tests and examinations proposed by the staff and its contractor will attempt to determine the cause of failure and if possible reproduce it.
The following summarizes the initial findings and lists the proposed tests.
DISCUSSION AND CIRCUMSTANCES A site visit was conducted on March 3,1983 by NRR and Franklin Research
- Center personnel to inspect the Type DB-50 Circuit Breaker Undervoitage Trip Attachment in an effort to determine the most probable cause of For A-Tr7-is1 Dh
^
. failure. The reactor trip switchgear rooms for Units 1 and 2, each of which contain four 08-50 circuit breakers, were visually inspected and resulted in the following observations.
All four 08-50 circuit breakers and undervoltage trip attachments for Unit I were removed from the switchgear cabinets. The enclosures were generally clean and free of dust. The ambient temperature was between 85'F and 95'F with warm exhaust air from inverter cabinets being directed at the DB-50 switchgear cabinets. The spacing between cabinets is approximately 3 feet.
All four 08-50 circuit breakers for Unit 2 were inspected.
The undervoltage trip attachments were removed, however. The circuit breaker cabinets contained a layer of loose dust approximately 1/16
.i of an inch thick. The ambient temperature was in the 70*F range.
The undervoltage trip attachment mounts on the top of the circuit breaker platform, to the right of the shunt trip which is several inches from the bottom of the switchgear cabinet.
Interviews were conducted with Mr. Ketchum, an electrical maintenance super-visor, who discussed the circumstances of the removal of the circuit breakers that were involved with the incident on Unit I and Mr. Leo Roland, another electrical supervisor, who had also worked on the circuit breakers in question in August of 1982. The information received was that the circuit breakers
. and their undervoltage trip device had been operated frequently and had operated during surveillance testing within a few days prior to the incident.
A request was made to Salem management to provide one of the undervoltage trip devices, and a shunt trip coil, for testing at Franklin Research Center (FRC). This request was complied with and an investigation of these devices is now underway at FRC.
RESULTS OF INITIAL EXAMINATION Initial investigations have noted roughness in the operation of the trip latch. There is some dragging of the mechanism, 'and portions of the latch mechanism have obvious signs of wear.
Possible contributing factors to the f
failure to operate are a lack of lubrication, wear, jarring of the under-
- i voltage device from the circuit ' breaker operation and more frequent opera-tion of the undervoltage trip device than was intended during design.
It is postulated that under most industrial applicaitons, the undervoltage device would be used very infrequently, and probably would only be operated during test sequences at perhaps yearly or longer intervals. Therefore, in industrial applications, the device would operate only a few time, perhaps l
20 or 30 cycles during its lifetime, and would not be 'a normal tripping mechanism for the breaker. However, in its use at Salem and other nuclea power plants, it is the prime tripping device for the circuit breaker, and as such is called upon to operate on the order of 50 times per year.
l This would mean, at its current age, in 1983 there would be possibly 400 to 500 trip operations on this device.
i
i 4-During the initial investigation, it was noted that the shunt trip coil has been operated since August,1982 once every 7 days rather than at longer intervals. This means that the circuit breaker is now tripped and closed
~
every 7 days, causing the jarring of the entire mechanism of the circuit breaker and its attached relays and coils due to the normal operation of the breaker. This may or may not be significant -in that the undervoltage relay would stay energized during these trips and its latch mechanism would
~
be jarred somewhat by operation of the breaker. This could possibly add to the friction which is building up in the latch mechanism from normal operation by causing the latch mechanism to just slightly nick the surface that it rides on thereby tending to prevent operation. Further investiga-
~
tion will try to determine whether this is indeed a contributor. It appears from initial inspect [on of the device that wear and roughness of mating 4
surfaces in the trip latch are contributing causes. Proper lubrication may have prevented the current situation or could reduce this roughness to the point where proper operation could occur.
Further investigation will t ttempt to determine whether the CRC-2-25 lubricating, cleaner spray added to the problem by either causing cor-rosion or removing all residual lubrication from initial construction and possible caking of the dust and dirt.
It appears that from the time of,,
initial construction of the undervoltage trip units, up until January of 1983, no lubricatioh whatsoever had been performed, and then in January of 1983, lubrication was performed by the maintenance personnel in conjunction 1
. with a Westinghouse technician. At this time, CRC-2-26 lubricant cleaner was sprayed on all four trip devices associated with Unit 1.
This lubricant is being procured by FRC for testing purposes.
LISTING OF EXPECTED INVESTIGATON BY NRC CONTRACTOR (FRC)
The first test will be to perform various deenergizations and energiza-1.
tions of the undervoltage trip unit and monitor the device under various conditions.
The second test will be to disassemble the latch mechanism to observe 2.
the surfaces of the various parts of the latch, and photograph these surfaces through a microscope to determine various levels of wear in these surfaces.,
The third test is to deterine the effects of CRC-2-26 spray on the 3.
various types of metals used in this device. An attempt will be made to use metals other than those in the actual device.
If possible, we will determine the chemical consistency of this spray, hopefully through the manufacturer.
To prove that the sample'undervoltage trip unit is identical to all Salem devices, a visual inspection of all existing Salem Unit I and 2 undervoltage trip units will be performed. This can take place at
. Salem.
No disassembly is needed. The devices tag be mounted on the circuit breakers or loose. This should be done'as soon as'possibli.
Tuesday, March 8, 1983 is recommended.
If further tests ar required they will be based upon the results of these-initial tests. All tests will be non-destructive tests such that the device will be able to be used for further testing
- d return to the utility.
~
ADDITIONAL TEST TO BE CONDUCTED BY THE LICENSEE AS REVISED BY STAFF This test wi.ll require the use of a spare circuit breaker. The undervoltage trip device and the shunt coil would be mounted upon the breaker, and the breaker would be operated repeatedly to determine the effect upon the shunt coil and undervoltage. trip unit.
It is surmised that while the device is energized and the breaker trips and closes a number of times, additional friction of the trip latch may occur from the vibration. This test is des-cribed in detail in the section titled, Revised Surveillance of Reactor l
Trip Breaker Operation.
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. II. REVISED SURVEILLANCE OF REACTOR TRIP BREAKER OPERATION AND VERIFICATION TESTING
SUMMARY
AND INITIAL FINDINGS PSE&G proposed the following increased surveillance of reactor trip breaker operation:
a.
Main and bypass breakers will be shunt-tripped weekly and, l
b.
Main breakers will be UV-tripped monthly.
The acceptability of the revised surveillance of reactor trip breaker opera-tion has been evaluated by the staff. - Based on an analysis conducted by the staff which considered reactor trip system unavailability, reactor trip breaker failure rates and test intervals, the following conclusions _have been made.
First, the proposed test of each reactor trip breaker under-
,y voltage trip attachment once every 30-days is acceptable. Second, the proposed test of the shunt trip attachment once every 7-days is considered to be excessive and may impact on the reliability of the reactor trip system l
by increasing the potential for a single failure. During testing, a single failure in the logic portion of the reactor trip system could prevent an automatic SCRAM. Thus, it is recommended that the shunt trip attachment be tested on the same schedule as the undervoltage trip attachment i.e., once every 30-days.
DISCUSSION The acceptability of the proposed test intervals for the reactor trip breakers was based on our review of reactor taip breaker failure rate data obtained from
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LERs. The reactor protection system unavailability goal of 3 X 10~
(used in both NtlREG-0460 " Anticipated Transients Without Scram for Light Water Reactors," and by 'the ATWS Task Force and Steering Group in the development of the proposed ATWS Rule) was used in evaluating the PSE&G proposed test intervals.
In addition, the following considerations were incorporated into the abo've staff recomrendation:
1.
The shunt trip coil provides a diverse means of tripping the reactor trip breaker which is electrically independent of the undervoltage trip coil. The undervoltage coil is supplied by a 48 Vdc source and is deenergized to trip. The shunt trip coil is supplied by a 125 Vdc source and is energized to trip.
2.
The shunt trip coil is an energize-to-actuate device and is.not
~
" fail safe" in that a loss of power will not cause a trip. However, the shunt trip is powered from a reliable Class IE battery backed source.
3.
Since the shunt trip coil is an energize-to-actuate device, it is not subject to the constant heating effects that the continuously energized undervoltage coil experiences. ThesehAatingeffectsmay contribute to the higher failu're of the undervoltage coil mechanism. /
4.
The mechanical construction of the shunt trip mechanism is less complex than the undervoltage trip mechanism. The shunt trip does not rely on the successful operation of the complex latching mechanism which has been attributed to be the source of the majority of failures of the undervoltage trip.
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5.
The majority of the electrical circuit breakers used in the high voltage 4
electrical distribution system have de powered energize-to-actuate shunt trip coil mechanisms. These breakers are used for manual as well as automatic trip functions for load shedding and power switching.
Reliability of energize-to-actuate shunt trips in similar applications 4
throughout the nuclear power industry has been shown to be signiff-cantly higher than for devices that are constantly energized.-
i 6.
PSE&G is revising procedures to require the operator to manually tirp the reactor following indication of an automatic reactor trip. Thus, on any. trip signal, diverse means will be used to trip the breakers.
7.
Over 70% of the known reactor trip breaker failures were caused by undervoltage coil mechanism failures.
8.
Most of the concerns relating to the events at Salem on February 22 and 25,1983 are related to the operation of the undervoltage coil.
During the events at Salem, the shunt trip functioned properly.
VERIFICATION TESTING i
It is recommended that a bench test be performed on one DB-50 reactor trip circuit breaker. The purpose of the test is to cycle the 0B-50 with the i
undervoltage trip attachment and shunt trip attachment in place fo'r a total of 2000 cycles to determine'if any adverse effects can be identified and provide that a properly maintained breaker and its sub-components can operate for an extended number of cycles. The breaker will be tripped
a.
10 -
with each cycle being alternated with the undervoltage trip and shunt coil trip. Ambient temperature should be 100*F to simulate the expected service environment and the circuit breaker should be cycled no sooner than once every thirty minutes to allow for return to steady state conditions. The results of each circuit breaker operation will be documented and a visual check made. Additional d'etails for this type of test will be provided at a later time.
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~C III. MAINTENANCE PROCEDURE AND PREOPERATIONAL VERIFICATION PROGRAM We have reviewed the licensee's maintenance procedure, Salem Generating Station Maintenance Department Manual Maintenance Procedure M3Q-2, Revi-sion 1.
This document includes a procedure for verifying proper operation of the undervoltage trip attachment and testing of the undervoltage trip i
attachment coil following their replacement. We have also reviewed the licensee's proposed reactor trip switchgear operational verification program which references M3Q-2. We have the following comments and re-commendations concering thee documents:
The maintenance procedure does not specify whether the maintenance o
and testing described are applicable to both the main and bypass breakers.
It should specify that it does.
The maintenance procedure should specify required actions to be taken o
in the event any acceptable tolerances, as identified in Enclosure 7 of M3Q-2, are not met.
?
The frequency of all maintenance and testing specified in the pro-o cedure, with the exception of the verification testing identified following undervoltage trip attachment replacement, should be specified.
We recommend that the procedure be modified to require cleaning of o
i the entire circuit breaker room, the removal of all four circuit breakers, the cabinets cleaned (vacuumed), and the breakers cleaned every refueling outage.
I -+ -- -
. Section 9.7.2.1 specifies that the undervoltage trip attachment be o
cleaned with a standard solvent. The procedure should specify the exact solvent to be used. We will request our contractors, FRC and BNL, to determine the adequacy of this solvent and any potential adverse effects from its use.
(This evaluation need not be completed prior to plant startup). '
Section 9.7.2.2 specifies the composition of lubrication to be, applied o
to specific points of the the undervoltage trip attachment. This section should. be specific as to whether the mechanism is to be lubricated each time maintenance is performed. We will request FRC and BNL to determine the adequacy of the lubricant and the points of application specified, and the frequency of lubrication.
9 Any undervoltage trip attachment that does not successfully complete o
25 consecutive cycles of testing to be performed by Westinghouse should not De accepted or installed by the licensee.
Section 9.7.4.15 specifies the testing to be performed ~ on the under-o voltage trip ar**chment coil following its replacement. The maintenance procedure should be revised to require that all replacement undervoltage trip attachments successfully complete 25 consecutive cycles of test.ing prior to installation in the plant and start of the 10 test cycle spect fied in the maintenance procedure. The time between each of the 10 test cycles should be specified. We recommend 30 minutes for the reasons
. specified in II above.
We believe the increase in test cycles and acceptance criteria specified if any failures occur during this testing are reasonable, and should be incorporated into maintenance procedure M3Q-2.
Technical Department Procedures Nos.11C-18.1.011 and IIC-18.1.010, o
referenced by the licensee, should be reviewed and acceptability de-termined by the Region.
Following revision of the maintenance procedure and associated proposed reactor trip switchgear operational verification program to incorporate the above comments and recommendations, we will re-review the documents and provide another report, which will include the results of our contractors
~
evaluations, documenting our final evaluation and conclusions concerer19 the adequacy of the maintenance procedure and preoperational verification program.
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F 12 (7 i R!chrd A.Uderrtz Public Service E:ecir.c and Gas Company P.D. Box 236. Hecocks Bridge.JU 06038 609 9354010
% os Pramount..Ya,s.6 March 1,1983 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. c.. 20555 Attention: Mr. Darrell G. Eisenhut, Director Division of Licensing Gentisment REACTOR TRIP BREAXER FAILURE NO. 1 UNIT SALEM GENERATING STATION DOCKET No. 50-272 The purpose of this letter is to document our investigation of two reactor tr.ip breaker failures and provide corrective actions to be taken. On February 22 and.-25,1983, the Salem Unit I reactor trip breakers failed to open upon receipt of a valid trip signal from the reactor protection system. In both instances, the manual trip was used to shut down the unit. PSE&G has determined that the reactor trip breaker undervoltage trip attachment failures were caused by a lack of proper lubrication on the latch. Westinghouse expert opinion cor. curs with this based upont a. An inspection of the undervoltage trip attachments. b. A review of PSEGG accounts of the tests performed after the failures. Previous; Westinghouse experience which indicates that I c. the lack of lubrication has besn the cause of simila~r Previous failures. d I Foro-t 7 !S L U D/$. l Y I I I ~
r o l .c i I ~ $r. Darrell G. E1'sei.qur 3-1-83 U. S. Nuclear Regulatory commission ,}fo.R b Es presented in cur meeting on February 28, 1983, our investigation of these incidents is summarized herein. PSE&G has reviewed the plant data from the events of February 22 and 25, 1983 to evaluate any potential safety impact on the primary system. Reviewrof the primary coolant parameters did not reveal any significant perturbations and followed trends that would be expected in a normal plant trip. I The bounding case'in the PSAR is the loss of normal feedwater at 102% power with only one auxiliary feedwater pump starting. In that transient, two steam generators boil dry and the other two drop to a level where approximately 50% of the tube bundles are expcsed. This.provided sufficient heat removal to preclude boiling in the primary system. This is a more limiting case than the two recent incidents at Sal'em, where on February 22nd as a result of the transient, the water level in three steam generators briefly dropped to a level equivalent to approximately 20% of the tube bundle exposed. On February-25th, the level in one steam generator again briefly dropped to approximately this same level, on both occasions, there was automatic auxiliary feedwater initiatl,en. The potential for'wetirhammer in the steam generator feed ring exists when the fee'dwater flow is interrupted long enough to allow the feed ring to drain. In both recent instances, there j was no flow interruption since auxiliary feedwater was initiated autematically. In addition, "J-tubes" have previously been installed in the feed rings. In conclusion, the events of February 22 and 25,.1983 were within bounds of FSAR analyses and did not have the necessary prerequisites for feedwater line waterhammer. I cur review of the breaker failures has resulted in a program of corrective actions to assure that such failures will not g) l recur.' These corrective actions are described below: 0 1. P3EAG has verified the salem surveillance testing meets ([It the technical specification requirements. ( Procedure PD18.1.004/5 Solid State Protection System Reactor Trip Br.eakers and Permissive P-4 Test Train A/3 satisfy the requirements for testing the reactor trip breakers. t i i t I ._ w._ --- -r-~~a. -= ~~ o n --r -"-- ur. . w'.
~. Mr. Darrell C. Eisenhut 3- -U. S. Nuclear Regulatory 3-1-83 com=ission i i Procedure PD18.1.008/9 Solid State Protection System P Test Train A/B satisfied the requirements for testing th unctional autcmatic trip logic. e 2. A detailed maintenance procedure M3Q-2 entitled Reactor Trip and Bypass ACB Inspection and Test, which include the undervoltage trip attachment, has been developed and s approved. data letter NSD-TB-74-2, Westinghouse Procedure and the Westinghouse Instruction Book for DB-50, DEF-16 and DBL-50-ACB's. This includes electrical testing of .the breaker, notification of the Technical Department of inspection. hold points.the need for post maintenance testing and a 3. hcuse and will be installed on each of the four N st breakers. q Va Westinghouse will provide technical assistance Aj,q to PSEEG to assure that No. 1 Unit undervoltage trip att
- are installed properly and that the breakers operate properly..
achments 4. Proper operation of the breakers will be verified prior to placing the breakers in service. to returning td.. service. proper operation will be developed and comp This program will take into provided by Westinghouse. consideration statistical data and recomme i p5. the UV attachments meet the specification requireme {tg(40 A for the original reactor trip switchgear. T/4 i 6. ine'reased as follows: Surveillance of reactor trip breaker operation will be i Main and bypass breakers will be shunt-tripped weekly. a. / b. Main breakers will be UV-tripped monthly. Proposed technical specification changes will be submitted l as appropriate. i l + ,n -,,-,-,,,_,,--nn a. ,,,-,,,,,O,.,,.~,, w,_ _,,,,,_,,,-.-,,,-.-.m
Mr. Darrell G. Eisenhut 3-1-83 U. S. Nuclear Regulatory coraission 7. The following tests will be performed after maintenance on reactor trip breakers to demonstrate operability prior i to return of the breaker to services a. Breaker will be shunt-tripped. , [,., ,'j / b. Breaker will be UV-tripped. O '/ ' f c. Breaker will be time-response tested.' 8. Emergency Instruction I-4.3, Reactor Trip, for Salem Units 1 and 2 will be revised to include the requirement to manually. trip the reactor trip breakers on all reactor trips. h The revision to this procedure and the basis for this additional action will be disseminated to all licensed operators. 9. A formal reactor trip / safety injection post trip review procedure will be developsd and issued as an Operations Department Directive. This procedure will specify the b-review and documentation necessary to determine the cause D of the event and'also determine that affected equipment ~ performed in its intended function. The procedure will also include management authorization requirements for l startup. All licensed operators will be informed of the l requirements of this docunent. 10. A review of L2R's, deficiency reports, maintenance work L sheets and work orders is in progress to identify items g' requiring preventative maintenance. Our preventative maintenance program will incorporate the results of this review to be completed by January 1, 1984. b A reactor trip and hypass breaker traceability programwill be eetablished to in 11 a the breakers will be traceable to a particular breaker and its location. This will be accomplished by April 1,/-. 1983. 12. Westinghouse has committed to provide PSESC with a compilation of all technical bulletins, manuals, etc., pertainhg to (,,b - Westinghouse equipment utilized at Salem. These will be V reviewed and incorporated into station documents as necessary.' in a timely manner. I ~: -
- .__ x _ _.--.
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o. Mr. Darrell G. Eisenhot 3-1-83 U. S. Nuclear Regulatory Com=1ssion 13. Work orders will be reviewed by QA to insure that there f is proper designation of safety related items. Por safety L-related work, QA will' establish proper inspection and/or / surveillance coverage. ( In addition, PSEEG istundertaking a thorough review of its Operational QA Program to identify changes necessary to improve performance. In our meeting with the staff on February 28, 1983, we were requested to clarify the safety classification of the reactor trip breakers. The reactor trip breakers are part of the Reactor Trip System which is a safety-related system. In the design and construction of Salem Generating Station, PSEaG considered as safety-related, those structures, systems and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. Salem UPSAR Section 7.1.1.1 states that the Reactor Trip System consists of equipment which initiates reactor trip or activates engineered safety features. Included is equipment from sensors to actuating devices.- The. reactor trip breakers and the under-voltage attachment'are safety-related. The shunt-trip attachment is not a functional part of the reactor trip system. Corrective action Items 1 through 9 will 'e completed prior b to startup. Corrective action Items 10 through 13 will be com-plated as described therein. We believe that accomplishment of the corrective actions identified ~ above will preclude recurrence of these and similar events and provide adequate confidence that Salem Unit 1 can be safely returned to service. sincerely, ,e ~
k MAR 4 1933 1 MEMORANDUM FOR: James P. Knight, Assistant Director for Components and Structures Engineering Division of Engineering i FROM: R. Wayne Houston, Assistant Director for Reactor Safety Division of Systems Integration
SUBJECT:
PROPOSED SCRAM BREAKER TEST FREQUENCIES AT SALEM UNIT 1 The purpose of this memorandum is to transmit ICSB input to the EQB SER justifying the restart of Salem Unit 1 following the events of February 22 and 25,1983 during which both reactor trip (scram) breakers failed to open on comand. The enclosed inforvation can be used to fom a basis for staff acceptance of the revised reactor trip breaker test frequencies proposed for Salem by Public Service Electric and Gas Company (PSE&G). pSE&G has proposed to test each reactor trip breaker undervoltage coil once every 31 days by simulating a solid state protection system automtic scram signal (e.3., pressurizer high pressure). The previous test interval (required by Technical Specifications) was once every 62 days. In addition, pSESG has proposed to test each reactor trip breaker shunt trip once every 9 7 days by manually energizing the coil. This is done using a pushbutton test switch at the breaker. The previous test interval (required by Tech-nical Specifications) was once within seven days prior to each startup. i During both undervoltage and shunt trip coil testing, the bypass breaker ' opposite the breaker under test is placed in service to avoid inadvertant reactor trips. , provides calculations of acceptable test frequencies for the reactor trip breakers based on reactor trip breaker failure rate data obtained from ICSB LgR searches. The reactor protection system unavaila-bility goal of 3x10- (used in both NUREG-0460 " Anticipated Transients Without Scram for Light Water Reactors,"and by the ATWS Task Force and Steering Group in the development of the proposed ATWS Rule) was used in arriving at these test frequencies. Two unavailability models are provided for your consideration. The first model treats the two series reactor trip breakers in the Westinghouse design as a single system. Thus, the test l frequene,y obtained for this model is that at which the system must be tested to achieve a system unavailability of 3x10-5 The second model treats the reactor trip breakers independently. The fact that only one g., of the two breakers must function for system success is designed into the pb model. The test frequencies obtained from these models are 6 and 35 days, I espectively. This roughly corresponds to, and therefore, tends to support e test frequencies proposed by PSE&G. Fo.r#--P7 /9
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~.' a J. P. Knight 2 The models used were not set up to obtain test frequencies based on the specific failure mechanism (i.e., undervoltage coil mechanism failure versus shunt trip coil failure). These calculations should be viewed as preliminary attempts to ascertain whether the proposed test frequencies are reasonable to the staff. To arrive at an ideal test frequency based on the history of reactor trip breaker failures would require a detailed analysis perfomed by reliability specialists using a more reliable data base. In addition to the above calculations, enclosure 2 contains information which we believe should be used as the basis for acceptance of the PSE&G proposed test frequencies based on engineering judgement. Based on this inforcation we believe that the Proposed 7 day test interval for the shunt trip coil may be too frequent in that the benefits of increased system availability due to increased testing may be more than offset by the potential for not restoring the system to its normal operating r.xxie following the test and the increased probability of system failure while testing is being performed. While a trip breaker is under test, both the other trip breaker and the bypass breaker replacing the breaker under test receive signals from a single protection systen logic train. Thus during testing, a single failure in the logic portion of the protection system could prevent an autoratic scram. It appears that testing the shunt trip every 7 days may be counterproductive. A detailed analyses perforced by qualified personnel would have to be performed to deterr.ine this point. In conclusion, ICSS supports the PSET.G proposed revised test frequency of q once per 30 days for,the undervoltage coil mechanism, but believes that a 30 day frequency for testing of the shunt trip coil is sufficient. R. Wayne Houston, Assistant Director for Reactor Safety i Division of Systems Integration
Enclosures:
Distribution: As stated Central File ICSB R/F l cc: R. 1'.attson Salem Unit 1 S/F G. Lainas R. Kendall (PF) T. Ippolito AD/RS Rdg. V. ?!oonan F. Rosa S. Varga C. Rossi R. LaGrange T. Dunning P. Shemanski D. Rubenstien R. Stevens D. Fisher A. Thadani J. Joyce J. Kennedy S. Newberry J. T. Beard P. Caranoviski l ~ 4 css /.qS.g,t,,n.ai=a, cS.8/DSI ,.,,,,,,I, css /DSg AD% 1 -= s,,,,,,,,,, _,9 s a,,,, _, ~ ",, @y, _, _ FR o 'h
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MAR 41983 e 9 ENCLOSURE 1 SCRAM BREAKER TEST FREQUENCIES =ess , p l i f I.-
3 + MAR 4 1983 MODEL 1 : Treating the two series scram breakers at Westinghouse plants as a single system in order to achieve a test frequency based on a 3 x 10-5 unavailability' (for ATWS considerations). The following formula is + r used: U= hat where: U = Unavailability of the system e A = Failure rate of the 4 system per year t = Test interval in years I h Sel'ecting A as g and U as 3 x 10-5 and substituting into the above p equation and solving for t yields: t = 15 x 10-3 years, or s 6 days. A was chosen based on one system failure (ATWS) in 250 Westinghouse reactor years. . In arriving at the above test interval of s 6 days, the following should be noted: 'l. The Salem event was counted as only one ATWS event (failure of the single system). Some people may consider the Salem event (s) as two system failures (ATWSs).
- 2. Only Westinghouse plants in the United States were considered in the number of reactor years.
- 3. This is a "best estimate" calculation.
- 4. This model may be grossly oversimplified. Time constraints did not 4
permit a detailed analysis to 'be performed. .y mw, ,,,y- ,..v..... ,..y,, ,-.%,....-.-y,-,. ,,..,-..e.,,_, %_,,.+_.,.+-,e
MODEL 2 : Treating the tw series scram breakers at Westinghouse plants as being totally independent in order to achieve a test frequencey based on a 3x10-5. unavailability of the system (for ATWS considerations) 0= fat 2 where U = Unavailability of system A = Failure rate for individual scram breakers per year t = Test interval in years Selecting A as 9.5 x 10-2 an[Uas3x10-5 and substituting into the above equation and solving for t yields: t = 9.9 x 10-2 years, or S 35 days. A was chosen based on 21 individual breaker failures in 220 Westinghouse reactor years (since 1973). i In arriving at the above test interval of 35 days, the following ~ should be noted. 1. Possible comon cause contributions to the breaker failure rate were not considered. 2. The riumber of Westinghouse scram breaker failures is based upon data obtained from ICSB LER searches. This number is not exact (the actual number is anticipated to be slightly l l higher). 3. Only known breaker failures in Westinghouse plants in the United States since 1973 were considered. 4. This is a "Best Estimate" calculation. 5. This model may be grossly oversimplified. Time constraints did not permit a detailed analysis to be performed. 6. If Westinghouse breaker failure data is used, the test frequency necessary to achieve an unavailability of 3x10-5 is once every 2.5 years. ~.
MAR 4 1983 h G ENCLOSURE 2 PROPOSED TEST FREQUENCY ASSESSMENT i m
y,. MAR 41983 l ENCLOSURE 2 PROPOSED TEST FREQUENCY ASSESSMENT f ICSB believes that the following information should be considerd in the determination of the acceptability of the proposed test frequencies for the undervoltage and shunt trip coil mechanisms for the ' reactor trip breakers at Salem. 1. The shunt trip coil provides a diverse means of tripping the reactor trip breaker which is electrically -independent of the undervoltage trip coil. The undervoltage coil is supplied by a 48 Vdc source and is deenergized to cause a trip whereas the shunt trip coil is supplied by a 125 Vdc source and 1s energized to cause a trip. 2. The shunt trip coil being an energize-to-actuate device is not " fail safe" in that a loss of power will not cause a trip. However, the shunt trip is powered from a highly reliable ' Class 1E battery backed source. 3. Since the shunt trip coil is an energize-to-actuate device, it is' not subject to the constant heating effects that the contin-uously energized undervoltage coil experiences. These heating effects may contribute to the higher failure rate of the under-voltage coil mechanism. ~ 4. The mechanical construction of the shunt trip mechanism is somewhat simpler than that of the undervoltage trip mechanism. The shunt J t trip does not rely on the successful operation of the complex latch-ing mechanism which has been' attributed to be the source of the - majority of failures of the undervoltage trip. 5. The majority of the electrical circuit breakers'used in the higher voltage electrical distribution system have de powered energize-to-actuate shunt trip coil mechanisms. These breakers are used for manual as well as automatic trip functions for load shedding and power switching. Reliability of energize-to-actuate shunt trips in similar applications throughout the nuclear power industry has been very high as demonstrated by the lack of LERs on these devices. 6. PSE&G is revising procedures to require the operator to manually trip the reactor following ir.dication of an automatic reactor trip. Thus, on any trip signal, diverse means will be used to trip the breakers. 7. Over 70% of the known reactor trip breaker failures were caused by undervoltage coil mechanism failures. m..,,
o 8. Most of the concerns relating to the events at Salem on February 22 and 25,1983 are related to the operation of the undervoltage coil (e.g., were the undervoltage trip mechanisms properly lubricated?). During the events at Salem, the shunt trip functioned properly. Based on the above, we conclude that the increased test frequency (from once per 62 days to once per 30 days) for the reactor trip breaker under-voltage coil mechanisms appear to be appropriate and should result in an increase in reactor protection system reliability. On the other hand, however, we do not feel that increasing the test frequency of the shunt trip coil mechanism to once every 7 days is necessary, and may be counter-p roductive. In our judgement, a 30 day test interval for the shunt coil may be more appropriate. m t = 4 4 S 2 ... ~........... -.. -. - --. - -- ---}}