ML19341B088
| ML19341B088 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 01/16/1981 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-2.B.4, TASK-2.E.4.2, TASK-2.K.3.17, TASK-2.K.3.30, TASK-2.K.3.44, TASK-3.D.3.3, TASK-TM LAC-7320, NUDOCS 8101300181 | |
| Download: ML19341B088 (18) | |
Text
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D DA/RYLAND COOFERAT/VE. e o oox sir 2615 EAST AV SOUTH LA Cr40SSE WisCONSW 54601
<cce> usaoco January 16, 1981 In reply, please refer to LAC-7320 DOCKET N0. 50-409 U. S. Nuclear Regulatory Commission ATIN:
Mr. Darrell G. Eisenhut, Director Division of Licensing g
Office of Nuclear Reactor Regulation s
Division of Operating Reactors 4/
Washington, D.C.
20555 d4gggIS 7
SUBJECT:
DAIRYLAND POWER COOPERATIVE
.g v t kg.Si m i-3 LA CROSSE BOILING WATER REACTOR (LACBWR) 5'
[4[Q PROVISIONAL OPERATING LICENSE NO. DPR-45 P_0ST TMI REQUIREMENTS
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REFERENCES:
(1) DPC Letter LAC-7296. Linder to Eisenhut,
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Dated December 31, 1980.
(2) NRC Letter, Eisenhut to All Operating Reactors, Dated October 31, 1980.
(3) DPC Letter LAC-6680, Linder to Denton, Dated December 6, 1979.
(4) DPC Letter LAC-6853, Linder to Dcnton, Dated April 9, 1980.
(5) LACBWR Technical Specifications.
(6) DPC Letter LAC-7266 Linder to Denton, Dated December 5, 1980.
(7) DPC Letter LAC-6705, Linder to Ziemann, Dated December 20, In70 (8)
"La Crosse Boilir Reactor Safeguards Report,"
ACNP 65544, Augu>t 1%/.
(9) " Description of Post-Accident Safeguards Provisions for the LACBWR," ACNP-66564, Amendment No. 29 to CAPR-5, September 1966.
(10)
" Additional Information on the La Crosse Boiling Water Reactor," ACNP-66572, Amendment No. 30 to CAPR-5, October 1966.
(11) " Technical Evaluation, Adequacy of La Crosse Boiling Water Reactor Emergency Core Cooling System," Report SS-942, Gulf United Nuclear Corporation, May 31, 1972.
(12) " Response to Questions by AEC/DL With Regard to Gulf United Report SS-942, Technical Evaluation, Adequacy of La Crosse Boiling Water Reactor Emergency Core Cooling System," Report SS-1075, Gulf United Nuclear Corporation, April 30, 1973.
(13) DPC Letter LAC-6706, Linder to Denton, Dated December 20, 1979.
(14) Licensees Response to Order to Show Cause, Dated January 22. 1930.
O o 8_302000/7/
/
Mr. Darrell G. Eisenhut, Director LAC-7320 Divison of Licensing January 16, 1981 Gentlemen:
0"r response (Reference 1) to your request for information required in NUREG-0737 (Reference 2) indicated our intention to submit certain topics by January 35, 1931.
The information requested is identified as clarification items of NUREG-0737.
I.A.1.1 SHIFT TECHNICAL ADVISOR-DESCRIPTION NO. 4 This item requires a description of the long-term STA program.
It also specifies a comparison be made for reference only to the INP0 training do ument which is Attachment C to Reference (2).
This comparison is to this response.
The long-tern STA program will be based on the existing program which is outlined in References (3) and (4) and Attachment 1 to this response.
An annual raqualification consisting of review of transient and accident conditions, a review of the Safeguards Evaluation section of the LACBWR Safeguards Report for operating authorization (ACNP-65544) which discusses and analyzes the effects of possible accidents, review of Licensee Event Reports for other nuclear plants and training to mitigate core damage will be provided.
Initial training of new STA's will include the existing program plus onsite experience of at least six oths for personnel with previous nuclear plant experience and at least one er for personnel without such experience.
DPC has no specific plans at this time for the eventual phasecut of the STA assignment at LACBWR. We are investigating the availability of nuclear engineering curriculum courses through the University of Wisconsin-Madison and La Crosse for staff personnel.
II.B.4 TRAINING FOR MITIGATING CORE DAMAGE The required training program has been developed and is available for review.
The training will be initiated by April 1,1981.
II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY-DESCRIPTION NO. 5 This item requests a reduction of the containment setpoint pressure for the initiation of isolation of non-essential penetrations to the minimum compatible with normal operating conditions.
The basis for this selection was to be provided by the containment pressure history during normal operation plus 1 psi for instrument drift or fluctuation. The point to be selected should be far enough above the maximum observed (or expected) pressure inside containment during nonnal operation so that inadvertent containment isolation does not occur.
Containment pressures at LACBWR during nonnal operation have reached 4-4.5 psig indicating that a maximum setpoint of 5.5 psig would comply with the guidelines of NUREG-0/3/..
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Ju Mr. Darrell G. Eisechut, Director LAC-7320 Divison of Licensing January 16, 1981 Currently, LACBWR Technical Specifications (Reference 5) permits isolation of the containment building with setpoints for high containment building pressure at 5 psig. The December 1980 calibration of the containment building pressure switches verified that the setpoints were at 4.8 psig.
As a License Amendment would be required to change the existing Technical Specifications to the 5.5 psig value and LACBWR has had no shutdowns due to operational difficulties with the 5 psig limit, no changes will be proposed.
II.K.3.17 REPORT ON OUTAGES OF EMERGENCY CORE COOLING SYSTEMS LICENSEE liEP6iiT7ND PROPMETf fRIT11AL SPECIflTCTil0N CHANGES The requiremer.t is to list all emergency core cooling system outages for five (5) years including:
(1) dates and duration. (2) cause, (3) system or component involved, and (4) corrective action.
The word " outage" was assumed to mean any time the system was not available for automatic operation. to this response covers the 1976-1980 5-year period as required.
In addition to the 4 items requested, a column has been added to note when equipment out of service was required to be operable by LACBWR Technical Specifications (Reference 5).
The total durations for the 5-year period are suanarized as follows for periods when its operability was required.
_ System or Component Total Time-5 Years 1A Emergency Diesel Generator 2.73 Hours IB Emergency Diesel Generator 5.77 Hours Both Diesel Generators 2.47 Hours (Note 1) 1A High Pressure Service Water Diesel 13.85 Hours IB High Pressure Service Water Diesel 12.18 Hours Both High Pressure Service Water Diesels 12.00 Hours (Note 2) 1A High Pressure Core Spray Pump 0
Hours 1B High Pressure Core Spray Pump 0
Hours Both High Pressure Core Spray Pumps 1.68 Hours (Note 3)
Both High Pressure Core Spray and Alternate Core Spray 0
Hours Note 1--Both emergency deisel generators were out only on one occasion in 5 years when availability was required.
This action was intention?1 to correct a control power problem following a scram.
Note 2--Both high pressure service water pumps were out only on one occasfon in 5 years when availability was required.
The reactor was below 17. power.
Note 3--Both high pressure core spray pumps were out only on one occasion in 5 years when availability was required.
This action was intentional *.o correct a control power problem following a scram.
3
Mr. Darrell G. Eisenhut, Director LAC-7320 Divison of Licensing January 16, 1981 This outage list was prepared using Control Room log books ar.d Maintenance Requests for 1976-1980.
The outages are of extremely short duration during periods when operability was required so that no Technical Specification changes are required.
II.K.3.30 REVISED SMALL-BREAK LOSS-0F-COOLANT-ACCIDENT METHODS TO SHOW iMRPLIANCE WITH 10CFR50, APPlNDIX K A conference to review analysis justification is scheduled for early February, 1981.
II.K.3.44 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE-FAILURE TO VERIFY N0 FUEL FAILURE An analysis of anticipated transients combined with equipment failure or operator error, "La Crosse Boiling Water Reactor Review of Transients,"
NES-81A0037, Revision 0, was forwarded to the NRC by Reference (6).
Reference (6) demonstrated that no core damage will result from enticipated transients combined with single-failure.
Because the relief valves are located in a branch of the 10-inch main steam line, a stuck-cpen relief valve would be the equivalent of a steam line break.
The modes for short-term and long-term cooling subsequent to a steam line break were discussed in DPC Response I.(1)(a) of Reference (7).
Since it has been reported that clad temperatures have been calculated to follow the decreasing saturation temperature of the froth in the vessel, it logically follows that no core damage would result from lifting a relief valve.
In addition to the specific transients cited above, it should be noted that References (8) through (12) show that the single-failure proof High Pressure Core Spray System, which includes an individual spray line for each fuel assembly, is capable of cooling the core regardless of reactor water level.
In addition, in the event that water level cannot be quickly re-established and maintained above the core, the plant can be depressurized using the single-failure proof Manual Depressurization System to allow operation of the single-failure proof Alternate Core Spray System. Therefore, no significant core damage will result from any anticipated transient combined with single-failure of equipment.
II.D.3.3 IMPROVED INPLANT I0 DINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS The equipment for post-accident analysis of iodine was described in Reference (13) and Reference (14).
Training of Health and Safety Technicians on the use of this single channel analysis has been completed.
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Mr. Darrell G. Eisenhut Director LAC-7320 Divison of Licensing January 16, 1981 If there are any questions regarding this letter, please let us know.
Very truly yours, DA PYLAND POWER COOPERATIVE
.WW$L/
lW rank Linder, General Manatjer FL:JDP: abs CC:
J. G. Keppler Reg. Dir., NRC-DRO III NRC Resident Inspectors l
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1 COMPARISON OF LACBWR STA'S TO THE PROGRAM DESCRIPTION 0F IHE INSTIfiTTE 0' NIICLTAR POWER OPERATITriS
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INP0 SECTION 5 - GENERAL EDUCATION AND EXPERIENCE Experience LACBWR STA's comply with the experience requirements listed with the exception that six months or greater experience at the station is considered a sufficient minimum.
Absence _s No absences from duties of significant duration have occurred and a specific policy has not been determined.
I_NPO SECTION 6 - EDUCATION AND TRAINING REQUIREMENTS 6.1.1 Prerequisites Beyond High School Mathematics--------------90 Hours (6 Credits)
Chemistry----------------30 Hours (2 Credits)
Physics-----------------150 Hours (10 Credits)
As STA's at LACBWR are engineers or supervisory personnel of long experience (Ref.1) the majority of the incumbents would ha*/c completed these course exposures as part of pre-engineering or science curriculum.
6.1.2 College Level Fundamental Education Mathematics--------------90 Hours (6 Credits)
Reactor Theory----------100 Hours (6-2/3 Credits)
Reactor Chemistry--------30 Hours (2 Credits)
Nuclear Materials--------40 Hours (2-2/3 Credits)
Thermal Science---------120 Hours (8 Credits)
Electrical Science-------60 Hours (4 Credits) 14uclear Instrumentation and Control------------40 Hours (2-2/3 Credi ts)
Nuclear Radiation Protection and Health Physics----------------40 Hours (2-2/3 Credits)
The STA's with engineering degrees will have portions of this section. As it is not a standard curriculum for any type of engineer, no individual would necessarily have this credit breakdown.
Individuals with nuclear engineering degrees, generally have the topics in mathematics, reactor theory, nuclear materials, and thermal sciences.
Those with degrees in other fields such as electrical, mechanical, or radiation protection engineering would have the mathematics plus the topic most related to their degree of specialization.
6.2 Appli.e_d Fundamentals - Plant Specific Topics in Reactor Technology, Chemistry /
Corrosion Control, Instrumentation, Materials, a nd Th e rmal Cyc l e s------------------------------120 Hours Plant specific training at LACBWR has required about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> in plant specifics beyond the individual's background experience. As only individuals of long experience at LACBWR or newer employees of significant nuclear background were utilized, this training is sufficient.
6.3 Mana_gement/ Supervisory Skills Various Management Topics------------------------40 Hours The assigned STA's are members of LACBWR management and professional staff.
No specific training is provided as part of the STA Program.
6.4 Plant Systems Speci fic Training on Plant Systems--------------200 Hours The assigned STA's of long tenure at LACBWR received an average of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of system training.
Additional training was provided for individuals of lesser tenure, all of whom had significant previous nuclear background.
It is estimated ; hat 50-60 contact hours would be required for college graduate engineers without previous plant exposure to familiarize them with plant systems.
The training wili increase approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> with the addition of Core Damage Mitigation training.
6.5 Administrative Controls Administrative Control Procedurc -
Familiarization in Selected Areas----------------80 Hours All assigned STA's are members of the LACBWR Operations Review Canmittee and as such, review most of the items listed in Appendix C of Reference (2).
They also receive the required training annually for security and 10 CFR 20.
It is estimated that annual time caanitment by the STA's exceeds the 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> recommended by the INP0 guide for initial STA training.
6.6 Genera _l,0perating Procedures Various Standard Operating Procedure Topics------30 Hours These topics do not fall specifically within the STA accident assessment or operation assessment function as they are more closely related to routine power operation. The assigned STA does have a constant involvement in the review of these procedures as a member of the Operations Review Committee, however, specific STA training in this area is not prescribed.
6.7 _ Transient / Accident Analysis and_ Emergency Procedures Transient and Accident Analysis Plant Abnormal and Emergency Procedures-------------------------30 Hours The LACBWR STA training in these areas is 8 to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> on the average.
It will increase by approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with the addition of Core Damage Mitigation training.
6.8 Simulato_r Training There exists no simulator module for this facility.
6.9 Annual Requal_ification Review of Transients and Accidents, Rev i ew o f In d u st ry LER---------------------------4 0 Ho u rs The annual requalification time will include a review of the Safeguards Evaluation section of the LACBWR Safeguards Report for Operating Authorization (ACNP-65544) which discusses and analyzes the effects of possible accidents, general transient and accident analysis, and individual review of industry LER's by each SlA.
It is estimated approximately 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> annually will be involved in this activity.
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40 Hou rs There is no simulator for the LACBWR facility.
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ECCS EQUIPMENT OUTAGES.- 1976-1980 j
OPERABILITY REQ'D DURING i
2 DATE DURATION
.AFFECTED EQUIPMENT PLANT CONDITION REASON /CAUSE CORRECTIVE ACTION _TAf.Eji i
2/27/76 1HR 44MN-1A HIGH PRESSURE N3 PREVENTIVE MAINTENANCE (PM) CHANGED OIL, AIR i
SERVICE WATER FILTERS, ETC.
I (HPSW) DIESEL i
2/27/76 1HR 1B HPSW DIESEL NO PM
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3/11/76 7HR 7 MN 1A & IB HIGH PRES-NO PM 0F SUCTION FILTER SURE CORE SPRAY (HPCS) PUMPS s
4/ 1/76 5 HR 55 MN 1A EMERGENCY NO PM DIESEL GENERATOR L
i-(EDG) l 4/ 5/76' 19 MN 1A EDG NO INSTALLATION OF IB EDG l
l 4/ 8/76 29 HR 5 MN 1B HPSW NO LEAK ON ENGINE HEATER INSTALLED NEW HEATER
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4/22/76 6 DAYS 1.25 HR 1B HPCS PUMP NO REWIRED POWER SUPPLY (DUE TO INSTALLATION ON NEW EDG)
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4/30/76 6HR'15 MN 1B HPCS PUMP NO SPRAY FOR ONE OF THE UNPLUGGED LINE PLUNGERS PLUGGED 5/11/76 16 DAYS 1A EDG NO LEAK ON EXHAUST PIPING INSTALLED NEW SEALS IN MANIFOLD EXPANSIO'1 JOINT t
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5/13/76 30 MN IB HPCS PUMP NO FAILED TC > TART ON LOW ADJUSTED AND CLEANED WATER SIGl.AL CONTACTS 5/18/76 20 DAYS 19H 35M 1A HPSW DIESEL NO GENERAL PUMP INSPECTION i'
6/ 7/76 24 MN IB HPCS PUMP NO BREAKER RACKED OUT 6/16/76 30 MN 1A EDG NO CHANGE BATTERIES 6/22/76 '42 MN 1B EDG NO INSTALL OIL PRESSURE SWITCH
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ECCS EQUIPMENT OUTAGES, 1976-1980 OPERABILITY REQ'D DURING DATE DURATION AFFECTED EQUIPMENT PLANT CONDITION REASON /CAUSE CORRECTIVE ACTION tan N 7/ 9/76 5 HR 29 MN 1A EDG NO OIL LEAK ON TURBOCHARGER TEMPORARILY REPLACED BEARING RETURN TO SUMP HOSE W/TYGON TUBING UNTIL RUBBER HOSE AVAILABLE 8/ 2/76 2 HR 20 MN 1B HPSW NO PM 8/ 2/76 1 HR 53 MN 1A HPSW NO PM 8/ 4/76
< 8 HR 1A EDG NO PM-CHANGE OIL & FILTER 9/ 1/76.
9 MN 1A EDG YES LO FILTER GASKET LEAK REPLACED GASKET 9/15/76 3 HR 25 MN 1B'EDG YES TEST LOAD ALTERNATELY RESET MANUAL EMERG.
LOADED A UNLOADED DURING SHUTDOWN RESET MONTHLY TESTS 11/ 5/76 50 MN 1B HPCS PUMP NO FAILED TO START ON B0RON RECONNECTED WIRE WHICH INJECT SIGNAL DURING HAD COME LOOSE, CHE:nEd TEST OTHER TERMINALS 11/16/76 17 MN 1A EDG N0 ELECTRICAL RESISTANCE MEASURING 2/28/77 29 MN 1B HPSW YES COOLING SOLEN 0ID VALVE CLEANED SAND FROM STICKS SHUT ACTUATOR PISTON 5/12/77 20 HR 7 MN IB HPSW.
NO LEAK ON COOLING WATER INSTALLED NEW BUSH '. 4 SUPPLY LINE PIECE OF PIPE
'5/20/77 5 DAYS 2211 35M 18 EDG NO LEAK ON COOLING WATER REWELDED BAD WELD ON LINE PIPE ELBOW 5/27/77
< 3 DAYS 1B EDG NO PM 5/30/77 3 DAYS 35 MN IB EDG NO PM
ECCS EQUIPMENT OUTAGES, 1976-1980 OPERABILITY REQ'D DURING DATE DURATION AFFECTED EQUIPMENT PLANT CONDITION REASON /CAUSE CORRECTIVE ACTION TAKEN 7/11/77 35 MN 1B EDG NO TEST OF CO2 SYSTEM S/26/77 3 HR 15 MN 1A EDG NO PM 8/26/77 1 HR 15 MN 13 EDG NO PM 9/ 9/77 2 HR 51 MN 1A EDG NO PM 10/10/77 2 HR 27 MN 1A EDG NO REPLACE WORK COOLING WATER REPLACED HOSE HOSE TO RADIATOR 11/ 2/77 6 itR 3 MN 1B HPSW NO COOLANT SYSTEM FLUSil IP/ 1/77 40 MN 1B HPSW NO OVERHEATED DUE TO FAILURE CLEANED SOLEN 0ID VALVE OF COOLING SOLEN 0ID TO OPERATE
!?/20/77 2 HR 1A EDG NO TIME DELAY ON LOW LUBE OIL TIME DELAY SAT. ALSO PRESSURE TRIP IN STARTING INSTALLED NEW LENGTH OF CIRCUIT HOSE AND CLAMPS ON WATER PUMP DISCHARGE Y
12/21/77 2 HR 1A EDG NO AUTO SHUTDOWN SOLEN 0ID REPLACED 7,UTO SHUTDOWN REPLACED S0LEN0ID VALVE C==D g IP/22/77 1 HR 25 MN 1A EDG NO CONTINUATION OF WO"<K ALSO REPAIRED FUEL LEAK COMMENCED 12/21 12/22/77 50 MN 1A EDG NO CONTINUATION OF WORK ALSO REPAIRED FUEL LEAK l
COM:ENCED 12/21 1/ 4/78 1 HR 14 MN 1A EDG NO RELOCATED FUEL SHUT 0FF 1g VALVE PER FACILITY CHANGE 1/ 4/78 56 HR 1A HPSW NO CRACKED DIESEL EXHAUST INSTALLED NEW EXPANSION u
EXPANSION JOINT JOINT & FLANGES
Ei.jl5_ EQUIPMENT GUTAGES, 1976-1980 OPERABILITY REQ'D DURING DURATION AFFECTED EQUIPMENT PLANT CONDITION REASON /CAUSE CORRECTIVE ACTION TAKEN UATE__
1/12/73 4 HR 1A EUG NO FUEL LEAK AT INLET TO FUEL TIGHTENED FITTING OIL SHUT 0FF VALVE SOLEN 0ID VALVE 2/20/78 2 HR 1B HPSW NO MAINTENANCE - PROBABLY PREVENTIVE 3/ 3/78 40 MN 1B HPSW NO DIESEL NOT RECEIVING REPLACED COOLING WATER SUFFICIENT COOLING FLOW SOLEN 0ID 3/15/73 6 HR 16 MN 1B HPSW YES EXHAUST LEAK REPLACED EXPANSION JOINT 3/21/78 12 HR BOTH HPSW PUMPS YES BOTH HPSW PUMP SWITCHES RETURN TO AUTO - RX OFF LESS THAN 5% POWER 1/23/78 8 DAYS 6.5 HR ACS NO PLANT SHUTDOWN 5/ 1/78 10 MN 1B EDG NO ELECTRICAL RESISTANCE MEASURING
,g 5/ ?/78 45 MN 1A EDG NO ELECTRICAL RESISTANCE b
MEASURING Ei@
5/ 7/78 46 HR 40 MN ACS NO PLANT SHUTDOWN g 5/29/18 20 HR 36 MN ACS NO PLANT SHl!IDOWN c===2
@ 6/20/78 15 MN 1A HPSW YES AIR IN FUEL OIL SYSTEM VENTED & REFILLED FUEL C=a OIL PRIMING CHAM 3ER g 8/11/78 12 MN 1A EDG YES CHANGE OIL 10/19/78 95 MN 1A EDG NO PM ON BATTERIES & DIESEL
& CHANGE CONTROL SWITCH TO ADD "NOT IN AUT0" ALARM
E Cf15 EQUIPMENT OUTAGES, 1976-1980 OPERABILITY REQ'D DURING
_DATE DURATION AFFECTED EQUIPMENT PLANT CONDITION REASON /CAUSE CARRECTIVE ACTION TAKEN 10/25/78 273 HR 45 MN 1A HPCS PUM?
NO PUMP LEAKING THRU INTERtlA'.LY 10/26/78 128 HR 15 MN 1A EDG NO CHECKED FOR FUEL OIL LEAKS REPLACED #2 INJECT 0P, N0ZZLE 11/22/78 61HR 10 MN ACS NO PLANT SHUTDOWN 11/29/78 2 HR 35 MN 1B HPSW DIESEL YES INSTALLED NEW AIR CONTROL VALVE FOR COOLING WATER 11/29/78 APPROX. 3 HR 1A HPS2 DIESEL YES INSTALLED NEW AIR CONTROL VALVE FOR COOLING WATER 12/15/78 5 MN 1B EDG YES FAILURE TO START ON RETESTED SAT. COULD NOT MONTHLY TEST - CAUSE DUPLICATE PROBLEM UNKNOWN 12/13/78 4 MN 1B EDG YES TRIPPED ON OVERSPEED RESET TRIP SETPOINT WHILE TESTING 1/13/79 215 HR 53 MM ACS NO PLANT SHUTDOWN g 1/16/79 8 MN 1B EDG NO ELECTRICAL RESISTANCE TESTING g
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')3/26/79 1 MONTH 29 DAYS ACS NO PLANT SHUTDOWN M
9 HOURS O 4/12/79 26 DAYS 18H 54M ECCS NO PLANT SHUTDOWN 5/ 3/79 31 HR 36 MN 1A EDG NO FUEL OIL CONTAMINATION IN REPLACED ENGINE HEATER Q
LUBE OIL HOSE N 5/10/79 6 HR 14 MN 1A HPCS NO 18-MONTH TEST 2EED 5/10/79 4 HR 16 MN 1B HPCS NO 18-MONTH TEST b
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B B
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W W
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0 F
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F A
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A A
A A
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C B
B A
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9 A
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1 1
1 1
1 A
1 1
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H 1
6 7
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M 3
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4 S
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N N
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N Y
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R RM R
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R R
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9 9
9 9
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9 9
9 9
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7 7
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7 7
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/
7
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q E
ECCS' EQUIPMENT OUTAGES, 1976-1980 OPERABILITY REQ'O DURING
_ DATE _
DURATION AFFECTED EQUIPMENT PLANT CONDITION REASON /CAUSE CORRECTIVE ACTION TAKEN l
4/11/80-25 MN +
HPCS PUMPS NO CHECK SUCTION STRAINER TIGHTENED C0VER 1 HR 42 MN LEAKS 4/13/80 35 MN 18.EDG NO WIRING MODIFICATIONS 1
4/19/80 11 MN 1A HPCS PUMP NO PM ON BREAKER i
4/19/80 4 MN 1B HPCS PUMP NO PM ON BREAKER g.
4/21/80
'.37 MN 1A HPCS PUMP, NO TMI MODIFICATIONS IA HPSW
!4/21/30 42 MN 1A HPCS PUMP, NO TMI MODIFICATIONS 1A HPSW 4/2t/80-S HR 12 MN 1A HPSW NO.
PM 4/24/80 4 HR 42 MN 1B EDG NO PM
'6/25/80 1 HR 7 MN 1A EDG NO PM 6/21/80 10 MN IB EDG N0
'~ % 8/10/80 5 DAYS '6.'S HR ACS NO.
PLANT SHUTDOWN
@ 8/14/80 g
2 HR 12 MN ~
1B HPSW NO CHANGE OIL N 8/19/30
.< 5 MN 1B HPSW YES FAILED TO START FROM REPLACED BATTERIES CONTROL ROOM
~
- 2. @
3/20/80 4 MN 1B HPSW YES CONNECT NEW BATTERY-h3/20/80 8 MN 1A HPSW YES
. CONNECT NEW BATTERY
/26/80
- 7. DAYS ACS N0 PLANT SHUTDOWN
?
/'7/80' 2 HR-12 MN 1B EDG YES CHANGED STARTING BATTERIES
LECC_5 E_QUIPMENT OUTAGES, 1976-1980 OPERABILITY REQ'D DURING DATE DURATION AFFECTED EQUIPMENT PLANT CONDITION REASON /CAUSE CORRECTIVE ACTION TAKEN 11/12/80 1 MONTH, 1 DA7, ACS NO PLANT SHUIDOWN 4 HR 50 MN 11/15/80 7 HR 2 MN 1B ECG N0 18-MONTH BATTERY TEST 11/16/80 1 HR 57 MN' HPCS NO WORK IN REACTOR CAVITY 11/24/80
< 1 DAY HPCS N0 11/25/80-3 HR 39 MN HPCS NO
'12/.4/80
_8 DAYS, 8 HR 18 HPSW NO 5-YEAR PM 12/ 9/80 1 HR 18 EDG NO ELECTRICAL RESISTANCE TESTING 12/11/80' 1 DAY 7 HR:25 MN.
1A HPSW N0 5-YEAR PM 12/12/80-
_7 DAYS 1A & IB HPCS PUMPS NO ILRT 12/16/80 7 DAYS'
.ACS NO FACILITY CHANGE TO ADD EMERGENCY SERVICE WATER SUPPLY SYSTEM 6%)
Q 12/25/80 16 HR 10 MN
.ACS NO FACILITY CHANGE TO ADD EMERGENCY SERVICE WATER g
SUPPLY SYSTEM 55'G) 12/29/80 APPROX. 8 DAYS ACS
.N0 FACILITY CHANGE TO ADD D
EMERGENCY SERVICE WATER LE'p)
SUPPLY SYSTEM IEFS 552 2 Essu 7_,.