ML19269C905
| ML19269C905 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 02/12/1979 |
| From: | Sjogren R Maine Yankee |
| To: | |
| Shared Package | |
| ML19269C904 | List: |
| References | |
| NUDOCS 7902200057 | |
| Download: ML19269C905 (28) | |
Text
1 l
MAINE YANKEE NUCLEAR POWlR STATION MONTHLY STATISTICAL REPORT 79-01 FOR THE MONTH OF JANUARY, 1979 7902200067
i OPERATING DATA REPORT 50-309 DOCKET NO.
DATE 790212 CO.\\1PLETED BY R.M.
Sjogren TELEPilONE (617)366-9011 x2281 OPERATING STATUS Maine Yankee N tes
- 1. Unit Name:
January 1979
- 2. Reporting Period:
2630
- 3. Licensed Thermal Power (31Wt):
864
- 4. Nameplate Rating (Gross.\\lWe):
825
- 5. Design Electrical Rating (Net 31We):
850
- 6. 31aximum Dependable Capacity (Gross SIWE):
810
- 7. Slaximum Dependable Capacity tNet SlWe).
S. If Changes Occur in Capacity Ratings titems Number 3 Through 7) Since Last Report. Give Reasons:
- 9. Power Lesel To which Restricted. lf Any (Net 51We):
864 MWe approximately 97%
Because of maximum allowable steam flow through the
- 10. Reasons For Restrictions. lf Any:
T.P.
turbine.
k This 31onth Yr.-to.Date Cumulative 744 744
- 11. Hours In Reporting Period
- 12. Number Of flours Reactor Was Critical 735.11 7.35. 11 45,778.82
- 13. Reactor Reserve Shuvown llours 0
0 0
- 14. Ilours Generator On.Line 729.68 729.68 44;252.85
- 15. Unit Reserve Shutdown flours 0
0 0
- 16. Gross Thermal Energy Generated (5tWil) 1,777,359 1,777,359 93,114,741
- 17. Gross Electrical Energy Generated (31Wil) 600,140,. _,
600,140 30,467,950 IS. Net Electrical Energy Generated (SIWil) 572.766 572,766 28,860,962
- 19. Unit Sersice Factor 98.0 93.0 81,0
- 20. Unit Asai! ability Factor 98.0 98.0 81.0 95.0 95.0 69.2
- 21. Unit Capacity Factor (Using SIDC Net) 93.3 93.3 66.7
- 22. Unit Capacity Factor (Using DER Net) 1.9 1.9 3.7
- 23. Unit Forced Outage Rate
- 24. Shutdowns Scheduled Oser Next 6.ilonths (Type. Date.and Duration of Each):
- 25. If Shut Down At End Of Report Period. Estimated Date of Startup:
- 26. Units In Test Status (Prior to Commercial Operation):
Forecast Achiesed INITIA L CRITICALITY INITIAL ELECTRICITY CO\\l51ERCIAL OPER ATION (9/77)
t AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-309 Maine Yankee UNIT DATE 790212 CONIPLETED BY R. M. Siogren W.EPHONE (617)366-9011 X2281 MONTil January 1979 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (M We-Net )
(MWe. Net )
824 772 1
37 779 2
823 33 800 3
298 g9 609 792 4
20 581 820 5
21 636 819 3
792 823 7
23 760 820 8
3 822 9
820 25 820 10 818 26 795 11 792 27 819 818 12 3
13 819 797 39 797 777 34 30 822 783 15 31 16 819 INSTRUCTIONS On this format. list the awrag: daily unit power lesel in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.
(0!77 )
i
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Page 1 of 3 UNITS!!UlDOWNS AND POWER REDUCTIONS DOCKET NO.
50-309 UNIT N AM E Maine Yankee DATE 790212 REPORT MONTil Jariuary 1979 COMPLET ED itY R. M. Sincren TELErilONE (617)366-9011
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a 1-79-4 1-3-79 F
10.70 A
3 NA IA INSTRU-X Electrical spike on RPS temperature channels resulting in a TM/LP trip.
2-79-4 1-5-79 F
3.62 A
3 NA IA INSTRU-X Electrical spike on RPS temperature channels resulting in a TM/LP trip.
L.R.
1-6-79 S
29.50 B
1 NA HF llTEXCll-D ECT of selected condenser waterbox to 757 1-7-79 tubes.
L.R.
1-8-79 F
11.42 A
1 NA IIF llTEXCH-D Condenser tube Icaks in the "C"
& "D" to 80%
waterboxes. Plugged 1 tube in each.
L.R.
1-11-79 F
7.25 A
1 NA IIF llTEXCll-D Condenser tube leak in the "A" waterbox to 80%
- Tube plugged.
L.R.
1-14-79 F
7.25 A
1 NA IIF llTEXCII-D Condenser tube leaks in the "D" o 80%
waterbox - 3 tubes plugged.
I 2
3 4
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Page 2 of 3 UNIT SilulDOWNS AND POWER REDUCTIONS DOCKET NO.
50-309
~
UNIT NANtE Maine Yankee DAIE 740212 COMPLET ED lty R.
M_
%,.,, n REPORT MONTil Januarv 1979 T ELLPilONE (617)366-9011 x.2281 5
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6 L.R.
1-17-79 F
12.00 A
1 NA IIF llTEXCll-D Condenser tube leak in the "D" waterbox to 80%
- 1 tube plugged.
L.R.
1-18-79 F
10.25 A
1 NA IlF llTEXCH-D Apparen'_ cube leak in the "D" waterbox to 802
- no leak could be found.
L.R.
1-19-79 F
7.75 A
1 NA IIF llTEXCll-D Condenser tube leak in the "D" waterbox to 801
- 1 tube plugged.
L.R.
1-20-79 F
7.50 A
1 NA IIF llTEXCll-D Condenser tube leak in the "D" waterbox to 802
- 3 tubes plugged.
L.R.
1-27-79 F
8.42 A
1 NA IIF IITEXCll-D Condenser tube Icak in the "D" waterbox to 80%
- 1 tube plugged.
L.R.
1-29-79 F
5.50 A
1 NA IIP llTEXCH-D Condenser tube leak in the "D" waterbox to 80%
- I tube plugged.
I 2
3 4
1:. I o r t eil ite.ison :
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I Page 3 of 3 UNIT S!!UlDOWNS AND POWER REDUCTIONS DOCKET NO.
50-309 UNil' N AM E Maine Yankee 790212 DALE January 1979 COMPl.ET ED lty R. M. Sjogren REl' ORT MONTl!
TELLPilONE (617)366-9011 x.2281 O
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L. R'.
1-30-79 F
10.00 A
1 NA IIF llTEXCll-D Condenser tube leak in the "A" waterbox to 80%
- 1 tube plugged.
L.R.
1-31-79 F
9.97 A
1 NA IIF llTEXCll-D Condenser tube leak in the "A" waterbo>
to 80%
- 1 tube plugged.
I I
2 3
4 l'. l'ori cil it e.non :
M e t h o,l.
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DOCKET NO.
50-309 UNIT Maine Yankee DATE 790212 COMPLETED BY R. M. Siocren TELEPIf0Nd (617)366-9011 x.2281 REPORT MONTli
.Ta n u a ry 1979 SUSNARY OF OPERATING EXPERIENCES At the beginning of the month the plant was operating at S 97%
reactor power limited by the 864 MWe maximum load permitted on the low pressure turbine blading.
On January 3rd and again on the 5th, electrical spikes on the station's grounding network resulted in two plant trips. The change in the ground potential momentarily spikes the temperature inputs to the TM/LP circuitry resulting in a trip.
The apparent cause was due to a summation of electrical noise which, due to the narrow margin to trip in the TM/LP channels, wa sufficient to spike the indication temperature signals to the trip setpoint.
Corrective action has been taken to reduce the electrical noise and it's not expected to cause further problems.
During the power escalation following the trip on 1/5/79, load was held at 75% for S 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> to permit Eddy Current Testing of a selected group of condenser tubes. Additionally, one condenser tube which had developed a leak was removed, while at power, to determine its failure mode.
For the remainder of the month the plant operated at essentially full load (m 97%) except the 10 load reductions to 80% to locate and plug leaking condenser tubes.
DESIGN PACKAGES COMPLETED DURING 19eb Plant Alterations (and Modifications)
PM 9-77 Waste Neutralization Tank Modification PA 12-77 Plant Fire Alarm System and Gaitronics Installation PA 1-78 Addition of Blowdown Demineralizer System PA 2-78 Aux. Charging Pump Cable Reroute PA 3-78 Addition of M.C.B.
Indication for E.H.C.
PA 10-78 Relocation of Containment Low RMS Channel Detection PA 11-78 Service Water Vent to Vacuum Priming Tank PA 13-78 Removal of Solenoid Valves From M.Y. FRV Control System PA 15-78 Forced Circulation Thru E-37 (PWST Thermosyphon Htr.)
Plant Design Change Requests PDCR 2-77 Insulate PCC & SCC Motor Bearing Capsules PDCR l-78 Installation of Check Valves in HPSI Suction Line PDCR 3-78 CIS Closure of Control Room Ventilation MOVs PDCR 4-78 Steam Generator Modification PDCR 6-78 Low Steam Generator Pressure Bypass Bistable Mod.
Engineering Desien Change Requests EDCR 75-11 Additional Filtration EDCR 76-17 TM/LP Trip Csiculator Mod.
EDCR 77-23 Diesel Generator Alcu. Modifications EDCR 78-8 Security Lighting Power Supply EDCR 78-10 Control Rod Modification EDCR 78-13 CEA Plug Assembly EDCR 78-15 CEDMTime Delay Latching Relay
PM 9-77 WASTE NEUTRALIZATION TANK MODIFICATION Due to inadequate corrosion resistance and inadequate volume, the 10,000 gallon placite lined carbon steel waste neutralization tank was replaced with a 22,500 gallon fiberglass tank enclosed in a new addition attached to the south side of the turbine building.
This modification does not constitute an unreviewed safety question as defined by 10CFR50.59 (a) (2).
PA 12-77 PLANT FIRE ALARM SYSTEM AND CAITRONICS INSTALLATION The first portion of this alteration involved the addition of a fire alarm switch to the B section of the main control board. The system uses the existing tone generator for the plant emergency evacuation alarm slightly altered to obtain interrupted tone that is distinctly different from the evacuation alarm. The tone is broadcast over the Femco system.
The second portion of the alteration involved the addition of Gaitronics phones in the following areas: on the lower level PAB mounted on the east wall of the degasifier area just south of the locked gate, outside the entrance to the nressurizer relief valve cubicle, and a hazardous duty unit located in the southeast corner of the Gas House.
This alteration does not constitute an unreviewed safety question as defined by 10CFR50.59 (a) (2).
PA 1-78 ADDITION OF BLONDOWN DEMINERALIZER SYSTEM In order to reuse the steam generator blowdown water (reducing demand on the make-up demineralizers), a steam generator blowdown demineralizer system was added.
Condensate from the steam generator blowdown tank is piped through the existing cooler to a new blowdown transfer pump which boosts the condensate pressure.
Condensate is pumped through the existing blowdown tank drain piping to a point just prior to the existing connection to the service water system. A new branch from this point carries the blowdown condensate to the demineralizer system located in the water treatment area.
Before passing through the demineralizer, the condensate is filtered through a wound fabric prefilter with replaceable cartridge elements. The demineralized condensate is then run through a post-filter, which is identical to the pre-filter, and returned to the condenser hotwell.
This alteration does not constitute an unreviewed safety question as defined by 10CFR50,59 (a) (2).
9 a
PA 2-78 AUXILIARY CHARGING PUMP CABLE REROUTE To preclude the possibility of a postulated fire in the cable vault from damaging the power supply cables to all reactor coolant system charging pumps. The cables for the Auxiliary Charging Pump were re-routed away from the cable vault.
MCC-9B was selected as the new power source since its feeder comes out of 480 volt bus 9 in the unprotected switchgear room, through the floor to the un-protected cable tray room and into the primary auxiliary building through a wall sleeve without passing through the cable vault.
The power and control cable runs from MCC-9B to the primary auxiliary building emergency shutdown panel and the local pump control enclosure primarily in conduit. The existing control and instrument cable that runs from the primary auxiliary building to the control room was lef t as originally installed.
The postulated cable vault fire could damage this cable but the auxiliary charging pump can be started locally from the PAB emergency panel and can be speed controlled manually at the pump.
This alteration does not constitute an unreviewed safety question as defined by 10CFR50.59 (a) (2).
'PA 3-78 ADDITION OF MAIN CONTROL BOARD INDICATION FOR ELECTRO-HYDRAULIC CONTROL During certain plant operations it is necessary that the control room operators monitor various turbine control valve indication.
Previously, these indications were only available at the Electro-Hydraulic Control System (EHC) cabinets.
This alteration provides the operators with tur'oine control valve indications and an indication selector switch located on the Display Panel Assembly of the EHC system on the Main Control Board.
Parallel connections were run from the Turbine Control Test Point Selector Switch on the Test and Calibration Panel to the new remote selector switch.
While future indicator positions have been provided, the system will currently monitor valve position limit, governor valve common, governor valve manual, and governor valve auto.
This alteration does not constitute an unreviewed safety question as defined by 10 CFR 50-59 (a) (2).
PA 10-78 RELOCATION OF CONTAINMENT LOW RMS CHANNEL DETFrTOR In order to enabic fuel movement in the upender area of the refueling cavity without setting off the Containment Low channel high alarm during normal refueling operations, the Containment Low channel detector and remote indicator assembly was moved from the west side to the east side of the refueling cavity.
This alteration does not constitute an unreviewed safety question as defined by 10 CFR 50.59 (a) (2).
PA 11-78 SERVICE WATER VENT TO VACUUM PRIMING TANK Because several different systems empty into the service water system, interaction between the systems had been causing improper draining of the turbine hall sump water settling tank.
A vent line was therefore run from a spare instrument connection near the Service water outlet top hat to a spare instrument drain connection on the vacuum priming tank.
The vent removes non-condensibles f rom the liquids entering the top hat.hus allowing the turbine hall sump water settling tank to drain properly.
This alteration does not constitute an unreviewed safety question as defined by 10CFR 50-59 (a) (2).
'PA 13-78 REMOVAL OF SOLEN 0ID VALVES FROM THE FEED'4ATER REGULATOR VALVE CONTROL SYSTEM Maine Yankee's Feedwater Regulator Valves have had a history of problems which have lead to, among other events, a plant trip in January of 1976.
During December 1976, the solenoid valves which cause the FRV valves to " fail as is" (open), were removed.
It was determined that stem leakage out of the bottom of the operator cylinders caused the pressure on the bottom of the piston operator to decrease forcing the FRV to close.
This defeated the purpose of the " fail as is" solenoid valve system.
The Systems Engineering Group in Westboro, Ma. conducted an independent evaluation on the removal of the solenoid valves and determined that the alteration does not change the FRV response in the accident in which its operation is required, nor will it significantly increase the probability of a feed system waterhammer event.
They also determined that the alteration did not affect the Plants FSAR or Tech. Spec nor did it constitute an unreviewed safety question as defined by 10 CRF 50.59 (a) (2).
PA 15-78 FORCED CIRCULATIO:: THROUGil E-37 (PWST THERM 0 SIPHON HEATER)
The thermosiphon heater for the Primary Water Storage Tank (PWST) was designed for natural circulation.
Because of this slow, natural circulation, part of the heated water turns to steam before 2ntering the PWST causing considerable water hammering.
This alteration solves the problem by forcing circulation through the thermo-siphon heater and rerouting the PWST recirculation from the two Primary Water Transfer Pumps to the thermosiphon heater inlet.
The higher flow rates through the heater do not allow the water to turn to steam.
This alteration does not constitute an unreviewed safety question as defined by 10 CFR 50 59 (a) (2).
PDCR 2-77 INSULATE PCC & SCC MOTOR BEARING CAPSULES Because of a number of bearing failures in the past several years on the Pr s. nary Component Cooling System pump motors and the Secondary Component Cooling System pump motors, the manufacturer was consulted.
The service engineer for the man-ufacturer suspected that the problem was caused by AC circulating shaft currents which are caused by motor magnetic circuit imbalances. These suspicions were confirmed.
The solution to the problem was to insulate the bearing capsules on the end of the motor opposite the coupling, which has been done.
This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 (a) (2).
PDCR 1-78 ADDITION OF CHECK VALVES IN HPSI SUCTION LINE The Maine Yankee Steam Line Break Analysis has shown that, for cycle 4 and beyond, a boron concentration of 17?0 ppm is required in the High Pressure Safety Injection (HPSI) line supplying the HPSI pumps. To assure this condition check valves were installed just upstream of the tee leading from the Volume Control Tank (VCT). The check valves allow flow to the HPSI pumps in the event of a safety injection actuation signal, but prevent any dilution of the HPSI suction line from the lower boron concentration of the reactor coolant system.
Just upstream of each check valve, a 3/4 inch line was installed off from the HPSI suction pipe. Attached to this is a 1/2 inch valve to be used as a chemistry sample point and a 3/4 inch valve which can be used to allow flow from the refueling water storage tank down the HPSI suction line in the event that the chemistry sample shows an unsatisfactory boron concentration.
Yankee Atomic Electric Co. performed a seismic evaluation on the modification and concluded that the piping was structurally adequate to withstand all design loading conditions, and that no support modifications or additions were necessary.
This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 (a) (2).
a PDCR 3-78 CIS CLOSURE OF CONTROL ROOM VENT MOV'S A NRC review of the radiological consequences to the control room operators as a result of the design basis loss of coolant accident was completed as part of the research necessary to grant Maine Yankee a license uprate to 2630 MWt.
The NRC estimated that the thyroid dose to the operators could potentially exceed the criteria of General Design Criterion 19.
Maine Yankee has therefore installed automatic isolation of the control room ventilation as a result of this estinate.
Separation criteria were adhered to during the original installation of these valves so the original power and control cables were determined to be adequate and since each emergency bus powers an intake and exhaust MOV, no power supply changes were made.
Automatic closure of the MOV's was thus accomplished by inserting a normally open contract in the close circuit of the MOV's and closing the contact on a Containment Isolation Signal (CIS).
This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 (a) (2).
4
e a
e PDCR 4-78 STEAM GENERATOR MODIFICATION Tube denting is a phenomenon that occurs in steam generators with drilled support plates.
Chlorides may hide out between the tubes and support plates causing stress build-up in the support plates. The stress is uniform throughout the plate except where no tubes exist, as on the rim, and causes the plate to expand in the horizontal plane. The plate is restricted from expanding by lugs attaching the rim of the plate to the shroud of the steam gm.erator and by the rim itself. This restraint causes additional stress concentration which can lead to cracking between the outer row of tube holes and flow holes.
The modification at Maine Yankee involved removing the rim material and support lugs from the #7 and #9 partial support plates in all three steam generators.
Permanent support of the #9 partial support plate has been provided by tube
" staking".
The procedure involved locating and expanding sleeves at the level of the support plate.
Five tubes were staked at each #9 support plate.
These stakes have been tested and shown to support the plate under main steam line break conditions. The #7 plate was previously supported in the middle by four tic rods in each plate. No additional support was deemed necessary.
Each staked tube was plugged.
This change does not constitute an unrevieweu safety question as defined by 10 CFR 50.59 (a) (2).
PDCR 6-78 LOW STEAM GENERATOR BY? ASS BISTABLE MODIFICATION The NT-4 in che auxiliary logic drawer is used as a bypass bistable in the Reactor Protective System and provides a contact which, when closed, allows the low steam generator pressure bistable to operate.
During the calibration of the BS13 section of the auxiliary logic drawer it had been difficult to get a steady setpoint voltage value. This was caused by a change on the input impedance of the autioneering inputs. The NT-4 design used for BS 13 uses an input current and a 15 volt de power supply to generate its setpoint value. A change in the input impedance changes this current which results in a change of set point.
In order to eliminate this dependence on input impedance, the " Front end" comparator operational amplifier was re-cnnfigured so as to increase the input impedance and isolate the setpoint voltage divider circuit from the input circuit.
This eliminated the interaction between the input signal current and the setpoint signal current.
This change does not constitute an unreviewed safety question as defined in 10 CFR 50.59 (a) (2).
EDCR 75-11 ADDITIONAL FILTRATION The Exhaust Ventilation System at Maine Yankee, as it existed prior to this change, provided exhaust ventilation for the Primary Auxiliary Building, the Fuel Building, and the purge / exhaust ventilation for the Reactor Containment.
The system was inadequate in that it could not provide particulate filtration and radiodine absorption capabilities for all three exhauststreams simultaneously.
To alleviate this situation two new charcoal filter assemblies were installed.
Tiiis change increases the filtration capability of the existing exhaust ventila-tion system and does not constitute an unreviewed safety question as defined in 10CFR 50.59 (a) (2).
EDCR 76-17 TM/LP TRIP CALCULATORY MODIFICATION One of the requirements to increase Maine Yankee's operation capacity to 2630 MWt was to change the old Thermal Margin / Low Pressure (TM/LP) Trip calculator.
The function of the TM/LP calculator is to generate a primary pressure trip point below which the reactor cannot operate for given values of reactor power level and primary coolant reactor inlet temperature, without exceeding thermal protection limits of the reactor.
Calculation of the previously existing set-point input value was done assuming a worst case value for symmetric offset.
This value of symmetric offset was overly conservative.
The change performed consists of the removal of the four old calculators from each of the four channels in the Reactor Protective System and addition of newer calculators that accept a variable symmetric offset function. Addi-tionally, several resistors were installed in the Reactor Protective System Calibration and Indication Panel to permit adjustment of the inputs to the digital panel meter scaling amplifiers located in the TM/LP calculator.
Finally several new as well as several existing adjustable TM/LP coefficient resistors were repackaged and relocated in the RPSCIP to permit optimum use of space.
This change does not constitute an unreviewed safety question as cefined by 10 CFR 50.59 (a) (2).
EDCR 77-23 DIESEL GENERATOR ALARM MODIFICATIONS In a letter to Maine Yankee the NRC indicated that it had found that misleading information concerning diesel generator operability could possibly be provided to the control room operator. The letter requested a review of diesel generator alarm circuitry to determine how each condition that renders a diesel generator unable to respond to an automatic emergency start signal is alarmed in the control room.
Diesel generator trouble is annunciated both locally at the Engine Control Panel (ECP) and at the main control room (typical of both diesel generators). The ECP annunciator has individual trouble lights for each diesel generator alarm condition.
At the main control board, however, a single window provides a general annunciation indicating that there is a diesel generator trouble alarm at the local ECP annunciator.
Previously, annunciatory operation was such that silencing an alarm at the ECP annunciator also cleared the annunciator at the main control room even though the alarm condition may have still existed. This situation was remedied by converting the existing single main control room annunciator window to a
" split" window indicating " Disabling Condition" on the top half and " Trouble" on the botton half.
The alarm contacts were modified such that the upper half window will annunciate only for disabling conditions and cannot be reset until the disabling condition clears. Additionally, the lower half window will now annunciate any alarm condition at the ECP annunciator (disabling or non-disabling) similar to the previous window.
A limit switch was installed on the manual operator assembly of the Emergency Fuel Stop Valve located at the engine to provide a contact closure when the valve is closed.
The contact remains closed until the valve is reopened.
"he contact has been included with the auxiliary contacts for disabling conditions at the ECP.
There is a common station service annunciator window for loss of control power to the diesel generator /,ir Circuit Breakers (ACB) the previous inscription of the window was " DIESEL GEN ACS LOSS OF CONTROL POWER".
For clarity, this inscription has been changed to read "DG 1A or 1B ACB LOSS OF CONTROL POWER".
This change does not constitute an unreviewed safety question ac defined by 10CFR50.59. (a) (2).
l-
EDCR 78-8 SECURITY LIGHTING POWER SUPPLY Federal Regulation 10 CFR 73.55 (c) (5) issued by the Nuclear Regulatory Commi.asion required improved security lighting systems at all nuclear power plants. Amongs; these requirements was the need for emergency power to provide lighting within one minute of a loss of normal power.
To provide a connection for the security lighting load, a thermal magnetic circuit breaker was installed in the 480 voit Emergency Motor Control Center located in the protected switchgear room.
This circuit breaker has a shunt trip which automatically trips the security lighting on receipt of an accident signal. The tripping of this non-safety claas load will help to assure the in-tegrity of the emergency bus during an accident. The automatic accident signal to the shunt trip is provided by a lock out relay of the Safety Injection Actuation System which is located in the main control board.
Control power for the shunt trip coil is provided by a control transformer located in the circuit breaker cubicle.
This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 (a) (2).
a.
EDCR 78-10 CONTROL ROD MODIFICATION In 1975 an " Auto-Off" relay feature was added to the Control Element Drive System (CEDS) which automatically returned the CEDS to the "off" mode within 3 seconds of release of the manual control (In-Hold-out) switch. The time delay was to ensure completion of the worst case Coil Power Programmer's (CPP) timing cycle prior to the CEDS Logic System 'aing placed into the "off" mode.
The problem which necessitated this EDCR was that there was a particular sequence of CEDS control panel actions, namely mode selection and manual control switch operation, which could create a shortened "Lif t" coil pulse to a single CEA during its first outward step.
The CEA could drop as a result of this phenomenon due to inadequate CEA movement cnd lack of magnetic jack mechanism engagement.
The shortened " Lift" coil pulse arose due to a signal conflict to the Group and Mode Select module located in the CEDS Logic.
Basically, for a short time period, the Group and Mode Select module could in effect be in two modes
("off" and one other) until the " Auto-Off" relay picked up.
The "off" mode is translated to an open CEA Li.'t Coil enable relay contact in each and every individual CEA Control Module. When any mode of operation other than "off" is selected at the CEDS control panel the respective CEA Lift Coil enable relay contact is closed allowing normal command signal interpretation by tb.
CEA's CPP.
Ti.c solution to alleviate the dual mode select logic problem and still main-tain the auto-o'f feature was the addition of an integrated circuit to the Group and Mode Select modu); which logically prohibits more than one active state at a time.
This change does not constitute an unreviewed safety question as defined in 10CFR50.59 (a) (2).
n,
EDCR 78-13 CEA PLUG ASSEMBL,Y Maine Yankee had eight part length control element assemblies (CEA's) which were originally incorporated in the NSSS design of the reactor for purposes of controlling Xenon oscillations.
Because of regulatory imposed constraints, technical specifications prohibited use of part length CEA's, and also since that would this time there is no reactor controlling Function anticipar-atinvolve use of these CEA's the eight part length CEA's and at
'ated drive shafts have been removed from the reactor.
The eight part length rods were replaced with a CEA plug. The plug device is engaged onto the top of the fuel assembly and essentially represents the These CEA plug assemblies were installed to lower end of a part length CEA.
the core hydraulics would be unaffected as a result of removal ensure that of the part length CEA's.
constitute an unreviewed safety question as defined by This change does not 10CFR50.59 (a) (2).
f
. <> e EDCR 78-15 CEDM TIME DELAY LATCHING RELAY Recently it was shown that any vibration when opening the door to each Control Element Drive System (CZDS) cabinet might cause a rod drop to the rod associated with the cabinet.
.1 further investigation revealed that the time delay relay added by EDCR 77-f (120 volt AC Power Supply to the Maine Yankee Coil Power Prog 71mmer Timer Module 15 VCD Power Supply) was susceptible to vibration.
To alleviate the situation a relay was added across the output of the time delay relay in the control element drive mechanism.
This relay energizes d
after the time delay reiay has energized, thereby latching in the delave signal. This delayed signal is used by the Coil Power Programmer Timer Module. The Timer Module circuit provides " Enable" command signals to various CEDS coils.
This change does not constitute an unreviewed safety question ac defined i' 10CFR50.59 (a) (2).
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