ML20128G734
| ML20128G734 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/12/1985 |
| From: | Whittier G Maine Yankee |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| 5784L-FWS, GDW-85-173, MN-85-116, NUDOCS 8507090303 | |
| Download: ML20128G734 (91) | |
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l EDISON DRIVE MAIRE HARHEE ATOMICP0MIERCOMPARUe avausra, un,ne oas3s q{
(207) 623-3521 9
June 12, 1985 MN-85-ll6 GDW-85-173 Region I United States Nuclear Regulatory Commission Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406 Attention:
Dr. Thomas E. Murley, Regional Administrator
References:
(a) License No. DPR-36 (00cket No. 50-309)
Subject:
Annual Report of Facility Changes and Relief and Safety Valve Failures and Challenges Gentlemen:
In accordance with 10 CFR 50.59, attached please find a report containing a brief description of the facility changes completed at the Maine Yankee Atomic Power Company plant during 1984.
In lieu of 39 copies required by 10 CFR 50.59, a master microfiche of this submittal will be provided under separate cover.
There were no relief and safety valve failures or challenges during 1984.
Very truly yours, MAINE YANKEE ATOMIC POWER C0bPANY hhW G. D. Whittier, Manager Nuclear Engineering & Licensing GDW/bjp Attachments: Annual Report of Facility Changes (91 Pages) cc: Mr. James R. Miller Mr. Cornelius F. Holden Mr. James Taylor, Director, I & E Washington 8507090303 850612 PDR ADOCK 05000309 ff R
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-s INDEX TO DESIGN PACKAGES COWLETED IN 1984 PA 28 Reroute EV-1 and 2 Relief Lines to Primary Vent Stack PA 27-82 Auxiliary. Steam From Crossunder Header PA 38-82 C-A1A, B, ~& C Aftercooler Drain Modification PA 3-83 Refurbishment of Refueling Machine PA 4-83 Circulating Water Stainless Steel Amertap Piping Replacement with PVC Piping PA 11-83 Containment Purge Exhaust Room Radiation Shield Door PA 14-83 Hydrogen Purity Instrumentation Replacement PA 18-83 SFP Deep Bed Filter Timer PA 22-83 Staff Building Undervoltage Trip PA 23-83 Uninterruptable Power Supply to FEMCO PA 25-83 Reactor Trip / Turbine Trip Redundancy PA 27-83 Independent Operation of Containment Sump Pumps PA 28-83 NNS Instrumentation Multiplexer PA 30-83 Alarm for TCB-9 or RPS Panel PA 33-83 Separation of Well Water and Domestic Water Systems PDCR 11-82 Replacement of Containment Electrical Penetration E-2 and E-ll PDCR 15-82 Replacement of Safaety Related Limitorque Motors PDCR 22-82 Manual Operator for MS-T-163 PDCR 4-83 Emergency MCC Feeder Remote Indicator Lights PDCR 5-83 Limitorque Motor Splice Upgrade PDCR 6-83 Modification of Housing Interlocks - W SI, LPSI, and Containment Spray Pumps PDCR 7-83 Personnel Airlock 0-Ring Retention Modification EDCR 79-36 Ground Detection System for DC and Vital AC Busses
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EDCR 80-50 Grid Undervoltage Protection EDCR 81-11 Mitigation of Harsh Environment 5669L-FWS
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O' EDCR 81-18 Replacement of Safety Related Terminal Blocks EDCR 82-31 Debris Screens for Containment Air Supply and Purge Exhaust Lines
~EDCR 83-09 Appendix R Modifications EDCR 83-18 Feedwater Spring Cans EDCR 83-20 Pressurizer Safety Relief Valve Piping Modification EDCR 83-22 79-02 Support Modifications EDCR 83-24 Capability of Main Steam Piping to Hold Water EDCR 83-28 Condensate Surge Tank System EDCR 83-29 Auxiliary Feedwater System Modifications EDCR 83-32 Inadequate Core Cooling Monitoring System EDCR 83-500 Isophase Bus Upgrade EDCR 83-501 Containment Hi Range Rad Monitor Cable Swap EDCR 83-505 Elimination of "ISH" Interlock EDCR 83-509 Main Steam Non-Return Valves Vacuum Assist System EDCR 83-512 Motor Driven Feed Pump Recirculation System EDCR 83-513 Main Turbine /P-2C Trip Interlock Bypass EDCR 83-514 ICI Cable Upgrade EUCR 83-515 AFW Seismic Upgrade
-EDCR 83-517 Condenstate Overboard Discharge Line Check Valve EDCR 83-518 Steam Generator Chem Feed Line Penetration Isolation EDCR 83-519 Feedwater Pipe Arrangement Modification EDCR 83-520 Removal of AFW-1 Internals EDCR 83-527 MSR Tube Bundle Replacement EDCR 83-529 Modification to Reactor Vessel Closure Head Lift Rig EDCR 83-530 Waste Oil Pump - Trip on High Tank Level EDCR 83-531 Auto Close of FW-M-336 on P-2C -Trip 5669L-FWS
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. EDCR 83-533 Addition of an Air Supply Station Near Containment Personnel r
Hatch EDCR 83-534 Traveling Water Screen Corrosion / Degradation Improvements EDCR 84-02 Qualified Level Transmitters for Containment Sump Indication EDCR 84-04 Performance Instrumentation Upgrade EDCR 84-06 Reroute Scavenging Vent to First Point Heaters EDCR 84-08 Modification to Steam Generator Primary Manway Gasket Retainer Plates EDCR 84-09 Loss of Il5KV Voltage Alarm EDCR 84-10 P-2C Low Suction Pressure Trip Time Delay EDCR 84-13 Containment Personnel Hatch Inprovements EDCR 84-14 Charging System Flow Restrictor for LTDP Concerns EDCR 84-15 Steam Generator Wet Lay-up Recirculation System Taps EDCR 84-16 Main and Auxiliary Feedwater and Main Steam Isolation Bypass Annunciation EDCR 84-19 Steam Dtrip Controller Output Test Feature EDCR 84-20 Main Steam Quality Measurement EDCR 84-21 Condensate /FW/CW Instrumentation Upgrade EDCR 84-23 PORV Block Valve Motor Operator Replacement EDCR 84-24 Replace Temperature Switches on C-5A & B EDCR 84-26 First Point Heater Vent Repairs EDCR 84-27 H.P. Turbine Moisture Pre-Separation EDCR 84-28 Jib Crane Mounting Reactor Containment El. 46' ED:R 84-30 S/G Separator Can Replacement EDCR 84-31 Addition of Locking Device to Polar Crane EDCR 84-32 Loose Parts / Thermal Shield Monitoring System Upgrade EDCR 84-36 Steam Generator Secondary Handhole Cover Thermowell EDCR 84-38 Check Valve Installation in Water Treatment Line EDCR 84-39 S/G Secondary Handhole RTD's 5669L-FWS
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EDCR 84-40 Replacement of MVAR Watt Transducer in Sect "A" K B r
EDCR 84-41 NNS Multiplexer RTD Inputs EDCR 84-43 Screen Wash Pump Strainer Replacement i
EDCR 84-47 011 Retention Curb for P-2C EDCR 84-50 Computer Cable for TSC and EOF EDCR 84-54 VAX/VMS Terminal Cable Runs EDCR 81-47 Reactor Vessel Head Temp Monitor EDCR 83-31 LTOP Variable Setpoint EDCR 84-11 H.F. Containment Parameters Relocation EDCR 84-12 H.F. Wide Range Log Recorder Replacement t
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MAINE YANKEE ATOMIC POWER COMPANY c
PA 28-81 Reroute EV-1 and 2 Relief Lines to Primary Vent Stack PA 28-81 installed a cross-connect line, connecting the relief lines from the evaporators, EV-1 and EV-2, with the line leading from the Blowdown Tank, TK-18, to the Primary Vent Stack.
The normal discharge from the evaporator relief valves, should an overpressurization condition occur, is to TK-18.
The new cross-connect line provides an alternate flowpath so that the Blowdown Tank can be isolated for maintenance or inspections and any discharge from the evaporators during this period will be directed to the Vent Stack.
This design change was implemented to ensure proper safety for those personnel performing tank maintenance.
The system modified is non-nuclear safety class. This modification does not alter the function of the evaporator relief line nor does it alter the function or operability of any other existing plant system. Additionally, this plant alteration does not create an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY 4
PA 27-82 Auxiliary Steam from Crossunder Header PA 27-82 involved the addition of a new auxiliary steam supply line from the 42" relief valve line for MSR's A & B to the existing auxiliary steam header.
The newly installed 8 inch line incorporates a new pressure control valve which is controlled through split ranging with the exisiting auxiliary steam supply valve to allow HP turbine exhaust steam to supply all plant auxiliary steam demands above approximately 85-88% plant power level.
This modification was installed to increase plant output by approximately~1250 KW (assuming auxiliary steam flow of 60,000 lbm/Hr. and high pressure turbine efficiency of 81%).
All components of this design change are classified as non-nuclear safety class. This alteration does not involve any unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PA 38-82 C-1A, B, and C Aftercooler Orain Modification PA 38-82 removed the solenoid operated drain valves from the service and instrument air conpressor aftercoolers and replaced them with ball float traps and level gauges. This change was to allow slow steady draining of the system with minimal air loss and to allow visual verification of condensate level.
This change was to a system which is non-nuclear safety class. This I
alteration does not constitute an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPA96Y PA 3-83 Refurbishment of Refueling Machine The refueling machine was refurbished in order to improve its reliability and speed of operation during refueling.
Mechanical improvements were made to facilitate maintenance, improve decontamination of hoist assembly shim plates, added a viewing window, a new electronic based control console, general refurbishment of the structure, new l
removable TV camera system, and state-of-the-art motor drives.' This alteration does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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O PA 4-83 Circulating Water Stainless Steel Amertap Piping Replacement with PVC Piping This alteration involved replacing the existing schedule 10 stainless steel piping on the circulating water Amertap System with PVC piping. The change was necessitated by corrosion of the stainless piping.
The circulating water Amertap System is non-nuclear safety class.
The replacement of the piping did not affect the operation of any other plant system. Therefore, this alteration does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PA 11-83 Containment' Purge Exhaust Room Radiation Shield Door Previously the auxiliary feed pump and containment purge exhaust rooms were interconnected by an open passageway. Since the containment atmosphere essentially extends into the purge room through the ventilation duct work up to the location of the first isolation valves, VP-A-3 and 4, radiation levels in this area following an accident could make the pump room inaccessible.
However, the pump room must remain accessible as various post accident valving l
operations are performed from this area, i.e., aligning RCS hot leg injection, hydrogen recombing, and containment purge / hydrogen contral systems.
PA-ll-83 installed a sliding radiation shield door to isolate the auxiliary feed pump room from the containment purge exhaust room following an accident to ensure that the habitability of this area is maintained.
This plant alteration does not adversely affect existing plant systems or components. The shield door was installed solely to provide radiation protection for personnel in the unlikely event of a major accident. Since the door is large in size and is located in an area near existing safety class equipment, the door was seismically mounted to ensure that a failure does not occur, which could damage vital station components. This alteration does not constitute an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PA 14-83 Hydrogen Purity Instrumentation Replacement PA-15-83 replaced the obsolete, mechanical hydrogen purity meter for the main generator with a state-of-the-art electronic system.
This enhanced maintainability and reliability as well as improving operation.
The generator hydrogen monitoring system is non-nuclear safety class.
The ability to monitor generator fill gas was enhanced, no other plant systems were affected. Therefore, this change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PA 18-83 SFP Deep Bed Filter Timer PA-18-83 installed a time totalizing meter on the safety class deep bed filter (FL-80, 81, 82). This timer keeps a running total of filter usage time to facilitate maintenance.
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The timer is non-nuclear safety class. Nothing was mounted on seismic portions of the filter. No control function is provided by the timer.
Therefore, this change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PA-22-83 Staff Building Undervoltage Trip T'his change automatically trips the staff building feeder breaker when bus 3 undervoltage occurs. The breaker will be reclosed after the voltage has been restored. This eliminates unnecessary loads when voltage is low. This plant alteration does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PA 23-83 Uninterruptable Power Supply to FEKO PA 23-83 changed the power supply for the plant paging system to a more reliable supply to assure continued operation and availability. By changing to a different battery-backed inverter power supply, previously encountered power interruptions due to bus switching during outages will be eliminated.
This alteration does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWEft COMPANY PA 25-83 Reactor Trip / Turbine Trip Redundancy PA-25-83 improved the reliability of turbine trip in the event of a reactor trip. This change provides for a separate and independent means of tripping the main turbine on CEDM bus undervoltage. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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e MAINE YANKEE ATOMIC POWER COMPANY PA 27-83 Independent Operation of Containment Sump Pumps, P-13 A & B PA 27-83 provided for an upgrade to the controllers of the containment sump pumps, P-13A & B.
It is now possible for control room operators to run either or both sump pumps as desired. The low level pump auto-stop function remains intact, as does the manual override of the low level pump cutout. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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PA 28-83 NNS Instrumentation Singal Multiplexer PA 28-83 installed a remotely operated signal selector in containment with its control unit in the control room. This allows one of sixty-four input signals to be selected for monitoring, using only one cable and electrical penetration. It is not to be used for control signals nor for safety class instrumentation.
It is not to be used for continuously monitored NNS instrumentation, nor for any signal important to plant operation or safety.
The multiplexer is non-nuclear safety class and is mounted away from any safety class equipment. This alteration does not involve an unreviewed safety question as defined in 10 CDR 50.59.
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M AINE YANKEE ATOMIC POWER COMPANY PA 30-83 Alarm for TC8-9 on RPS Panel An audible alarm has been added to the RPS monitoring panel to indicate the opening of TCB-9 (tie breaker for MG-1A & B). This alarm should make the operators aware that the two motor-generator sets could be operating in a non-synchronous mode. This change improves the monitoring capability for important electrical equipment. It does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PA 33-83 Separation of Well Water and Domestic Water Systems PA 33-38 separated the well water and domestic water systems. As a result, each system is now totally dedicated to serving a specific plant site water requirement. The domestic water supplies flushes, showers, wash basins and the well water supplies all the potable water fixtures.
Both of the water systems affected by this PA are non-nuclear safety class.
This changes does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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. MAINE YANKEE ATOMIC POWER COMPANY POCR 11-82 Replacement of Containment Electrical Penetration E-2 & E-ll i
Penetrations for the control element drive mechanism fans were replaced since parts are no longer available for the previously installed penetrations. The new penetrations are equivalent replacement of an existing component. This change does not involve an unrevien d safety question as defined by 10 CFR 50.59.
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s MAINE YANKEE ATOMIC POWER COMPANY PDCR 15-82 Replacement of Safety Related Limitorque Motors This design change involved the motor replacement of six safety related Limitorque motor operators. The six motors were replaced with radiation exposure-qualified direct replacements as part of Maine Yankee's commitment to environmental qualification of safety related electrical equipment per I&E Bulletin 79-01B.
This modification did not involve an unreviewed safety question as defined by 10 CFR 50.59.
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PDCR 22-82 added a sidi mounted handwheel to MS-T-163 to provide manual operating capability. /MS-T-163 is a CIS/SIAS, air operated isolation valve to y-e t
the turbine driven auxiliary feedwater pump (TDAFP). During a total station
.". i. blackout, MS-T-163 vill remain open for an estimated four hours (until the air i
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' accumulator loses p'ressure),., With the manual handwheel the valve can be
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s opened and 'stegam con be.fedl.a the TDAFP, should the need arise. Steam pressure for the TDAFP can be regulated by'the hindwheel on MS-P-168 located down stream from MS-T-163.,
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The handwhe,el wilfnohnally be in the neutral position and locked.
Thus the
.-,4 normal functon,of tre valve will not be altered. This change does not involve 8
_:1' an unreviewed safety question as per 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PDCR 4-83 Emergency MCC Feeder Remote Indicator Lights PDCR 4-83 installed four red and green indicator light pairs on the electric control board. These lights indicate the positions of the feeder breakers from busses 7 and 8 to MCC's 7A, 78, 8A, 88. This provides useful indication of MCC readiness status in the control room.
1 This change does involve a safety class system. Safety class parts were utilized and seismic acceptability of component mounting has been assured.
This change does not invovle an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY PDCR 5-83 i
Limitorque Motor Splice Upgrade-Motor.. splices to some Limitorque operators were replaced with environmentally qualified splices. This change is a direct replacement of existing equipment with functionally equivalent but environmentally qualified equipment. This alteration does not. involve an unreviewed safety question as defined by 10 CFR 50.59.
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M AINE VANKEE ATOMIC POWER COMPAMY PDCR 6-83 Modification of Housing Interlocks - WSI, LPSI and Containment Spray Pumps Under PDCR 6-83 the housing interlock for the high pressure safety injection (HPSI), low pressure safety injection (LPSI), and containment spray pump (CS')
have been modified to allow the use of a ground truck in an inoperable pump cubicle without disabling its respective spare pump. Relocating and replacing the interlock switch allows the spare pump to operate if the A or B pump cubicle has a ground truck and the elevating mechanism is racked up. This PDCR improves the reliability of the WSI, LPSI, and containment spray pump power supplies and, therefore, does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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M AINE Y ANKEE ATOMIC POWER COMPANY PDCR 7-83 Personnel Airlock 0-Ring Retention Modification This modification involved the addition of 3/32" filler wire on the outer edge of the a-ring grooves on the inner and outer doors of the personnel airlock.
The wire will hold the o-rings in place more firmly and thus reduce maintenance.
This change increased the reliability and reduced the maintenance on the personnel air lock. The airlock is a Safety Class 2 containment isolation boundary. This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 79-36 l
Ground Detection System for DC and Vital AC Busses l
A high impedance ground detection system was installed under EDCR 79-36 in the NNS DC bus. The system will be' activated and evaluated before a similar system is installed in the safety class busses.
This does not involve an j
unreviewed safety question as defined by 10 CFR 50.59.
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M AINE Y ANKEE ATOMIC POWEQ COMPANV ECCR 80-50 Grid Undervoltage Protection This change upgraded the undervoltage protection for 4160 volt safeguard busses. The system will alarm on low voltage.
Coincident with SIAS, the bus will be supplied by the emergency diesels.
This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 81-11 Mitigation of Harsh Environment I&E Bulletin No.79-018 requires that a high energy line break (HELB) outside containment must not result in an environment capable of causing failure of safety class electrical equipment.
EDCR 81-11 modified three high energy piping systems.
The modification provides automatic isolation of the auxiliary steam, blowdown, and letdown systems when a high ambient temperature is sensed by the Air Temperature Monitoring System (ATMS). The ATMS observes the ambient temperature at various plant locations and initiates a trip signal to close isolation valves following a HELB, thereby, ensuring the harsh environment is short term and operability of vital electrical equipment is not compromised.
The modifications performed on the letdown and blowdown systems are safety class 1 and 2, respectively, and the auxiliary steam system changes are non-nuclear safety class. This change inproves the reliability of both safety i
related and non-safety related systems.
This change does not create an l
unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 81-18 Replacement of Safety Related Terminal Blocks I.E.Bulletin 79-018, " Environmental Qualification of Class lE Equipment",
requires that all equipment in safety related systems subject to harsh environments be qualified to that environment. The safety-related terminal blocks inside the Reactor Containment lacked any qualification documentation and were replaced with Weidmuller-type environmentally qualified terminal blocks. This modification does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EOCR 81-47 Reactor Vessel Head Temperature Monitor I
EDCR 81-47 installed a conax dual RTO in a thermowell on the Reactor Head.
Cabling was added to connect the RTO through containment penetrations and landed on terminal blocks in the Main Control Board. The cables, connectors and dual RTD are all Class lE. Future EOCRs will utilize this modification to provide the actual indications for R.V. Head Temperature on the Main Control Board. This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EOCR 82-31 Debris Screens for Containment Air Supply and Purge Exhaust Lines EDCR 82-31 installed debris screens in the containment purge supply and l
exhaust piping systems to prevent the remote possibility of debris being blown into the purge lines during a LOCA event. This debris could prevent the isolation valves from fully closing and consequently failing to seal the Containment Building should on-line purge be in progress coincident with a Loss of Coolant Accident.
This design change is safety class 2 as the purge supply debris screen is attached to the containment pressure boundary.
The operation of the purge supply and exhaust system is not affected by the change.
This EDCR does not constitute an unreviewed safety question as defined in 10 CFR 50.59.
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EDCR 83-09A, B, & C i'
Appendix R Modifications This design change consisted of modifications to meet Maine Yankee's commitment to the requirements of 10 CFR 50 Appendix R.
These modifications included the following:
1.
The installation of an alternate shutdown power supply consisting of a 230 KV diesel generator, a 480V motor control center, and 125V DC station battery and battery charger, and a 2KVA inverter.
2.
The installation of an alternate shutdown panel to replace the existing steam generator emergency panel in the Auxiliary Feedwater Pump Room.
3.
The installation of all necessary process instrumentation to develop the necessary parameters for the Alternate Shutdown Panel.
4.
The installation of all necessary alternate pump and valve control i
circuits to allow control from the Alternate Shutdown Panel.
5.
The installation of all necessary fire barrier material to separate normal and alternate shutdown system.
This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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EDCR 83-18 Feedwater Spring Cans EDCR 83-18 involved the replacement and/or resetting of six spring cans on the three feedwater lines in the main steam valve house.
The supports affected by this design change are located within the Safety Class 2 boundaries on the feedwater lines. The modification to the six supports was performed to ensure that the supports will serve their intended function without jeopardizing the piping system.
Therefore, this design change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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Pressurizer Safety Relief Valve Piping Modification t
The SRV's and their inlet piping are Safety Class 1.
The modification modified flow characteristics which inproved stability and thus inproved reliability of the SRV's.
This modification did not ir.volve an unreviewed safety question as defined in 10 CFR 50.59.
I This modification involved minimizing the length of the SRV inlet piping, i
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MAINE YANKEE ATOMIC POWER COMPANV EOCR 83-22 i
i 79-02 Support Modification EDCR 83-22 involved the modification of several supports on several different safety class systems. These modifications are a result of analysis performed using criteria of IE Bulletin 79-02, " Baseplate Flexibility Using Concrete Expansion Anchors".
This design change ensures that the modified supports are capable of performing their intended function. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 83-24 t
Capability of Main Steam Piping to Hold Water l
Cold shutdown capability for Appendix R considerations necessitates flowing water from the steam generators to the condenser via the main steam piping.
This design change modified three supports on the main steam piping and two l
supports on the atmospheric steam dump line in order for the main steam lines to adequately support static water loading. This change does not involve an unroviewed safety queston as defined by 10 CFR 50.59.
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EDCR 83-28 i
Condensate Surge Tank System j
i EDCR 83-28 added a Condensate Surge Tank, TK-122, to serve as surge volume for j
l the secondary plant condensate makeup and spill system.
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performed by the DWST but Technical Specification revisions require that the j
l DWST be isolated from NNS systems during plant power operation. The exclusivu l
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function of the OWST, during plant power operation, will be to provide storage capacity for the safety class emergency feedwater system.
EDCR 83-28 includes a 120,000 gallon stainless steel tank with concrete foundation (located outdoors),10" condensate piping to connect to existing secondary condensato makeup header, steam injection for tank heating, tank level monitors, high/ low level alarms, temperature monitor, and required heat l
tracing. The new equipment is NNS but the tank is seismically designed to f
protect nearby safety class pumps.
This modification improves the safety design of the plant. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEF ATOMIC POWER COMPANY EOCR 83-29 Emergency Feedwater System Modifications This design change modified the Emergency Feedwater system by adding an additional isolation valve in each feed line.
The new valves automatically close when a low pressure condition in the valve-associated steam generator is sensed.
The associated Emergency Feedwater control valve also isolates on the same low pressure signal.
The two valve isolation guards against a single active failure during S/G depressurization caused by a main steam line break.
The valves will automatically reopen upon restoration of S/G pressure.
The modifications performed are safety class 3.
This design change enhances nucicar plant safety and does not constitute an unreviwed safety question as defined in 10 CFR 50.59.
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B EDCR 83-31 LTOP Variable Setpoint i
l EOCR 83-31 replaced PT-103 and PT-103-1 Pressurizer Pressure channels (NNS) i l
with two 0-2500 psig pressure transmitters PT-103 and PT-105, as part of a I
l program to address Low Temperature Overpressurization Protection (LTOP) i questions. The signals from these transmitters were each connected to their respective controller. The controller setpoint and process indication are l
compared in order to actuate the PORVs if the process is above the setpoint.
The LTOP system is armed in the variable pressure setpoint relief position r
under administrative control.
The two trains, each having a controller, process indication and separate digital setpoint indication, provide two l
independent trains to preclude electrical failure from interrupting the correct actuation. Power for each train is from independent power sources.
Alarms are retained for the purpose of ensuring the proper arming and disarming of the LTOP system.
f This modification did not involve an unreviewed safety question as defined in 10 CFR 50.57.
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l ECCR 83-32 l
Inadequate Core Cooling Monitoring System 1
Three new saturation monitors were installed.
They provide indication of the margin to saturation at three places in the Reactor Coolant System (RCS) - the core, the reactor vessel head, and the S/G tube region (RCS side).
This EDCR does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEC ATOMIC POWER COMPANY EOCR 83-500 Isophase Bus Upgrade Under EDCR 83-500 the isophase bus system bolted grounding straps were replaced by welded laminated expandable shunts. This change reduces yearly maintenance and does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANV EDCR 83-501 i
Containment Hi Range Rad Monitor Cable Penetration Swap EDCR 83-501 moved the high range radiation monitor cables from triaxial i
penetrations, to coaxial penetrations, which are correct for the cable type.
This change inproves the reliability of instruments important in the unlikely event of a serious accident. Safety class, environmentally qualified parts were used. This change does not involve an unroviewed safety question as defined by 10 CFR 50.59.
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l EOCR 83-505 Elimination of "ISi" Interlock EOCR 83-505 permanently eliminated an interlock in the Control Element Drive i
Control System which had been temporarily de-activated since 1980.
This allows a four step variability in shutdown rod group position to slow guide tube wear.
The CEDCS is non-nucicar safety class.
Technical specifications allow the change, and four years experience assured its functionality. This change does not involve on unreviewed safety question as defined by 10 CFR 50.59.
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M AlNE YANKEE ATOMIC POWEQ COMPQNV EDCR 83-509 Main Steam Non Return Valves Vacuum Assist System This change involved adding a line from the main condenser to the top of each main steam non return valve. The vacuum from the condenser holds the non return valves open during low steam flow conditions.
This reduces
" fluttering" of the valves and also allows the main steam system to be used for decay heat removal, t
This chan0e does not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE VANKEC ATOMIC POWEN COMPANY EDCR 83-512 Motor Driven Feed Pump Recirculation System EDCR 83-512 increased the recirculation flow capabilities for both motor driven main feed pumps (P-2A & 0).
The original recirculation systems did not provido adequate pump flow during recirculation operation. Previous measurements of pump pressure pulsations have indicated that P-2A & D experienced cavlation during operation in the recirculation modo. This cavitating flow condition has contributed to severe pump impeller cyn damago and high pump vibrations.
Increasing the recirculation flow capability to 4000 gpm will reduco pump internal recirculation and inlet vano cavitation, and provido the capability to operate for increased periods of timo on recirculation without risking severe pump damage.
This chango does not constitute an unroviewed safety question as defined in 10 CFR 50.59.
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M AINE YANNEC ATOMIC POWE A COMPANY EOCR 83-513 Main Turbine /P-2C Trip Interlock Bypass This EDCR installed a key lock bypass switch on the main control board.
The switch may be used to defeat the coincidence trip functions betwoon the main turbino and the turbine driven feedpump, P-2C.
An annunciator has been installed to alarm the bypass function.
This change allows plant startup on the turbino driven feedpump when the main turbino is tripped.
The coincidence trip circuits between the main turbine and P-2C are bypassed under administrativo control with the key lock switch.
This alteration does not constituto an unroviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 83-514 ICI Cable Upgrade EDCR 83-514 installed new, qualified cable sets for seven core locations, in which new detectors with qualified core exit thermocouples were installed.
This ensures that cables for these core locations inside the containment will remain operable in the unlikely event of a LDCA.This change affects a safety clacs lE system. However, this is an upgrade of the environmental qualification of the components with no functional change. This change does not involve an unreveiwod safety question as defined by 10 CFR 50.59.
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i EDCE 83-515 EFW Seismic Upgrade The design change involved the modification of existng supports and addition of new supports to the electric Emergengy Feedwater punps recirculation lines and other EFW piping.
Also, a new check valve was added to the steam-driven AFW pump recirculation line. The check valve will serve as the seismic pressure boundary between the seismic electric-driven EFW pump recirculation piping and the non-seismic turbine-driven AFW pump piping.
The AFW system operation was not affected by this change. This change does l
not involve an unreviewed safety question as defined in 10 CFR 50.59.
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Thid.changeinvolvedtheinstallationofa6"checkvalveinthecondensate g overboard line which ties into thedervice water discharge header.
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- [, /g' eliminates b e possibility of introdeding seawater, and hence, chlorides into the condensarfif an isolation valve is inadvertantly left open.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 83-518 Steam Generator Chemical Feedline Penetration' Isolation f
This design change involved the cutting and capping of 3 small bore pipes on both sides of a containment penetration. Since the lines were no longer utilized, this design change provided for.a means of permanently modifying the
- penetration such that annual IST penetration testing is not required.
This change does not involve an unreviewed safety question as defined by 10 CFR
- 50.59.
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MQlP$E YANKEE ATOMIC POWER COMPANY EDCR 83-519 Feedwater Pipe Arrangement Modification EDCR 83-519 replaced sections of the main feedwater piping and steam generator inlet nozzle thermal sleeves. The S/G nozzle thermal sleeves were replaced as they were slightly deformed during the 1983 water hammer event and the piping was replaced as a precautionary measure. The feedwater pipe configuration was changed in order to minimize the number of welded joints in the horizontal pipe sections adjacent to the S/G inlet nozzles, which previously experienced thermal stress cracking.
This modification is safety class 2.
The new pipe configuration does not adversely affect the operation of the feedwater system.
This EDCR does not present an unreviewed safety question as defined in 10 CFR 50.59.
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M AINE YANKEE ATOMIC POWER COMPANY,
EDCR 83-520 Removal of AFW-1 Internals The DWST provides condensate storage for the suction supply to the auxiliary feed pumps and AFW-1 is' a gate valve which isolates the.DWST from the suction header feeding the auxiliary feed pumps. Single failure of this valve in the closed position would prevent operation of the auxiliary feed system during an emergency. Therefore, EDCR 83-520, essentially eliminates AFW-1 by removing the valve internals (disc)~and handwheel thus modifying the auxiliary feed system to satisfy single failure criteria.
This modification inproves the safety design of the plant.
This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 83-527 MSR Tube Bundle Replacement This change involved the replacement of the existing MSR tube bundles with optimized tube bundles and the addition of perforated plates in the inlet side of the chevron banks in the MSRs B and D.
These modifications result in
. increased reliability and efficiency.
The MSRs are in a non-nuclear safety class system. This change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 83-529 Modification to Reactor Vessel Closure Head Lift Rig EDCR 83-529 replaced the heavy lift linkage components that are utilized to attach the reactor vessel closure head lift rig to the containment polar The replacement components were upgraded to comply with Section 3.2 of crane.
ANSI 14.6 - 1978, "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4,500 kg) or More for Nuclear Materials", as outlined in NUREG-612, " Control of Heavy Loads at Nuclear Power Plants".
The reactor vessel closure head lifting rig is not a safety related structure. The replacement components were designed, fabricated, and installed in accordance with ANSI 14.6 - 1978, which encompasses and/or exceeds the design criteria and installation requirements of Maine Yankee's safety class structures.
The replacement components were designed with a larger factor of safety than the original design. This change does not involve an unresolved safety question as defined by 10 CFR 50.59.
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M AINE YANKEE ATOMIC POWER COMPANY EDCR 83-530 Waste Oil Pump - Trip on High Tank Level EDCR 83-530 installed a float switch on the waste oil storage tank. This switch prevents the pump from pumping oil into a full tank, causing overflow.
The waste oil tank, pump and new level switch are non-nuclear safety class.
This change safeguards against accidental tank overflow. This change does not
~ involve an unreviewed safety question as defined by 10 CFR 50.59.
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I MAINE YANKEE ATOMIC PCWVER COMPANY EDCR 83-531 Auto Close of FW-M-336 on P-2C Trip EDCR 83-531 wired contacts of' relay 83-FC-1303 into the FW-M-336 (P-2C Discharge Valve) control circuit. The relay energizes on P-2C trip and in this application closes the discharge valve. This prevents P-2C from being driven backward by pressure from P-2A, B, if its discharge check valve fails.
This change and all associated equipment are non-nuclear safety class.
This design prevents potential equipment damage. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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e MAINE YANKEE ATOMIC POWER COMPANY EDCR 83-533 Addition of an Air Supply Station Near Containment Personnel Hatch This modification involved the addition of a service air hose station near the perscnnel air lock.
A new run of 1" piping 'was tied into the existing air line running overhead.
This change affected a non-nuclear safety class system and did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 83-534 Traveling Water Screen Corrosion / Degradation Improvements EDCR 83-534 included several modifications which were aimed at reducing the level of periodic maintenance required for the four traveling water screens which serve to remove debris from the plant intake cooling water. A new elastomeric urethane coating material was selected to replace the historically ineffective bitumastic coating. Also, an impressed current cathodic protection system was added to all four intake bays to protect areas which were left uncoated or subjected to coating damage / wear.
In addition, a spare screen unit adapted for universal drive configuration was purchased to establish quick change out capability and allow more flexible and less costly routine maintenance programs for the screen units.
All aspects of this design change were classified as non nuclear safety related, and thus the alterations outlined above do not involve any unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EOCR 84-02 l
Qualified Level Transmitter for Containment Sump Indication i
The level transmitters in the containment sump have been replaced by environmentally qualified transmitters. They have a range of 0-10 feet and indicate in the control room. An existing system was upgraded to assure operability under post accident conditions. This change does not involve an unreviewed safety geustion as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-04 Performance Instrumentation Upgrade Additional instrumentation has been added locally on the Moisture Separator Reheaters to monitor their performance.
The information available from this new instrumentation, coupled with our existing instrumentation for the MSR System will enable a calculation of the following, based on actual data for each MSR:
1.
Shell side inlet conditions including pressure, flow and quality.
2.
Chevron pressure drop and moisture carryover.
3.
Shell side tube bundle pressure drop.
4.
Shell outlet conditions including flow, pressure and temperature.
5.
Reheater heat transfer to shell side fluid.
This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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EDCR 84-06 Reroute Scavenging Vent to First Point Feedwater Heaters Changes under EDCR 84-06 allow the drains from the MSR Scavenging Steam Vent Condensor to be automatically routed to either the MSR shell (inlet) or to the first point heaters. This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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. MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-08 Modification to Steam Generator Primary Manway Nozzles EDCR 84-08 instituted the use of a modified steam generator primary manway gasket retainer plate.
The new plates allow gasket installation separately from the primary manway covers. They provide easier and more accurate positioning of the primary manway gaskets.
The gasket retainer plates are non-nuclear safety class. The tapped holes for the plate's positioning cap screws were located in the primary manway flange, which is safety class 1.
The modified components continued to comply with the appropriate design criteria. Thus, this modification does not increase the probability of an accident, equipment malfunction, or the occurrence of an accident not previously evaluated. Therefore, this change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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e M AINE YANKEE ATOMIC POWER COMPANY EDCR 84-09 Loss of 115 KV Voltage Alarm EDCR 84-09 upgraded the alarm functions associated with the loss of 115 KV offsite power supply. This change installed an audible alarm in addition to the visible Panalarm which annunciates after a ten second time delay to minimize nuisance alarms.
The control room personnel are alerted by this alarm function to loss of reserve station power under all operating conditions, including problems with transformer X-14 when in the standby mode. This change does not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-10 P-2C Low Suction Pressure Trip Time Delay EDCR 84-10 modified the method of initiation of a P-2C low suction pressure trip. The pressure switch was removed, and a new Sigma Meter-Relay inserted in an existing instrument loop. Relay alarm points (2) drive an annunciator and a time delay relay. The relay trips P-2C if pressure stays low for 15 2 1
seconds.
This equipment is non-nuclear safety class.
The change provides new indication and prevents nuisance trips. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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9 MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-11 Human Factors Containment Parameters Relocation EDCR 84-11 relocated several instruments located on the Main Control Board that measure containment pressure, sump water level, hydrogen concentration and High Range radiation to another location in the safeguards section dealing with containment isolation. This was done to logically group instruments of similar functionality to increase the effeciency and effectiveness af their The functions of each individual meter or instrument did not change, use.
only the location.
These instruments are Class lE for post accident monitoring. They were relocated while retaining all of Class IE characteristics such as channel separation, seismic mounting, etc.
This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MQlNE VAWKEE ATOMIC POWER COMPANY EDCR 84-12 Human Factors Wide Range Log Recorder Replacement EDCR 84-12 replaced a Wide Range Log Power Recorder with a computer driven CRT Trend. The recorder will be retained in another location until the CRT trend can be programmed and any problems worked out in order to satisfy Fuman Factors criteria. Long term historical data trending in hard copy will be resolved before completion of the recorder removal. The new CRT will provide improved monitoring capabilities by eliminating the Human Engineering Deficiencies in the chart recorder and not creating any new ones.
The inputs necessary to drive the CRT are generated by the plant process computer.
This isolation feature retains the necessary class lE to NNS separation.
This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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4 MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-13 Containment Personnel Hatch Improvements This design change involved several mechanical changes to the containment personnel hatch in order to inprove the hatch's overall reliability. The changes are as follows:
1.
Installed new wear rings, with additional screws, on the inner door locking ring.
2.
Increased the quantity of grease fittings on both locking rings.
3.
Replaced the wedges on both doors and locking rings.
4.
Replaced the hinges (both doors) with an updated version of the hinge.
5.
Installed new adjustable locking pin holders.
6.
Installed a system to provide the capability of testing between the 0-rings of the inner door from outside of containment. This system is similar to the one installed for the outer door.
The operation of the personnel hatch was not affected by this design change.
This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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Charging System Flow Restricter for LTCP Concerns l
l This change involved the installation of a multi-stage flow restricting orifice in the chemical and volume control system line. This orifice reduces the effects of low temperature overpressure transients when the RCS pressure j
is less than 400 psig by restricting the high pressure safety injection flow to 200 gallons per minute or less.
At RCS pressures below 400 psig the flow l
restricter is valved in and the W SI valves are locked out.
This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-15 Steam Generator Wet Lay-Up Recirculation System Taps In order to provide a more efficient method of controlling steam generator water chemistry during wet lay-up, a recirculation system has been designed for each steam generator. This design change provided the pipe taps and root i
valves that will be used to connect the recirculation system to the recirculation loops.
The taps were installed on the main steam lines and the blowdown lines. This modification did not involve an unreviewed safety i
question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-16 Main and Auxiliary Feedwater and Main Steam Isolation Bypass Annunciation This design change involved a change to the main and auxiliary feedwater regulator valves steam generator low pressure trip circuitry that provided an alarm directly from the control switch whenever any switch is in the bypass position.
This modified the old arrangement which drove a bypass alarm from an auxiliary relay. This design change also provided an alarm whenever one of the excess flow check valve control switches are left in the "0 PEN" position which also bypasses a low steam generator pressure trip. This modification did not involve an unreviewewd safety question as defined in 10 CFR 50.59.
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M AINE YANKEE ATOMIC POWER COMPANY EDCR 84-19 Steam Dump Controller Output Test Feature EDCR 84-19 installed an on-line test feature to the steam dump control circuitry.
This new test capability permits verification of the availability of the steam dump signal from the steam dump controllers. This change inproves the reliability of the steam dump controller circuitry. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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EOCR 84-20 Main Steam Quality Measurement Isokinetic steam sampling nozzles have been added to each of three main steam lines to be used to obtain samples of main steam for testing purposes.
These nozzles are located inside the containment building, upstream of the main steam non-return valves. As such, this installation was designed, installed and tested to nuclear safety class II requirements.
This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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e MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-21 Condensate /FW/CW Instrumentation Upgrade This design change provided for the installation of additional instrumentation to more effectively monitor the performance of the circulating water and condensate systems.
The modifications consisted of the following installations:
1.
Pitot tube taps in the circulating water inlet lines and DP transmitters across the circulating water pumps to allow for accurate measurement of circulating water flow rates.
2.
Circulating water discharge line thermocouples and conductivity probes to allow for accurate determination of discharge water temperatures and water levels in the discharge lines.
3.
Parallel plate pressure devices in the condensers to provide for more representative turbine bac5ressure indication.
4.
Thermowells on condensate pumps discharge to be used to measure condensate pump discharge temperature.
5.
OP cells downstream of the condensate header (connected to existing orifice plates) to be used to measure flow rate.
This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-23 PORV Block Valve Motor Operator Replacement This design change involved the replacement of the existing PORV block valve motor operators with identical safety class lE environmentally qualified replacements. This change was part of Maine Yankee's environmental qualification program in response to I&E Bulletin 79-01B. This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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EDCR 84-24 Replace Temperature Switches C-5A & B EDCR 84-24 replaced mechanical temperature switches in the containment air compressor alarm circuits with RTD's and indicating controllers. This provided new indication, eased calibration and slowed setpoint drift. Die compressors and alarms are non-nuclear safety class. New indication is provided to ease fault diagnosis. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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EDCR 84-26 i
First Point Heater Vent Repairs l
The original carbon steel vent piping from of the shell side of the first
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point feedwater heaters was replaced with a stainless steel material to eliminate erosion and leakage problems.
This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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the four turbine outlets and piping liquid water to the heater drain tank.
m C.'901sture cemoval improves the efficiency of the plant and reduces the erosion
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-28 Jib Crane Mounting Reactor Containment Elevation 46' EDCR 84-28 provided for the installation of a jib crane on the structural steel that supports the storage platform above steam generator number three.
The EDCR also provided for a positive anchorage for the cavity seal ring and auxiliary bridges stored on the platforms.
The jib crane is rated for three tons.
The frame to which the crane was mounted is safety class structural steel.
The structural steel frame was reanalyzed to include the effect of the design crane load, and modified as appropriate to ensure continued compliance with the plant's safety class structural steel criteria.
This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-30 S/G Separator Can Replacement EDCR 84-30 involved replacement of 94 of 98 moisture separator cans with new high capacity cans. To allow greater accessability to the center region of the steam generator, four original cans extending from the manway to the center region remained. Due to the possibility of overloading, these original cans were orificed. As a result of the height increase associated with the new cans, new dryer drain lines with relocated bends were installed.
These modifications did not affect the pressure boundary of the steam generators and as such were assigned only a Q.A. related classification. This design change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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EDCR 84-31 Addition of Locking Device to Polar Crane l
EDCR 84-31 installed a locking mechanism on the hook of the polar crane to prevent the rotation of the crane hook on its thrust bearing. The modification involv'ed only the immediate components about the crane hook.
The modifications had no effect on the strength of the structural components.
This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-32 Loose Parts / Thermal Shield Monitoring System Upgrade Five new continuous monitoring channels have been installed in this system which increases the system sensitivity and monitoring capability.
The upper head system consists of three magnetically mounted accelerometers while the lower head system consists of two accelerometers attached to the incore instrumentation guide tubes.
This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-36 Steam Generator Secondary Handhole cover Thermowell This design change provided for the installation of a RTD thermowell in each of two steam generator secondary handhole covers on steam generator E-1A.
These thermowells will be used in conjunction with the RTD's installed under EDCR 84-39 to measure the temperature of the downcomer water in S/G E-1A.
This information will be used to calculate the steam generator recirculation ratio.
Since these thermowells are installed in the secondary side boundary of the steam generator, they were designed, installed and tested to nuclear safety class II requirements. This modification did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-38 Check valve Installation in Water Treatment Line Technical Specifications require that the DWST be isolated from NNS systems during power operation and that the DWST contain a minimum of 100,000 l
gallons.
In order to allow addition of makeup water to the DWST during power operation and, at the same time, preserve the safety class boundary, EDCR 84-38 adds two check valves in the condensate makeup line at the safety class boundary thus providing a seismic boundary and satisfying single failure criteria. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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i EOCR 84-39 S/G Secondary Hanchole RTDs i
i EDCR 84-39 installed resistance temperature detectors (RTDs) to monitor secondary side temperature in Steam Generator No. 1.
These RTDs were installed in thermowells in the secondary handhole covers described under EOCR 84-36. This change does not involve an unreviewed safety question as defined l.
by 10 CFR 50.59.
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.o MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-40
- Replacement of MVAR Wett Transducer in Sect "A" W B This transducer replacement replaced an existing non-linear unit with a linear one. -Its output feeds the computer and logs the power output of the plant-to the computer.
This EDCR does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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EDCR 84-41 NNS Multiplexer RTD Inputs Temperature transmitters were installed in the containment and connected to three inputs of the signal multiplexer installed under PA 28-83. This allows three 100 ohm Platinum RTDs to be monitored in the control room. Initially, RTDs were installed in the steam generator secondary handholds for use in calculating the steam generator recirculation ratio. These transmitters may be used for other RTDs in the future. This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-43 Screen Wash Pump Strainer Replacement EDCR 84-41 replaced the original "Y" type strainer located in the suction line of each screen wash pump with a basket type duplex strainer. This new installation allows strainer basket removal.and cleaning without interrupting the opration of the screen wash pumps.
This design change modifies a non-nuclear safety class system. The reliability and availability of the system will be enhanced by this modification.
This EDCR does not create an unreviewed safety question as defined in 10 CFR 50.59.
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1 MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-47 011 Retention Curb for P-2C An oil rentention berm was constructed around the steam driven feedpump (P-2C). The oil reservoir on P-2C holds a substantial amount of oil. The func'tlon of this berm is to prevent spreading of oil, should the reservoir rupture, thus preventing damage by the oil, the spread of fire, or the l
inadvertant overboard discharge of oil. This change is to a non-nuclear safety system.
This change does not does not constitute an unreviewed safety question as defined by 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-30 Computer Cable for TSC and EOF Cables were installed from the main computer to the Technical Support Center and Emergency Operating Facility. This change provides the capability for these facilities to monitor actual plant conditions during drills or in the unlikely event of an actual accident.
This change does not involve an unreviewed safety question as defined by 10 CFR 50.59.
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-O EDCR 84-54 VAX/VMS Terminal Cable Runs Under EOCR 84-54 cabling was provided to several locations within the plant and the Staff Building from the simulator computer room. This cabling provides access to the Digital Equipment Corporation VAX computer which is used to store data for maintenance and operations purposes. This cable permits accessing the VAX computer system from terminals throughout the plant. This change does not involve an.unreviewed safety question as defined in 10 CFR 50.59.
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