ML20199F993
| ML20199F993 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/06/1986 |
| From: | Whittier G Maine Yankee |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| 7556L-SDE, GDW-86-129, MN-86-74, NUDOCS 8606240631 | |
| Download: ML20199F993 (58) | |
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MAIRE HARHEE ATOMICF0WERComPARUe ,vaug,,Yl,,Te?,"a's h (207) 623-3521 O June 6, 1986 MN-86-74 GDH-86-129 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 (Docket No. 50-309)
(b) MYAPCo Letter to USNRC dated March 11, 1981 (FMY-81-33)
Subject:
Annual Report of Facility Changes and Relief and Safety Valve Failures and Challenges Gentlemen:
In accordance with 10 CFR 50.59, attached is a report containing a brief description of the facility changes completed at the Maine Yankee Atomic Power :
Station during 1985. In lieu of 39 copies required by 10 CFR 50.59, a master j microfiche of this submittal will be provided under separate letter. In '
Reference (b), Maine Yankee committed to reporting any challenges and/or failures of PORV and pressurizer safety valves. During 1985 there were no such events.
1 Very truly yours, l MAINE YANKEE ATOMIC POWER COMPANY khN G. D. Whittier, Manager i Nuclear Engineering and Licensing GDH/bjp
Enclosure:
Annual Report of Facility Changes cc: Mr. Ashok C. Thadani Mr. Pat Sears Mr. Cornelius F. Holden Mr. James M. Taylor, Director, Office of Inspection and Enforcement 7556L-SDE 8606240631 860606 9 h
DR ADOCK 0500 i )
. e e MAINE YANKEE ATOMIC POWER COMPANY
_ DESIGN PACKAGES COMPLETED IN 1985 EDCR 81-13 New Spent Fuel Racks EDCR 82-05 Refueling Canal Block Shields EDCR 83-504 Hi Resin Annunciator for Resin Hold-Up Tank EDCR 83-506 Improve Feedwater Temperature Indication EDCR 84-03 Replace RHST Level Switches with Transmitters EDCR 84-34 Boric Acid Mix Tank for S/G Treatment EDCR 84-42 S/G Het Lay-Up Recirculation System EDCR 84-45 Permanent Polymetrics Truck Hook-Ups EDCR 84-48 Automated Condenser Chloride Monitor System EDCR 84-49 PVS Particulate and Iodine Monitoring System EDCR 84-53 Activation of 20AST & 20ET on S.0.E.
EDCR 84-58 Exciter H2 Monitor EDCR 84-60 Improve SIAS Loss of Control Annunciator EDCR 84-63 Eliminate Seal Contacts for Spray Chemical Inlet Valve Control EDCR 84-66 Haste Transfer Backflush EDCR 84-68 Elimination of Unused Components EDCR 84-69 Control & Indication for VCT Isolation Valve EDCR 84-70 Prevention of Spurious P-2C Recirc EDCR 84-71 Reactor Head Vent Selector Switches EDCR 84-72 Reg Guide 1.97 Modifications - Pressurizer Pressure i EDCR 84-74 Interchange P-61A & P-61S Control Switches EDCR 84-75 Provide Indication for Steam Dump Overrides & PR-A-38 EDCR 84-78 PHST Heating System Modifications EDCR 84-81 Pressure Gauge Installation to Fuel Oil Transfer System EDCR 85-01 Remove Throttle Bushing d/p Instrumentation i EDCR 85-02 Pressurizer Panel Modifications EDCR 85-04 Installation of Aux Feedwater 1" Diam Check Valve EDCR 85-05 AFH Support H-317 Modifications EDCR 85-06 HELB II EDCR 85-07 Reg Guide 1.97 Modifications-Pressurizer Level EDCR 85-08 Reg Guide 1.97 Modifications-Steam Generator Level EDCR 85-09 Reg Guide 1.97 Modifications-RCS Temperature EDCR 85-10 Reroute Spray Bldg Aux Steam Piping EDCR 85-11 Main Control board Panel Modifications EDCR 85-12 Improve CVCS/ Loop Panel Layout EDCR 85-14 Seismic Anchorage of Cont RM & BATT RM Light Fixtures EDCR 85-15 IE Bulletin 80-11 Block Hall Modifications EDCR 85-16 Increase Aux Steam Supply for Start Up EDCR 85-17 Appendix J Primary Water Systein Modifications EDCR 85-18 Primary Inventory Trend System Modifications EDCR 85-19 Appendix J Primary Sampling System Modifications EDCR 85-20 Reactor Coolant Pump Lube Oil Collection System l EDCR 85-22 Post Accident Sampling System Modifications EDCR 85-23 P-2A&B Trip on Hi S/G Level EDCR 85-24 Appendix J Primary Component Cooling Modification EDCR 85-25 ECCS Valves Lightboxes Upgrade EDCR 85-26 Safety Injection Tank Panel Modifitdions EDCR 85-28 Replacement of First Point Heater Vent Lines EDCR 85-29 Improve Containment Air Sample & H2 Analyzer Tubing EDCR 85-30 Appendix J Primary Drain System Modifications EDCR 85-31 HP Turbine Skimmer Performance Instrumentation EDCR 85-34 ICI Cable Replacement .
EDCR 85-39 HP Turbine Moisture Preseparator Upgrade i EDCR 85-40 Post Accident Stack Sampling System I EDCR 85-42 Generator Stator RTD Installation l j
Service Hater Header Seismic Upgrade EDCR 85-44 hq 7556L-SDE g t1 i
t 9 MAINE YANKEE ATOMIC POWER COMPANY EDCR 81-13 NEN SPENT FUEL RACKS EDCR 81-13 replaces the existing sixteen aluminum fuel racks with new stainless steel poison racks which have a smaller center-to-center spacing.
This increases existing storage from 953 to 1,476 fuel assemblies, extending on-site spent fuel storage capacity to approximately the year 1995.
The new spent fuel racks perform the same function as the old equipment and are shown, by conservative analysis, to adequately meet applicable design requirements. This design modification was approved by the NRC in their
" Safety Evaluation and Environmental Impact Appraisal Regarding Maine Yankee Spent Fuel Storage" dated June 16, 1982.
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' MQ1NE YANKEE ATOMIC POWER COMPA93Y EDCR 82-05 REFUELING CANAL BLOCK SHIELDS EDCR 82-05 installed vertical steel ribs with attached strong grating shields at the opening at the north end of the refueling cavity at elevations 20' and -2'. The steel deflection shields restrain loosely stacked, solid, concrete blocks which are placed in the refueling cavity opening to reduce radiation levels in the containment annulus area during the transfer of spent fuel. The concrete block wall alone does not have any lateral stability, and in the event of a failure, could damage safety-related equipment in the
, proximity, if not protected.
The steel deflection shields are not considered safety c; ass, but since safety class systems are in close proximity, the modifications have been designed and installed as if they were quality assurance related. There is no increase in the probability of an accident or equipment malfunction or of an accident of a different type which has not been previously analyzed, therefore, this design changes does not constitute an unreviewed safety question as defined by 10 CFR 50.59(a)(2).
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- ' MAINE YANKEE ATOMIC POWER COMPANY EDCR 83-504 HI RESIN ANNUNCIATOR FOR RESIN HOLD-UP TANK EDCR 83-504 provides annunciation indication for a high resin level condition in the resin hold-up tank (TK-109). This was accomplished by wiring the unterminated TK-109 level switch leads to an annunciator and reconnecting the resin inlet valve (HSS-A-28) to its control circuit eliminating the yellow tagged jumper cable which performed this function in the past.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the !
possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the '
margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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- MAINE VANKEE ATOMIC POWER COMPANY EDCR 83-506 IMPROVE FEEDHATER TEMPERATURE INDICATION EDCR 83-506 improved the accuracy of feedwater temperature indication.
This was accomplished by replacing the 100 Ohm feedwater temperature RTD's (1220A, B, C) with 200 Ohm Rosemount RTD's. The existing RTD bridges located in Computer Cabinet "H" were disconnected and marked spare. Their function was replaced by Rosemount Series 444 temperature transmitters. A digital readout and selector switch was installed which allows manual monitoring of feedwater temperature for hand calorimeter calculations when the computer is unavailable. The new equipment was mounted in Computer Cabinet "H" and uses existing cabling to the RTD's and to Cabinet "D".
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59. It provides greater assurance of operation within the specified conditions as stated within accident analyses.
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4 O MAINE YANKEE ATOMIC POWER COMPAP$V EDCR 84-03 REPLACE RHST LEVEL SHITCHES WITH TRANSHITTERS EDCR 84-03 replaced the RHST RAS level switches (LS303AK, BK, CK and LS304AK, BK, CK) with the combination of a Rosemount 11538 level transmitter and an Acromag bistable. In addition, level transmitters LT30K and LT302K were removed. The increased accuracy of the new RAS initiation setup will provide greater assurance that at least 200,000 gallons are transferred before RAS.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it 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 84-34 BORIC ACID HIX TANK FOR S/G TREATMENT EDCR 84-34 replaces the secondary side boration system with a simpler, faster system having increased capacity. The new system consists of a 550 ghilon stainless steel tank piped to the condenser hotwell through a 1 1/2" S.S. pipe. The boric acid will be added manually to the water and mixed by an electric mixer. The boric acid solution will be drawn into the condenser by the vacuum in the condenser. It will take approximately 10 minutes to complete the injection process during start-up and by the use of control valves, the system can supply the smaller flow rates required for normal plant operation.
This change does not affect the FSAR, Technical Specifications, or any safety class systems. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANV EDCR 84-42 S/G HET LAY-UP RECIRCULATION SYSTEM EDCR 84-42 provided a means of chemically treating, sampling, and mixing steam generator water during wet lay-up. Three portable pump / tank units were installed, one for each steam generator. Recirculation is accomplished using a 3 HP, 50 GPM portable pumping unit. The time for one steam generator complete mixing is approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The recirculation units are located at El. 21'-0" in the containment outer annulus adjacent to the personnel hatch. This recirculation system provides better chemistry control by recirculating the steam generator water to assure a more uniform condition, it also provides taps for a strainer unit and a convenient sampling location.
This system will be used only during plant outages. During normal plant operations, the system components will be disconnected from the main steam and blowdown lines. This method of operation will classify the recirculation system as non-nuclear safety class. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MQiNE YANKEE ATOMIC POWER COMPANY EDCR 84-45 PERMANENT POLYMETRICS TRUCK HOOK-UPS EDCR 84-45 installed a permanent 2 1/2" schedule PUC pipe from the treated water pumps (P-79A, & B) to the polymetrics influent (both inside and outside locations) and then from the polymetrics effluent to the 6" common water supply header downstream of the resin trap (FL-58). This EDCR eliminates running a 21/2" fire hose from these locations.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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s MQ1NE VAGeKEE ATOMIC POWER COMPANY EDCR 84-48 ,
AUTOMATED CONDENSER CHLORIDE MONITOR SYSTEM EDCR 84-48 installed an automated three pump system, with a manual operation option for continuous monitoring of chlorides in the condenser hot wells. This upgraded system will achieve a sampling cycle, better trending and improved reliability, also with a continuous automated sampling system, the direction, identification and location of chloride contamination due to leaks is much quicker.
Existing condenser penetrations and isolation valves from each bay (A,B,C & D) were utilized for this system. New pumps were installed which tied into existing lines and did not increase the probability of causing loss of condensate or feedwater. Check valves on the suction lines to the pumps prevent any vacuum leaks to the condenser. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previcusly evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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i MAINE YANKEE ATOMIC POWER COMPAP5Y EDCR 84-49 PVS PARTICULATE AND IODINE MONITORING SYSTEM EDCR 84-49 replaced the tygon tubing that routes the sample slipstream through a series of instrumentation with stainless steel piping. To prevent the condensation problem, the existing insulation and heat tracing was upgraded and expanded to maintain a consistent temperature from the PVS to the
- sample air dryer. The sample slipstream is continuously drawn through the 1 sample line by one vacuum pump. To prevent an unmonitored gaseous release due to pump failure, a second pump was added which can be placed into service by manually transferring power and valving.
Reliability and improved operation of the PVS monitoring system is gained l by implementation of this design package. This change does not increase.the l probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of ,
Technical Specifications, therefore, it does not present an unreviewed safety <
question as defined in 10 CFR 50.59. ,
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MAINE YANKEE ATOMIC POWER COMPANV EDCR 84-53 i
ACTIVATION OF 20 AST & 20 ET ON SEQUENCE OF EVENTS EDCR-84-53 installed relays that will provide the computer with a contact status change to indicate the presence of a trip signal to trip coils 20 AST 1 and 20 ET. These devices provide control grade trips to the main turbine.
Adding the activation of 20 AST and 20 ET in the sequence of events computer 1 log will aid in determining the initiating cause of a plant trip.
These components will affect no existing safety class electrical system, and are non-nuclear safety class designation. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it 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 84-58 EXCITER H2 MONITOR EDCR 84-58 installed a hydrogen monitoring system for the main generator exciter. This consists of two sensors, mounted in the top of the exciter housing, and a monitor mounted near the generator H2 panel. A common meter switch selectively displays Channel A, Channel B, or automatically displays the higher of the two. Alarm contacts are set at 40 and 60% LEL (Lower Explosive Limit) by the manufacturer and are field resettable. These will be tied to separate panalarm windows on the H9 panel to provide separate alarms at alert and danger levels. This design change provides a warning of H2 leakage into the exciter before explosive concentrations are reached.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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M AINE VANKEE OTOMIC POWER COMPAPdY i
EDCR 84-60 IMPROVE SIAS LOSS OF CONTROL ANNUNCIATOR EDCR 84-60 relocates the electrical sensing point of the four ECCS sets of
, alarm relays ("74" devices) that monitor for an open circuit in the trip path
[ for each of the final actuation relays ("86" devices). The four associated
, " Loss of Control" annunciator windows on the RH annunciator panel (MCB i Section C) were relabelled to clearly indicate their function with the words "86 DEVICE TRIP PATH FAULT" preceded with the respective system designation (SIAS, CIS, CSAS, or RAS). Also, the annunciator's electrical logic was modified to reflect the slight sensing functionality change of the "74" l devices.
No changes were made to the ECCS actuation logic scheme. Implementation of this change eliminates false annunciation after safeguards system actuation, and provides more consistent information to operators, thereby 1 improving safeguards indication. This change does not increase the probability of occurrence or consequences of an accident or malfunction of j equipment nor does it create the possibility for an accident or malfunction of
- equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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M AlNE Y ANKEE ATOMIC POWER COMPONY EDCR 84-63 ELIMINATE SEAL CONTACTS FOR SPRAY CHEMICAL INLET VALVE CONTROL EDCR 84-63 removed the " Seal-in" contacts for the control circuits of valves CS-M-66 and CS-M-71. This change permits reversal of the MOV direction after an operator manually selects either "0 PEN" or "CLOSE" without waiting for the valve to complete full stroke. The valves now function in a throttable mode, and safeguards functions are retained. The operator is required to hold the control switch in the desired position for the full stroke time which is approximately 35 seconds. The operators have been provided with direction on waiting several seconds between valve direction changes to eliminate any potential for damage to the motor operators.
! The valve stroke time and response to an accident signal remain unchanged. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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- MAINE YANKEE ATOMIC POWER COMPANY EDCR 34-66 HASTE TRANSFER BACKFLUSH EDCR 84-66 added a selector switch that allows the existing control functions of valves HSS-A-41 and HSS-A-26, which can not be operated simultaneously, to remain "As-Is" with the added feature of simultaneously opening these two valves. The addition of the switch allows backflushing capabilities from TK-80 through HSS-A-41 and HSS-A-26 to EV-2. This achieves a thorough flush without adding to the waste system.
The addition of the switch will affect no existing safety class electrical system. Failure of the switch will not result in the release of radioactivity. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAIRE YANKEE ATOMIC POWER COMPAMV EDCR 84-68 ELIMINATION OF UNUSED COMPONENTS EDCR 84-68 eliminated various unused components that were taking up valuable room on the Main Control Board or present a distraction to the Control Room operators.
No changes were made to any Class lE or safety related circuits. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an
- accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as
- defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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e MAINE YANKEE ATOMIC POWER COMPANY EDCR 84-69 CONTROL AND INDICATION FOR VCT ISOLATION VALVE EDCR 84-69 relocated the control switch and associated "0 PEN-CLOSE" indicating light for CH-M-87 (VCT Isolation Valve) from the rear of Section C MCB to the front vertical panel of Section C MCB. This relocation resulted from a Human Factors detailed review of the Control Room which states that controls are not visible from the front of the board. The new switch location fits into the new CVCS PANEL M/H/C. Functionally, the controls are the same as before the change, and the system operation will be unchanged.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in i 10 CFR 50.59.
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MAINE VANKEE ATOMIC POWER COMPANY
. f EDCR 84-70 PREVENTION OF SPURIOUS P-2C RECIRC EDCR 84-70 installed a new solenoid valve, S0V-1303, between FN-A-342 and its positioner. When energized, this S0V will divert control of the valve away from its positioner, and will apply a fixed 65 psi signal to FH-A-342.
This will prevent the valve from opening due to controller, E/P on positioner failure. To prevent failure of a pressure regulator from allowing the valve to open a second regulator and two check valves were installed. This increases air supply reliability during blocked shut operation.
Failure of newly installed components will either have no affect or will transfer recirc control back to the existing systems. Failure of most existing system components has no affect above 80% power. This change is classified non-nuclear safety (NNS). This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAIME YANKEE ATOMIC POWER COMFANY EDCR 84-71 REACTOR HEAD VENT SELECTOR SHITCHES EDCR 84-71 replaced and relocated four control switches associated with four valves, two (RC-M-54 & 56) are used to vent the reactor head and two (PR-M-89 & 90) are used to vent the pressurizer. The selector switches were replaced with a key-lock electroswitch type 20K control switches. The switches were moved to the ESF panels. These new controls and associated valve position indicators are similar in design, size and operation to other controls with similar functions and comply with the " Economy of Space" concept.
This change does not alter the electrical design or functional characteristics of the circuit. It merely relocated these infrequently used controls to reduce visual clutter of the Main Control Board. All materials are Safety Class lE. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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f MQlPdC VANKEE ATOMIC POWER COMPANY 7
EDCR 84-72 REG. GUIDE 1.97 HODIFICATIONS - PRESSURIZER PRESSURE EDCR 84-72 provides the operator with two redundant 0 to 3250 psig pressurizer pressure channels, PT-102X and 102Y. The two channels were physically and electrically isolated and each includes a single pen strip chart recorder on the Main Control Board. This change also enables the operator to select the pressure input into the saturation margin monitor even though each channel is powered from a different vital bus. An isolator was installed in the PT-102Y channel input to the saturation monitor to provide the necessary separation between equipment powered from different vital busses.
The two pressurizer pressure instrument channels for input to the saturation monitors are part of a Class 1E system. All of the n'ew equipment was purchased and installed as Class lE components. Both channels provide input to the Safety Parameter Display System (SPDS). This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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s M AINE YANKEE ATOMIC POWER COMPANY EDCR 84-74 INTERCHANGE P-61A &P-61S CONTROL SHITCHES EDCR 84-74 consisted of swapping the P-61A and P-61S bench board mounted control switches together with the associated indicating lights and control switch labels. No functional changes resulted from this EDCR.
This change does not increase the probability of occurrence or '
consequences of an accident or malfunction of equipment nor does it create the possioility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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M AINE VAP6KEE ATOMIC POWER COMPANV EDCR 84-75 PROVIDE INDICATION FOR STEAM DUMP OVERRIDES & PR-A-38 EDCR 84-75 installed three amber indicating lights on the MCB Section 8 bench board. approximately one inch above each of the steam dump override pushbuttons. A flashing amber light gives positive indication that control of the feedwater regulatory bypass valves (FH-A-112, 212, 312) was transferred from the MCB controller to the trip set controller on the feedwater control cabinets behind the MCB. Also, two new NAMCO limit switches were installed for Valve Position Indication (VPI) on Valve PR-A-38. One green and one red indicating light we.e installed above the existing PR-A-38 selector switch on the MCB, green light "on" indicates valve is shut, red light "on" indicates valve is open and both lights "on" indicates the valve is in the intermediate position. These changes increase operator awareness of system status.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for in accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPORY EDCR 84-78 PHST HEATING SYSTEM MODIFICATIONS EDCR 84-78 added a self-contained temperature control valve in the steam line to the tank heater just downstream of the existing on/off control valve.
Control for this valve is obtained from a remote bulb installed in the recirculation line between the heat exchanger and the tank. This " hot leg" to the tank is maintained at approximately 220*F maximum to eliminate the formation of steam bubbles in the return line. By controlling the " hot leg" recirculation temperature the steam / water hammer previously experienced will be eliminated. Due to the absence of a sparger inside the PHST to disperse bubbles, the return line temperature is kept in the 220*F range or less. This EDCR also replaced the existing shell discharge impulse trap, which was incorrect for this heat exchanger system. A float and thermostatic trap was used for replacement. A manual bypass valve was installed in parallel with the control valve to allow for larger steam flows if necessary.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, l therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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M AIME YANKEE ATOMIC POWER COMPANY EDCR 84-81 PRESSURE GAUGE INSTALLATION TO FUEL OIL TRANSFER SYSTEM EDCR 84-81 added a pressure gauge to the discharge of each of the two auxiliary fuel oil pumps, P-33A and P-338. These gauges were fitted to existing valves F0-1 and F0-6. This design change was the result of adding the fuel oil transfer system to Maine Yankee's Inservice Testing Program.
! This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the 4 margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YONKEE ATOMIC POWER COMPAP$Y EDCR 85-01 REMOVE THROTTLE BUSHING d/p INSTRUMENTATION EDCR 85-01 disconnected and removed MCB indicators PDI-151, 161 and 171 and annunciator windows 12-8, 9, & 10. This throttle bushing d/p instrumentation was no longer used nor meaningful and created nuisance alarms, therefore, it was eliminated. The throttle bushing d/p transmitter outputs remain available from the plant computer and the transmitters remain valved in to the RCP throttle bushing d/p sensing lines.
No changes were made to any Class 1E or safety related circuits. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as definad in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.-
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-02 PRESSURIZER PANEL MODIFICATIONS EDCR 85-02 replaced the existing temperature selector switch with a new electroswitch HS-150. The new switch is typical of other switches on the Main Control Board performing similar functions. This switch is mounted directly below the temperature indication TI-150. In addition, a Rochester Instrument Systems card cage was installed in Panel 2, MCB Section C which will accept temperature input signals and provide outputs which can be selected by switch HS-150 and indicated on TI-150.
Valve stem leakoff temperature indicator, TI-3301 was removed from the MCB. The valve stem leakoff temperature will be displayed on TI-150 and is one of the temperature signals that can be selected by temperature selector switch HS-150.
The existing selector switch for selecting the controlling channel for pressure level HS-101-1 was replaced with a Hestinghouse type OT-2 selector switch. The new switch is similar in shape to others performing a similar switching function. This switch is located directly below the pressurizer level recorder LR-101.
The components affected are non-nuclear safety class (NNS). This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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EDCR 85-04 INSTALLATION OF AUXILIARY FEEDHATER 1" DIAMETER CHECK VALVE EDCR 85-04 installed a Safety Class 3 check valve in line 1"-HAPD-15-151 to limit inventory loss from the DHST should a pipe failure occur in the non-analyzed lube oil cooling and turbine bearing cooling water piping. The ,
single check is used as the seismic pressure boundary. A test connection and root valve was installed upstream to permit inservice testing in accordance with Section XI of the ASME Code to verify proper operation of the check valve. Pipe support upgrades on 1"-HAPD-15-151 ensure the seismic adequacy from the pump suction connection through and including the check valve to a point of orthogonal restraint.
This design change was implemented in accordance with Safety Class 3 requirements. The portion of the auxiliary feedwater system affected by this modification is classified Safety Class 3. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does*it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-05 AFH SUPPORT H-317 H0DIFICATIONS EDCR 85-05 modified hanger H-317. The 6 inch auxiliary feedwater line, No. 6"-HAPD-25-60lR3, supplying S/G #3 was in close proximity with the left side spring can on the hanger. This design change increased the clearance between the spring can and the pipe, thereby, reducing the possibility of the pipe contacting the support during a seismic event, which could affect the hanger's load carrying capacity.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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.7 MAINE VANKEE ATOMIC POWER COMPAP$V O
EDCR 85-06 1
HELB II l EDCR 85-06 modified the PAB HELB trip circuit. This was achieved by
! revising the trip logic of the Air Temperature Monitoring System so that it no 9
longer attempts to identify the source of high energy released to the PAB, but
- causes all high energy lines in the PAB to close (isolate) on a high j temperature trip signal generated by any of the sensor pairs located in the j PAB. This design change eliminates possible uncertainties regarding pipe break location and the flow path of the released steam. It ensures that the proper line has been isolated regardless of which high energy line is affected
! or where the high temperature is sensed.
The Air Temperature Monitoring System is a Safety Class lE system. This
- change does not increase the probability of occurrence or consequences of an 1 accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not I present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE VANKEE ATOMIC POWER COMPANY j
! EDCR 85-07 REG. GUIDE 1.97 MODIFICATIONS - PRESSURIZER LEVEL 4
. EDCR 85-07 provides the operator with two redundant and electrically 1 isolated pressurizer level channels, LT-101X and LT-10lY. It also provides redundant trending information by supplying the operator with a qualified i recorder and the ability to call up a trending display on the plant computer
. system, through the Safety Parameter Display System. An electrical isolator 1 was installed in each of the pressurizer level instrument loops that separates
> each channel into an indication loop that is maintained as Safety Class lE and a control loop that is maintained as NNS. Each indication loop contains a differential pressure transmitter, power supply, sigma indicator, isolator to the plant computer and an isolation to the control loop. The control loop contains the remainder equipment needed to control the level in the pressurizer. To provide trending for channel LT-10lX, the strip chart i recorder on the front of the MCB was replaced with a seismically qualified
! recorder and will be permanently wired into channel LT-101X. Trending for l channel LT-10lY was connected to an isolator card in SPDS cabinet 2, then connected to the plant computer as in the past (MPX Point 360).
j Each of the level channels are powered from two separate breakers on the
, same vital bus. One breaker for the indication loop and the second breaker l 1
for the control loop. The power supply was designed so that a failure in the l NNS equipment (control loop) would not cause the loss of the Safety Class lE indication (indication loop).
This change does not increase the probability of occurrence or
, consequences of an accident or malfunction of equipment nor does it create the i possibility for an accident or malfunction of equipment important to safety
. previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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M AINE YANKEE ATOMIC POWER COMPANY i
EDCR 85-08 REG. GUIDE 1.97 MODIFICATIONS - STEAM GENERATOR LEVEL EDCR 85-08 provides the operator with two redundant ways of trending steam generator level for each steam generator for post-accident monitoring. It provides the operator with a qualified recorder for each steam generator on the Main Control Board and the ability to call up a trending display on the plant computer,tThrough the Safety Parameter Display System. Three existing unqualified recorders (LR-1211X,1221X and 1231X) were removed from Section "B" of the MCB and replaced with qualified recorders to provide trending for steam generator level. These recorders were wired into Channel "A" of the RPS steam generator level channels (LT-1213A, 1223A, and 1233A). For a redundant means of trending, Channel "C" of the RPS steam generator level channels (LT-1213C, 1223C, and 1233C) were inputted to the plant computer for trending through qualified isolators.
l The steam generator level channels are Safety Class lE instrument loops.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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s M AINE YANKEE ATOMIC POWEM COMPANY EDCR 85-09 REG. GUIDE 1.97 H0DIFICATIONS - RCS TEMPERATURE EDCR 85-09 installed six dual output temperature transmitters. Three in hot let channels TT-111X,121X and 131X, and three in the cold leg channels TT-111Y, 121Y, and 131Y of the Reactor Coolant System temperature instrumentation. These transmitters have a wide range of 0* to 750*F and a narrow range of 515*F to 665'F. The hot leg channels are displayed on the back of Section "B" of the MCB and the cold leg channels are displayed on the front of the MCB in the loop control section.
The existing three cold leg 0* to 600*F, NNS, single pen recorders and indicators for TT-115Y, 125Y and 135Y were removed. They were replaced by three 0* to 600*F dual display cold leg indicators purchased for this EDCR and installed by EDCR 85-11. Their qualified 0* to 750*F dual pen recorders were also purchased by this EDCR and installed by EDCR 85-11.
This design change provides the operator with a trend of both hot and cold leg teniperature on a dual pen recorder for each loop on the front of the MCB.
The hot and 31d leg temperature channels are Safety Class lE instrument loops. All equipment was purchased and installed as IE components except for the 0* to 600*F dual indicator that is used in NNS instrument loops. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated i in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY I
EDCR 85-10 REROUTE SPRAY BUILDING AUXILIARY STEAM PIPING EDCR 85-10 rerouted two auxiliary steam lines. This allowed for total removal of the lines that ran through the Personnel Hatch area, the area around fans FN44A & B and the spray building supplying heating steam to the DHST and RHST tank heaters and several building heaters. The two new auxiliary steam piping supplies tie into the existing system downstream of AS-P-61 and AS-P-477 respectively. They exit the valve house via an existing 12 inch wall penetration on the East wall. This routes these lines outside which creates a passive system for protection of any safety related equipment in the event of a steam leak. Existing control valve stations were tied into at both the DHST and RHST heaters. Traps were added at these two low points for protection of the control valves and heaters. The new steam line to HV-7 and HV-9 penetrates the wall at HV-7 and feeds the existing regulating and drain systems to each unit and space heaters PUH-14A and B. The two steam fed space heaters (PUH-15A and B) for the cubicle outside the personnel hatch were removed and replaced with two electric heating units.
The auxiliary steam lines are an NNS system. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-11 MAIN CONTROL BOARD (MCB) MODIFICATIONS EDCR 85-11 installed a modular designed main control board face. Several design changes were made to the types and location of instruments on the main control board due to Human Factors considerations required by the NRC. Since the extent of those changes made it impractical to salvage, a modular design allowing easy installation and removal was developed for the areas that required extensive rework.
This EDCR was required to close out several design changes since this package documents the necessary seismic analysis required for those modifications. Those EDCR's which required seismic integrity verification were:
84-60 84-63 84-68 84-69 84-71 84-72 84-74 84-75 84-80 84-82 85-01 85-02 85-07 85-08 i 85-09 85-12 85-25 85-26 A seismic analysis of all affected areas was conducted to ensure that the seismic integrity of the structural supports for the Safety Class 1E equipment on the MCB was not reduced. This change does not increase the probability of i occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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1 MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-12 l
IMPROVE CVCS/ LOOP PANEL LAYOUT
, EDCR 85-12 provided for the relocation of controllers, indicators, 1 switches and replacement of a limiter, controller, and indicator by a programmable indicating controller. For these changes in location of instruments, for CVCS and Loop Control Systems, refer to sketches #1 and #2 (ESK-4C & 4G) and see Table 1 of this EDCR. The new panel cutouts and facial i
plates were completed in accordance with EDCR 85-11 " Main Control Board Facial Plates for H.F. Design Changes".
The intention of this change was to provide a more logical configuration to the operators and reduce the possibility of error.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety j previously evaluated in the FSAR. Also, this change does not reduce the
, margin of safety as defined in the basis of Technical Specifications, i therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE VANKEE ATOMIC POWER COMPAPSV EDCR 85-14 SEISMIC ANCHORAGE OF CONTROL ROOM AND BATTERY ROOM LIGHT FIXTURES EDCR 85-14 modified the ceiling and light fixtures in the Control Room and the light fixtures in the Battery Room at elevation 45'-6". It was determined that these lacked sufficient anchorage to withstand a seismic event.
Additional ties were added to prevent the ceiling and light fixtures from falling and possibly injuring personnel and/or equipment. Also, two lighting fixtures were relocated in the Control Room. The new locations provide better lighting conditions that enables the operators to perform their job more efficiently.
The ceiling and light fixtures are not Safety Class. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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s MAINE YANKEE ATOMIC POWER COMPAD3Y EDCR 85-15 IE BULLETIN 80-11: MASONRY BLOCK HALL MODIFICATIONS EDCR 85-15 modified several existing block walls. Halls which were previously acceptable as analyzed with the arching method were modified as a result of the arching method not being approved by the NRC.
The walls that required modifications in the service building were the West and South walls between cols. 8 & 10 of the El. 45'-6" Battery Room. In containment the elevator North wall above El 46'-0" was modified. The South and Hest walls were not modified. However, the cables and conduit for the high range containment radiation monitors (RM-6113A, 8), reactor head vent
- valve (M0V-3007) and pressurizer vent valve (MOV-3009) were relocated to be outside the affected area of a wall collapse. Radiation detector RM-6113B was moved to the West side of the Containment structure to ensure that it is widely separated from RM-6113A.
In addition to the above changes, the PCC surge line was protected from a potential wall collapse and the 7 line between columns D & E between the service building and switch gear roon was reinforced. Modifications were also performed in the PAB in the vicinity of the I&C Hot Shop.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in l 10 CFR 50.59.
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s MAINE YANKEE ATOMIC POWER COMPONY EDCR 85-16 INCREASE AUXILIARY STEAM SUPPLY FOR START-UP EDCR 85-16 increased the supply of auxiliary steam during plant startup.
The capacity of the auxiliary boilers was not adequate to provide the entire maximum auxiliary steam demand during plant startup. The capacity of the new steam source provides the additional steam needed and reduces auxiliary boiler use during startup.
The existing connection on the main steam system, used to supply the auxiliary steam system through AS-P-3, is downstream of the main steam NRV's.
Since the NRV's are used to isolate the main steam from the turbine before condenser vacuum is established, main steam was not available to supply the auxiliary steam system during periods when maximum auxiliary steam demands occur. At the same time, main steam upstream of the NRV's was vented to atmosphere in order to control reactor coolant temperature. Providing a new connection in the main steam system upstream of the NRV's increases auxiliary steam supply and reduces the amount of expensive condensate lost when steam is dumped to atmosphere.
Pressure control valve AS-P-56 was supplying a small amount of main steam (from upstream of the NRV's) to the auxiliary steam system. This source of steam will not be necessary, therefore, AS-P-56 and associated piping and valves were removed in order to simplify the plant systems.
The new steam line is designated Safety Class 2 from the tie-in to and including, the first isolation valve. The isolation valve is a " Containment Isolation" valve and is remote operated from the control room. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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s M AINE Y ANKEE ATOMIC POWER COMPANY i
- EDCR 85-17 i
l APPENDIX J PRIMARY HATER SYSTEMS MODIFICATIONS i
} EDCR 85-17 modified the primary water system to conduct penetration leak l testing in a manner whicn will fully satisfy Appendix J requirements. A two i inch manual ball valve and two test connection / drains were installed in the-i primary water supply line which enters the containment building through
! penetration No. 37. The new two inch valve is located downstream of PH-80, t which is the inside containment leakage barrier. The new valve serves as a system test boundary during leak tests of the primary water system containment isolation valves. A 3/4" branch line was rerouted away from PH-80 allowing for the new boundary valve to be as close as possible to PH-80. Two test connection / drains were added, one between PH-80 and the new test boundary j valve, and one upstream of PH-A-78.
These changes were necessary to allow leak testing the primary water I
system containment isolation valves with air or nitrogen in accordance with Appendix J.
l The section of the primary water system associated with this change is
, classified as non nuclear safety class (NNS). This change does not increase
- the probability of occurrence or consequences of an accident or malfunction of
- equipment nor does it create the possibility for an accident or malfunction of
- equipment important to safety previously evaluated in the FSAR. Also, this j change does not reduce the margin of safety as defined in the basis of j Technical Specifications, therefore, it does not present an unreviewed safety '
question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-18 I PRIMARY INVENTORY TREND SYSTEM N0DIFICATIONS EDCR 85-18 provides protection from simultaneous loss of all instrument reference lines due to a single pipe break; it minimizes the effects of partial drawing of the reference leg and it improves the submergence capability of the Primary Inventory Trend System (PITS) transmitters.
l To provide separation of tubing, a new reference line was installed for d/p transmitter 3003 and connected separately to the reactor head. The effect
- of reference leg draining was limited by shortening the vertical length of leg
- directly above the instrument stalks. To protect the PITS cables from i post-LOCA submergence, the existing conduit was replaced with a water tight arrangement of tubing and fittings.
q The PITS tubing is Safety Class 1 and seismic. The PITS transmitters and j
associated cable and connectors are Class IE components. This change does not increase the probability of occurrence or consequences of an accident or
- malfunction of equipment nor does it create the possibility for an accident or
! malfunction of equipment important to safety previously evaluated in the
{ FSAR. Also, this change does not reduce the margin of safety as defined in
- the basis of Technical Specifications, therefore, it does not present an
- unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-19 -
APPENDIX J PRIMARY SAMPLING SYSTEM MODIFICATIONS EDCR 85-19 was necessary to allow containment leak testing of the primary
- sampling system in accordance with 10 CFR 50 Appendix J criterion.
Previously, three sampling lines ran from the primary sampling header inside containment through penetrations 62A, B, and C to the sample sink in the
, Primary Auxiliary Building (P.A.B.). The sample line penetrating containment
- at 628 was eliminated and the penetration capped. The two remaining lines provide sufficient redundancy to ensure an adequate system for obtaining reactor coolant samples. A new cross-connect of the two remaining sample
- lines was added in containment. This serves as an alternate flow path to ensure transportability of primary plant samples to the P.A.B. In the event
- one of the two flow paths become unavailable. Test connections and system test boundary valves were added to allow leak testing the sampling system containment isolation valves with air or nitrogen in accordance with Appendix 3.
l EDCR 85-19 modifications reduced the number of containment isolation (CI) valves in the primary sampling system from 15 to 4. Isolation of the system on SIAS will also be accomplished by the automatic closure of only four valves.
This change does not increase the probability of occurrence or j consequences of an accident or malfunction of equipment nor does it create the
, possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, 4
therefore, it does not present an unreviewed safety question as defined in 4
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s MAINE YANKEE ATOMIC POWER COMPANY O
EDCR 85-20 REACTOR COOLANT PUMP LUBE OIL COLLECTION SYSTEM EDCR 85-20 installed a separator lube oil collection system on all three reactor coolant pump motors as required by Appendix R. This system will reduce the potential of a lube oil fire by catching oil leaks before the oil could reach an ignition source. The upper bearing housing was fitted with a steel curb to contain any oil that leaks out of the system. The system (pumps, coolers, pipe, fittings, etc.) is contained within the limits of the curb. Spray shields were installed to ensure that oil from pressurized lines is contained within the curb. Oil collected within the curb drains to a vented collection tank located inside the loop area. The lower bearing exterior lines were fitted with a tray mounted directly below them. Oil that drops into the tray will flow out a separate drain and into the collection tank along with oil from the upper bearing.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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s MAINE YANKEE ATOMIC POWER COMPANY s
EDCR 85-22 POST ACCIDENT SAMPLING SYSTEM H0DIFICATIONS EDCR 85-22 corrected four deficiencies in the post-accident sampling
, system. These were the capability to obtain an undiluted sample, eliminate the water collecting in the gas septum line, correct the improper relieving capability for the flush-water pump, and eliminate the potential for over pressurizing the charging pump suction piping and instruments.
This change does not increase the probability of occurrence or
> consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE VANKEE ATOMIC POWER COMPANY 4 .
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EDCR 85-23 l
P-2A, & B TRIP ON HI S/G LEVEL i
l l EDCR 85-23 installed a control grade trip for the motor driven main feed j pumps, P-2A and P-2B on high steam generator level. The matrices of contacts l
which were used to trip the main turbine via 20AST and 20ET on high steam generator level will energize two new BFD-445 relays, UA-SGX1 and UA-SGX2.
These relays will activate 20 AST and 20ET and also trip and lock-out P-2A and
] P-28. The two new relays are isolated from AST and ET power supplies by fuses i such that relay failure can not disable the power supply for other trips.
} Either relay will trip both circuits (AST and ET) and will trip both feed
, pumps when required. Relays are normally de-energized so failure will not cause a trip. After a high level trip, both relays will seal-in and both
! motor driven feed pumps wilf stay tripped even after the S/G high level
! cordition clears. Each pump can be restarted by repositioning the control j switch to the "stop" or " pull-to-lock" position and returning to " start".
I This cht.nge does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety j previously evaluated in the FSAR. Also, this change does not reduce the ;
margin of safety as defined in the basis of Technical Specifications, i therefore, it does not present an unreviewed safety question as defined in j 10 CFR 50.59.
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EDCR 85-24
} APPENDIX J PRIMARY COMPONENT COOLING MODIFICATION l
i EDCR 85-24 addressed two issues which dealt with the penetration leak-l testing of the PCC lines that penetrate the containment building. One was the
. upgrading of piping and associated valves from the existing Safety Class 2
) boundary up to and including PCC-M-219 from Safety Class 3 to Safety Class 2.
l The second issue was the modification of the PCC lines that penetrate the
! containment structure to meet the penetration leak testing criteria as set forth in 10 CFR 50, Appendix J.
This design package has no affect on the mode of operation, and negligible
- affects on the delivery of PCCW to the cooling loads within containment. A j seismic analysis / evaluation was performed to ensure that the seismic
! qualification of the PCC system was not adversely affected by this EDCR. This j change does not increase the probability of occurrence or consequences of an i accident or malfunction of equipment nor does it create the possibility for an i accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as
- defined in the basis of Technical Specifications, therefore, it does not-1 present an unreviewed safety question as defined in 10 CFR 50.59.
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O MolNE YANKEE ATOMIC POWER COMPAPSV EDCR 85-25 ECCS VALVES LIGHT B0XES UPGRADE EDCR 85-25 electrically modified the light boxes on the main control board such that either of the two lamps in each " window" can be independently powered. The front faces were painted and relabeled. All ECCS position indicators are both " position" and color discriminable in that only the left side of each valve's window will illuminate green. All " blue lighted" position indicators are affected through the illumination of the right side of each applicable valve's window, which is blue as opposed to green for the ECCS position. To achieve the illumination logic as well as the remaining design basis considerations, a microprocessor was added between the valve position limit switches and the light box lamps.
Implementation of this EDCR resulted in a more easily interpreted ECCS systems valve status to the control room operators. No changes were made to the base functional concept of the light boxes. No Class 1E circuitry was modified or impacted and the seismic integrity of any system important to plant safety was not challenged. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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s MAINE YANKEE ATOMIC POWER COMPANV i
EDCR 85-26 i
SAFETY INJECTION TANK PANEL MODIFICATIONS EDCR 85-26 rearranged the existing safety injection tank panel components on the Main Control Board. This results in better system grouping which :
increases the efficiency of their use. This EDCR does not add any new ,
equipment to the Main Control Board and does not affect the functionality of l the existing components. '
The components relocated by this design change were seismically mounted to ensure that they will not compromise the operation of any Class IE equipment within the Control Board. This change does not increase the probability of j occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question i as defined in 10 CFR 50.59.
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M AINE YANKEE ATOMIC POWER COMPANY i EDCR 85-28 i
! REPLACEMENT OF FIRST POINT HEATER VENT LINES EDCR 85-28 completed the replacement of both carbon steel first point heater vent lines and valves with 304 stainless steel to eliminate erosion concerns. The steam venting system runs from the shell side of the first point feedwater heaters to the main condensers.
This change does not increase the probability of occurrence or i consequences of an accident or malfunction of equipment nor does it create the
, possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications,
- therefore, it does not present an unreviewed safety question as defined in I 10 CFR 50.59.
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e MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-29 IMPROVE CONTAINHENT AIR SAMPLE AND H2 ANALYZER TUBING EDCR 85-29 corrected several potential deficiencies in the containment air sampling system. A 1" bypass line and valve was installed around PAP-1 to allow continuous draining of 2" - ACC-39-151 if PAP-1 is inoperative and will be used for the alternate sample and H2 purge flow path. The bypass line was heat traced and insulated to be consistent with 2"-ACC-39-151. To ensure containment integrity the new valve (PAP-49) is locked closed during normal plant operation. A new sampling station was installed in the supply line to the 00MSIP. This was the location occupied by the containment air sample cylinder in the lower Primary Auxiliary Building (PAB). The grab sample line was insulated and heat traced to prevent condensation. A line was run from the sample station to the PAB ventilation system as a vent path to depressurize the containment air sample.
This EDCR improves the ability to determine the post-LOCA status of the containment atmosphere. The modified sampler system gives Maine Yankee better information to allow appropriate action post-LOCA to avoid the buildup of potentially explosive hydrogen gas in the containment building. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY 3
EDCR 85-30 l APPENDIX J PRIMARY DRAIN SYSTEM MODIFICATIONS I
EDCR 85-30 installed a two inch manual valve and a test connection / drain in the primary drain line which exits containment through penetratica No. 39.
- This permits leak testing the containment isolation (CI) valves associated
- with this penetration (PR-A-40, 41 and 42) in accordance with 10 CFR 50, Appendix J. The new two inch valve is located upstream of PR-A-40 which is
! the inside containment leakage barrier. The new valve serves as a test boundary during Type C test of the primary drain system containment isolation valves. A 3/4 inch branch construction was added directly adjacent to the new test boundary valve and is used as a low point drain and test connection. The function and pressure rating of the system remain unchanged.
i This change does not increase the probability of occurrence or '
! consequences of an accident or malfunction of aquipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in j 10 CFR 50.59.
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MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-31 HP TURBINE SKIMMER PERFORMANCE INSTRUMENTATION EDCR 85-31 relocated the two existing test taps to measure condensate flow through the 16 inch drain line from the HP turbine moisture separator (skimmer) system. This was necessary to allow more accurate measurement of the condensate flow through the skimmer drain line to the heater drain tank (TK-19). The injection tap was relocated upstream of NRV HD-349. This allows more complete mixing of the chemical tracer with the condensate by passing through the check valve. The sampling tap was relocated to a location on the drain pipe where 100% liquid samples can be expected to be obtained rather than a steam-water mixture as in the past. A pressure indicator was installed at the 24 inch header which collects moisture and feeds it to the skimmer drain line leading to the heater drain tank.
This EDCR does not alter the function or operation of the moisture separation system nor does it adversely affect any other plant system or component. The modifications are classified as non-nuclear safety class (NNS). This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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M AINE VANKEE ATOMIC POWER COMPANY EDCR 85-34 ICI CABLE REPLACEMENT EDCR 85-34 continued the upgrading process of the 45 In-Core Instrumentation (ICI) and signal sets by installing 10 more post-LOCA environmentally qualified cable and detector connectors. To date, 18 sets have been changed. This design change brings the total of environmentally qualified cable sets to 28.
This EDCR does not alter the function or description of the system as specified in the FSAR or have any impact on the technical specification requirements. This change upgrades the system to Safety Class by replacing existing cable sets with environmentally qualified materials. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it 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 85-39 HP TURBINE M0ISTURE PRESEPARATOR UPGRADE EDCR 85-39 provided for the orificing of one of the four 4 inch drain lines from each HP turbine exhaust nozzle to the 24" common vertical header with the goal of reducing the flow through the orifice line to as low as practical, while providing some small nominal flow for temperature equalization across the orificing device.
Each HP turbine ring separator (4 total) is drained by four 4" pipes which join together in a common 24" vertical header located under the turbine. The common vertical header is vented via a 12" line to the second point feedwater heater extraction system, and drained via a 16" line to the bottom of TK-19 (heater drain tank).
In each of the 4" drain lines that were orificed, a short vertical section of the existing carbon steel pipe was removed and replaced with a section of stainless steel pipe with an orifice plate welded inside.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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.6 MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-40 POST ACCIDENT STACK SAMPLING SYSTEM EDCR 85-40 installed a new post-accident stack sampling system designed to take an isokinetic sample from the primary stack sampling system. The existing primary stack sampling system takes an isokinetic sample from the primary vent stack. The new post-accident stack sampling system was installed to satisfy the requirements specified in NUREG-0737 and provide isokinetic and representative samples in accordance with ANSI N13.1-1969.
This EDCR does not affect any operating plant systems or other systems required for plant safety and does not constitute an unreviewed safety question as defined in 10 CFR 50.59. This system operates only during post accident conditions and is only required to monitor releases to atmosphere, therefore, plant safety is not affected. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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- M AINE YANKEE ATnMIC POWER COMPANY i
EDCR 85-42 GENERATOR STATOR RTD INSTALLATION EDCR 85-42 installed 36 conductors for the 12 additional RTDs on the new generator stator. The new cabling was run from the new generator stator terminal block to cabinet "H" located in the computer room. Changes were required to cabinet "H" to accommodate the new type of temperature transmitters.
The new generator stator is equipped with 36 RTD terminal points.
Twenty-four RTDs were utilized to measure stator winding hot gas discharge temperature (12 existing and 12 new) and four sets for spares on one terminal block. Eight sets for existing cold and intermediate gas RTDs are on the second terminal block. This provides for improved evaluations of temperature patterns and thus determine inspection and maintenance requirements and improve overall generator stator reliability for the life of the unit.
This EDCR does not affect the operation of any safety class components.
It improves evaluations of temperature trending for the main generator stator. The electrical cabling and temperature transmitters were purchased and installed as NNS components. This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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- 9 MAINE YANKEE ATOMIC POWER COMPANY EDCR 85-44 SERVICL HATER HEADER SEISHIC UPGRADE EDCR 85-44 seismically qualified the 24 inch service water pump discharge header in the circulation water pump house. This was accomplished by adding 2 new pipe restraints and modifying 4 existing supports. These modifications ensure continued system operation by restraining the 24 inch line thus preventing excessive line movement in the east / west directions which would exceed the manufacturer's lateral displacement of the expansion joints at the discharge of the service water pumps.
This design change improves the reliability of the service water system during a seismic event. The header is a common link between independent trains of the system and must be available during all design basis accidents.
This change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment nor does it create the possibility for an accident or malfunction of equipment important to safety previously evaluated in the FSAR. Also, this change does not reduce the margin of safety as defined in the basis of Technical Specifications, therefore, it does not present an unreviewed safety question as defined in 10 CFR 50.59.
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