ML19296D463

Annual Operating Rept,1979

ML19296D463
Person / Time
Site:	Millstone
Issue date:	02/29/1980
From:	NORTHEAST UTILITIES
To:
Shared Package
ML19296D459	List: ML19296D458 ML19296D463
References
NUDOCS 8003040621
Download: ML19296D463 (42)

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MILLSTONE UNITS 1 AND 2 ANNUAL REPORT This Annual Report has been prepared pursuant to the requirements of Title 10, Code of Federal Regulations, Section 50.59 and Sections 6.9.1.4 and 6.9.1.5 of the units' Appendix A Technical Specifications.

Common site reporting requirements are addressed in this section.

Common site facility changes and tests are administered under the control of only one unit, and their evaluations are provided in the section applicable to that unit assigned the responsibility. Common site procedure changes are addressed here.

The following procedures common to both units' FSAR required reviews pursuant to 10CFR50.59 due to the development of a new procedure or revisions during 1979:

1.

Revision 3 to the Millstone Modified Amended Security Plan (MASP).

The plan.was revised to include the latest NRC position on personnel searches. The revision did not affect the design or operation of either unit.

2.

Approval of an Off-Site Dose Calculation Manual (0DCM). The ODCM is to be used 'c.o implement the new Appendix I Technical Specifications and was submitted to the NRC for approval.

3.

Revision 6 of the Millstone Station Emergency Plan was approved to upgrade the plan to the requirements of NUREG 0610, Regulatory Guide 1.101 and 10CFR50, Appendix E.

Neither the design nor the operation of either unit were affected with regards to

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safety analysis requirements.

MILLSTONE UNIT 1 CONTENTS Page CHANGES Design Changes 1/1

- 1/10 Procedure Changes 1/11 TESTS 1/12 - 1/14 RADIATION EXPOSURE 1/15

Page 1/1 PLANT DESIGN CHANGES The following list Ly design change number summarizes those design changes completed in 1979, relating to safety related equiprent, which could have a potential impact on safety related systems, could potentially impact the environment, or required a change to the FSAR.

PDCR l-51 Radwaste Discharge Flow Transmitter The Radwaste Discharge Flow Transmitter was replaced with a new model to increase reliability and accuracy.

PDCR l-48 Annunicator/ Sequence of Events Recorder A sequence of events recorder was added to provide a hard copy history of events during plant operation. The system utilizes existing annunciator points for its input.

PDCR l-7 Gas Turbine Governor Mode Switch A switch was installed on CRP931 which allows the gas turbine governor mode to be changed from isochronous to droop to allow Gas Turbine Generator tie-in following loss of normal power conditions.

PDCR l-28 Spill Barriers for Fire Zones Barriers were installed around oil tanks and lines to limit the spread of oil should a break occur.

This was done as part of the fire protection program to prevent propagation of fire from one zone into another.

PDCR 1-29 Torus to Drywell Vacuum Breaker Modification The stuffing boxes on ten torus-to-drywell vacuum breakers were modified by replacing the packing with bushings and o-rings. This modification will alleviate binding between the valve shaft and packing.

PDCR l-36 Static De-excitation Circuit A generator field static de-excitation circuit was installed to provide back-up protection to the generator against the inability to remove excitation should the field breaker fail.

Page 1/2 PDCR l-46 Turbine Bearing Thermocouoles Thermocouples were attached to the eight turbine bearings to give better indication of bearing temperatures and possible problems.

PDCR l-61 Control Room Annunicator Power Switch / Lights A power-on indicator and a power on/off switch was installed to the annunciator power supplies to provide a power-on indication and a means to turn power on or off while servicing.

PDCR l-78 SJAE Pressure Control The steam jet air ejector pressure controllers were relocated from the SJAE Room to the Condensate Pump Room. This will decrease exposure to personnel performing maintenance.

PDCR l-81 Gas Turbine Generator Rotating Field The Gas Turbine Generator Rotating Field was replaced with a newer design field with a different type of slot insulation to reduce potential for grounding of windings.

PDCR l-97 Control Room Smoke Barrier A smoke barrier was installed between Unit 1 and Unit 2 Control Rooms to meet requirements governing Control Room smoke / fire separation.

PDCR l-99 B4C Fuel Pool Racks The spent fuel pool capacity was increased to 2184 fuel assemblies, utilizing stainless steel and B4C high density fuel storage racks.

To facilitate this modification, two additional modifications were made, the spent fuel pool discharge headers were rotated 90 to clear interfer-

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ences and the high density fuel racks were modified to permit the gas pressure to relieve itself to the spent fuel pool water.

PDCR l-100 Heat Exchanger Inlet Strainers Stainless steel strainers were installed in the inlet heads of various closed cooling water heat exchangers to allow rapid manual cleaning of debris and to prevent large mussel shells from lodging in tubes which could lead to tube failure.

Page 1/3 PDCR l-104 Gas Turbine Start on Turbine Trip / Removal The signal which caused the gas turbine to start whenever the main turbine generator tripped was eliminated.

It was determined that the logic was anticipatory only and elimination did not create a condition adverse to safety.

PDCR l-105 Digital Displays CRP905 Reactor temperature, pressure and electrical output digital meters were

'nstalled on CRP905 to aid operations during evolutions.

PDCR l-107 Drywell-Torus D.P. Indication Drywell-to-torus differential pressure transmitters were installed with a Control Room recorder to facilitate operation with a one pound per square inch pressure differential.

PDCR l-110 Moisture Separator Drain Controls The Turbine Moisture Separator Drain Controls located at the ci n tank were replaced with transmitters.

The new controls are now remott,

located in a low radiation area to allow adjustments to be made wh4'e operating and hence improving reliability.

PDCR l-111 Drywell Head Bolts The Drywell Head Bolts were replaced with a design that allows tensioners to be used for drywell head closure hence insuring containment integrity.

PDCR l-112 TIP Computer Logic The traversing incore probe monitoring computer logic curcuit board was rer. aced as recommended by G.E. Company to improve the quality of the sigt als going to the computer.

PDCR l-114 Gas Turbine Speed Switches The speed switches and tachometers were replaced with a high frequency digital type system to improve reliability.

Page 1/4 PDCR l-117 LPCI Heat Exchanger Head Zinc blocks were added to the upper and lower heads of the LPCI heat exchangers which are fed with salt water via the emergency service water system. The zincs will act as sacrificial anodes and prevent any corro-sion of the heads and attached piping.

PDCR l-123 Safety / Relief Valve Top Works Replacement The six Target Rock main steam safety / relief valves were modified from a 3-stage configuration to a 2-stage configuration.

The newer design valves are. not leakage sensitive hence less prone to inadvertent blowdown.

PDCR l-1 Diesel Generator Breaker Synchronizing Check Relay Installation of a synchronizing check relay on the Diesel Generator output breaker was accomplished to provide a backup to the operator during surveillance testing to insure the Diesel Generator is properly phased.

PDCR l-3 Drywell Sump Discharge Piping Maintenance stop valves were installed in the piping from the drywell sumps located upstream of the isolation valves to facilitate maintenance work on the isolation valves should it be required.

PDCR l-4 Safety / Relief Valve Piping Modifications In order to ensure torus integrity and reduce torus loads the discharge lines from the six safety / relief valves have been modified.

In the torus, the ram's head on each discharge line was removed and replaced with a Tee-Quencher.

In the drywell, another larger vacuum breaker has been added to each discharge line.

PDCR l-11 Feedwater Nozzle Thermocouple Installation Thermocouples have been added externally to the feedwater nozzles.

Temperatures measured will provide data to provde an indication of sparger bypass flow.

Page 1/5 PDCR l-14 Emergency Service Water to Service Tie-in A cross-tie has been installed which will allow the Emergency Service Water System to cool the RBCCW heat exchangers so that Service Water System inspections and repairs can be made during shutdowns.

PDCR l-15 MSIV Byoass and Drain Line The MSIV Bypass and Drain Lines were enlarged to allow the main steam lines to warm-up and equalize across the MSIV's at a faster rate with the reactor at operating pressure.

Change will also allow faster draining of condensate from steam lines to condenser.

PDCR 1 18 Standby Gas Treatment Blower Pumps The Standby Gas Treatment Blower control switches were modified to remove the pull-to-lock maintenanct feature to prevent the blowers from becoming disabled without an indication of the condition other than a visual check of the switch position.

PDCR 1-20 Core Spray Injection Val ces (5A & SB) Logic Change Valves 5A and 5B control wiring was modified so that the valve travel in the closing direction will be terminated by the torque switch when the valve disc is in the seating region. This change was done to provide further insurance that the valve is seated correctly.

PDCR l-22 Drywell Snubber Relocation Three snubbers located in the drywell were relocated because of interfer-ence with the new safety / relief valve top works. This change did not alter their intended purpose.

PDCR 1-28 Stop Valve and Control Valve Test Modification The closing time on the turbine stop and control valves was increased to permit testing of the valves at 100% power without causing unwanted perturba tions.

PDCR l-29 Feedwater Control System The density compensators were removed from the Feedwater Control System to increase reliability and reduce maintenance to the system.

Page 1/6 PDCR l-30 Reactor Buildinq 108 Ft. Level Camera A T.V. camera was installed on the 108 ft. level, Reactor Building, and a CRT monitor in the Control Room to allow monitoring of activities on the Refuel Floor (108 ft. level) by the Control Room.

PDCR l-32 Feedwater Blocking Valves The Feedwater Blocking Valves circuitry and motor operators have been modified to allow the valves to remain closed during startups and still meet FWCI operation requirements. This change will reduce normal leakage into the vessel through the Feedwater Regulating Valves.

PDCR l-33 Fuel Assemblies 148, 3 x 8R new design fuel assemblies were installed during the 1979 refueling outage.

PDCR l-34 LPRM Replacement The LPRM's replaced in 1979 have been replaced with a breeder type (G.E.

or Reuter Stokes) which have a greater life expectancy (7 years) than the older ones.

PDCR l-35 Control Rod Replacement Four test control rods have been installed in the reactor.

This is a research and development effort to determine feasibility of various materials for pin cladding.

No adverse effects ere expected.

PDCR l-36 Primary System Deaeration The mechanical vacuum pump will be used to provide primary system deaeration

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during shutdown and startup periods to lower the oxygen content in the primary system to reduce potential for stress corrosion cracking.

Appropriate piping changes were made to allow this to occur.

PDCR l-37 Recirc. Pump Suction Valve Break Detection Logic Circuitry has been changed in the recirc. pump suction valve logic to insure that the recirc. suction valves remain open in the ev.qt of a LOCA. This change replaces previous administrative controls that called for breaker de-energization.

Page 1/7 PDCR l-42 Hydrogen Analyzer Return Line The Hydrogen Analyzer Return Line was re-routed from the off-gas monitor to an individJal return line to eliminate system interaction and unwanted perturbations.

PDCR l-43 SRM-IRM ' Drive in Button' Modification The SRM-IRM ' drive in button' was converted from momentary action to an alternate action to allow the SRM-IRM detectors to be driven in without having to hold the button depressed.

PDCR l-46 Drywell Floor Drain Pumo Switch The drywell floor drain pump start switch was relocated from behind the Control Room panels to the front panels to allow operators to start the pump without going behind the panels.

PDCR l-49 Cleanuo Demin Effluent Sample Lines Sample lines have been installed and routed to the third floor Reactor Building sample hood to allow remote. sampling in a low radiation area.

PDCR l-50 Stem Mounted Limit Switches on the Main Steam Isolation Valves The stem mounted limit switches on the four Main Steam Isolation Valves located in the drywell were replaced with switches that are environmen-tally qualified.

PDCR l-51 Anchor Bolts Replacement Replaced anchor bolts and base plates on seven safe shutdown systems and inaccessible portions of LPCI, CS, FW and Main Steam Systems to meet criteria of NRC Bulletin 79-02.

PDCR l-52 Gas Turbine and Diesel Generator Status Annunciation The separation of vital and non-vital alanas for gas turbine and diesel generator has been completed.

This provides the operator with information on starting readiness of emergency power supplies.

Page 1/8 PDCR l-55 Auto Trio Tie Breaker 12C/12D on LNP A logic change was implemented to trip the tie breaker between 480 volt buses 12C and 12D to prevent any unnecessary load on the gas turbine should a LNP occur while bus 12D is beino fed from 12C through the cross tie breaker.

PDCR l-59 T-Ouencher Test Instrumentation Pressure transducers were installed in the torus to facilitate determina-tion of torus loads for safety / relief valve discharge through a T-Quencher.

PDCR l-61 Instrument Air Supply Modifications / Valves An air filter, regulator, lubricator and corrosion resistant piping, along with valves, were installed in the instrument air supply lines upstream of certain air opera *. ors to prevent valve sticking due to corrosion products.

PDCR l-62 Service Water /ESW Strainer Instrumentation Modification Tubing was installed on the service water and emergency service water AP

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gage lines.

This will allow plugged lines to be cleared using air or water.

PDCR l-65 Transformer Replacement A new General Electric 4160/480 volt transformer to supply 480 volt bus 12E was installed to replace a tranformer that became burned out.

PDCR l-66 Anchor Supports New anchor supports for the containment spray cooling heat exchangers A and B were installed as a result of recent seismic evaluations.

PDCR l-68 Transformer Replacement A new ITE 4160'480 volt transformer, with 80*C rise, was installed to replace a G.E. transformer, with 150 C rise, to preclude the po;sibility of generic G.E. transformer failures.

Page 1/9 PDCR l-72 Transformer Conduit Passage A hole was cut in the Turbine Building elevation 34'6" floor, for conduit to pass through to feed the cables to a new ITE transformer.

PDCR l-76 Off-Gas Recombiner Cooling Water Orifice A restricting orifice was it. stalled upstream of the conderiser return isolation valve to reduce cavitation at the valve.

PDCR l-79 Gas Turbitte Soeed Switch Noise Suppression A suppression network was added to the Gas Turbine speed switch input to eliminate spurious signals.

PnCP.1-88 Reactor Buildino Vent Radiation Monitor Two Reactor Building Vent Radiation Monitors were relocated to a lower radiation background area. This PDCR did not alter the purpose of the radiation monitors.

PDCR l-97 Loss of Normal Power Logic The loss of normal power logic was modified to correct a condition that was created previously with the addition of a feeder from the Unit 1 4160 volt system to Unit 2.

The change insures that intent of the original LNP logic remains the same.

PDCR l-99 Torus Strain Gage Installation Strain gages were installed on the torus shell to facilitate determination

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of torus shell loads for safety / relief valve discharge through a T-Quencher.

PDCR l-105 Service Water System Seismic Supports Installed temporary seismic restraints on service water piping pending resolution of ' As Built' condition to design documents in accordance with NRC Bulletin 79-14.

PDCR l-106 Seismic Restraints for ECT Sucoly Line Installed temporary seismic restraint on Emergency Condensate Transfer Punp Suction Line pending resolution of 'As Built' condition to design documents in accordance with NRC Bulletin 79-14.

Page 1/10 PDCR l-111 Test Fittings for Feedwater Control Valves Test fittings were added to the feedwater control air system to facilitate feedwater control system testing and future modifications.

PDCR l-113 PORV and SV Position Indicator Instrumentation was added to all six safety / relief valve discharge lines to give the operator additional information alerting him to flow in the lines.

PDCR l-115 Site Technical Support Center A closed circuit camera was installed in the Control Room, and a monitor in the Computer Room and one in the Condensate Polishing Facility to permit remote observation of Control Room instrumentation.

PDCR l-116 Group I Isolation Reset Logic Change A logic reset change was implemented to prevent inadvertent opening of the Main Steam Isolation Valves when a logic reset occurred and containment i-solation conditions had cleared.

PDCR l-117 Stack Area Radiation Monitor A stack area radiation monitor was installed to provide post accident monitoring information for the Emergency Plan.

PDCR l-119 Grcuo 2/3 Isolation Reset Logic Change A logic reset change was implemented to prevent inadvertent opening of the drywell floor drain valves and the drywell equipment drain valves when a logic reset occurred and isolation conditions had cleared.

PDCR l-120 TMI 2.1.4 Containment Isolation Atmosphere Control Valves A logic reset change was implemented for the containment isolation atmospheric control valves. This change prevents inadvertent opening of the valves upon reset of the logic after isolation conditions had cleared.

Page 1/11 PROCEDURE CHANGES No procedure changes as listed in the FSAR in accordance with the provisions of Title 10, Code of Federal Regulations, Sections 50.50, existed during 1979.

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Page 1/12 TESTS The following list by test number summarizes those tests performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59.

None of the tests were evaluated as unreviewed safety questions.

T79-1-1 Pressure transducers were installed in the torus to facilitate determination of torus loads for safety /rclief valve discharge through a T-Quencher.

T79-1-2 This test measured station auxiliary system loads and voltages to verify adequacy of calculated auxiliary system voltages.

This includes measure-ment of voltage, current and power factor for the supplies to each 4.16 KV and 480 volt bus and measurement of the main generator voltage.

These measurements were required with the plant operating at normal full load.

In addition, the main generator voltage and the bus voltage for 480 volt buses E and F.was measured during the starting of core spray pump "B".

During this motor start, the voltage, current and power factor was measured at the supply breaker to the motor.

T79-1-3 This test measured the current and voltage on the secondary of the Emergency Diesel Generator Field Exciter Power Transformer to establish that the transformer was ;perating within its proper nameplate rating.

T79-1-4 This test measured the current and voltage on the secondary of the Gas Turbine Generator Field Exciter Power Transformer to establish that the transformer was operating within its proper nameplate rating.

SP79-1-2 Pressure Leak Test - A pressure leak test was performed of the reactor cleanup system between valves 1-CU-28 (outside drywell) and 1-CV-30 (inside drywell) to prove integrity of all piping, welds or flanges.

Page 1/13 SP79-1-3 Stop Valve Seat Leakage - A pressure leak test was performed to insure that the drywell floor drain and equipment drain piping maintenance stop valves have ainimal seat leakage and packing leakage.

SP79-1-5 LPCI Hydrostatic Testing - This test was to hydrostatically test the Class 2 LPCI system from LPCI pumps discharge check valves to the Class 1 boundary valves 1-LP-10A,1-LP-10B,1-LP-16A,1-LP-16B,1-LP-14A and 1-lP-14B at a pressure equal to or greater than 200 PSIG.

SP79-1-13 Emergency Service Water System Hydrostatic Testing - n hydrostatic test was performed on the "B" Emergency Service Water System, including a portion of new tie-in to service water cystem.

SP79-1-14 Hydrostatic Test of Service Water Lines - This performed a hydrostatic test of the service water lines between the 24" valve (V4-300) and the inlet to the RBCCU heat exchangers, including a portion of the new emergency service water to service water cross-tie.

SP79-1-22 MSIV Bypass Testing - A hydrostatic test of the MSIV bypass and drain piping was performed to verifyno leakage in piping, welds or flanges.

SP79-1-31 LPCI/ Containment Cooling Operating Logic - A test was performed to check the Low Pressure Coolant Injection (LPCI)/ Containment Cooling Operating Logic by proving the LPCI path is capable of closure 10 minutes after the Loss of Coolant Accident (LOCA) and in the presence of an existing LOCA.

Page 1/14 SP79-1-41 Main Steam Safety / Relief Valves Integrity Test - This test was performed to insure the integrity of the air operator assembly on Target Rock model 7567F main steam safety / relief valves 1-MS-3a,1-MS-3b,1-MS-3c, 1-MS-3d, 1-MS-3e and 1-MS-3f.

A visual inspection was performed to verify proper installation of tr.2 air operator assembly and a pneumatic leak check to verify integrity of o-rings and joints.

SP79-1-42 M.G. Set Scoop Tube Linkage - The objective of this test was to collect data from each M.G. set scoop tube linkage and develop a correlation to total core flow with the position of the scoop tube linkage.

SP79.-51 MSIV Group I Isolation Reset - A Group I isolation reset test was performed as a restit of a logic change per PDCR l-116-79.

SP79-1-52 MSIV Group II Isolation Reset - A Group II isolation reset test was performed as a result of a logic change.

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MILLSTONE UNIT 2 CONTENTS Page CHANGES Design Changes 2/1

- 2/7 Procedure Changes 2/8 TESTS 2/9 - 2/13 STEAM GENERATOR TUBE ISI 2/14 - 2/21 RADIATION EXPOSURE 2/22

Page 2/1 S. ANT DESIGN CHANGES The following list by design change number summarizes those design changes completed in 197s. relating to safety related equipment, which could have a potential impact on safety related systems, could potentially impact the environment, or required a change to the FSAR.

PDCR 2-250-76 Modification to the Auxiliary Steam Condensate Radiation Monitor.

This change does not adversly affect any safety related system, improved operability and allowed for cortinuous operation of the radiation monitor.

PDCR 2-13-77 Provision for a permanent test signal of zero percent power for use with the Delta-T power calibration swit' h.

The change does not affect the c

safety requirements of the RPS and improves the calib.'ation procadure.

PDCR 2-25-77 Modification to the RPS TH and TC calibration potenticmeters to narrow their range.

The change improves the RPS temperature related calibration accuracies and does not affect the safety related " Operate' position of the test switch.

PDRC 2-230-77 Installation of a station oscillograph to monitor and record electrical information related to the 6.9 KV and 4.16 KV auxiliary buses and the

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24 Ki raain generator connectors.

The change does not adversly affect any safety related system and provides a means to analyze the causes of any trips or undervoltage conditions which occur.

PDCR 2-275-77 Recoating of the service water side of the RBCCW Heat Exchanger heads with "Carboglass 1678".

This change improved corrosion resistance and was compatible with system materials.

Page 2/2 PDCR 2-282-77 Addition of RBCCW three inch bypass valves on tne Service Water temperature control valves to provide finer temperature control. The change improves operability and does not affect the accident function or response of the RBCCW system.

PDCR 2-13-78 Provision for an Automatic Synchronization Check for Diesel Generator Output Breakers. This change enhances the long term reliability of the Diesel Generators and does not affect the Diesel Generator control circuit's accident functions.

PDCR 2-64-78 Addition of a remote Diesel Generator Disabled Alarm. The change improves diesel annunciation and does not affect any functions for diesel operability.

PDCR 2-78-78 Replacement of the Fire Protection System Test Line Flow Meter with a bellows type gauge tu facilitate calibration. This change does not adversely affect any safety related system and improves the reliability of the system.

PDCR 2-125-78 Relocation of service water pump seal water check valves 2-SW-163/164 in a horizontal piping run. This change enhances reliability by ensuring that the check valves will operate as designed and in no othe" way affects system operations.

PDCR 2-130-78 Relocation of two safety related instrument loop power supplies in control board C03R to provide space for future instrumentation.

This change did not adversely affect any safety related system since the power supplies were only relocated and separation criteria was maintained.

Page 2/3 PDCR 2-152-78 Installation of new seismically qualified support brackets for S/G safety related level channels LT 1113A and LT ll23A as reported in LER 78-26/3L-0 submitted on 11/2/78.

The change upgraded the system to existing design standards.

PDCR's 2-164-78, 2-1-79, 2-8-79 and 2-9-79 Addition of a Reactor Trip to the Reactor Protection System (RPS) for Reactor Coolant Pump underspeed.

This addition was designed and engineared in accordance with the criteria used for existing RPS protective functions and therefore does not adversely affect any RPS function.

PDCR 2-173-78 Additio of an undervoltage relay to the Reactor Trip Switchgear.

This change enhances the reliability of the undervoltage trip setpoint and does not adversely affect the operation of any safety related system.

PDCR 2-11-79 Changing of the Sump Recirculation Actuation Signal (SRAS) setpoint fron-30 inches to 48 + 9 inches, referenced to the bottom of the Refueling Water Storage Tank. This change implemented an approved Technical Specification change.

PDCR 2-21-79 CEA Guide Tube Sleeving of selected tubes - Cycle 3 to support the refueling outage. The installation of sleeves ensures structural integrity of the guide tubes and was approved via the Cycle 3 reload submittal.

PDCR 2-22-79 Installation of a Reactor Cavity Neutron Shield and storage structure.

This change reduced neutron and gamma radiation levels in che Containment and Enclosure Building and does not affect the design function or operability of any adjacent safety related equipment.

PDCR 2-31-79 Allowance for power operation with reduced Guide Tube flow holes.

This change will reduce CEA vibration and Guide Tube wear and was approved via the Cycle 3 reload submittal.

Page 2/4 PDCR 2-41-79 Upgrading of the Containment Nitrogen System to 900 PSIG.

This change facilitates pressurization of the Steam Generators and the Safety Injection Tanks by analysing the system's ability to withstand the increased forces. Compone7ts not suitable for increased pressure are isolated during high pressure operations.

The change does not adversely affect the operation of any safety system.

PD_rR 2-55-79 Installation of a cross-tie from the Turbirc Building sumps to the Condensate Polishing Facility Radwaste Process lng Facility. The cross-tie allows monitoring and processing of potentially contaminated drains in the event of a S/G tube leak and does not affect any safety systems.

PDCR 2-58-79 Increase Refuel Machine Trolly speed. This change conforms to the original design specifications and does not increase the probability or consequences of a fuel handling accident.

PDCR 2-60-79 Modification to the Pressurizer Pressure Loop Wiring.

This change restored the Pressurizer Pressure Loop Wiring to its original design after having the crimary loop isolated from the RPS by a temporary PDCR due to a ground. The safety margin and probability or consequences of an accident remain unaffected.

PDCR 2-67-79 Installation of key lock switches for loop charging header stops, 2-CH-518 and 2-CH-519. This design change will not adversely affect the operation of the system and provides administrative control of a boric acid flow path as part of the Stergency Core Cooling System flow.

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PDCR 2-77-79 Deletion of the Containment Sump Pump Auto-Start feature.

This change prevents the inadvertent transfer of water from the containment sump to radwaste, the potential for gaseous release and does not affect any safety related system.

Page 2/5 PDCR 2-79-79 Replacement of stem mounted limit switches on selected valves inside Containment with environmentally qualified limit switches for post LOCA conditions. This change increases the reliability of the circuits by

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ensuring functionability during the worst LOCA conditions and does not adversely affect the operability of the valves.

PDCR 2-113-79 Auxiliary Feedwater Pump shaf t material specification change.

This change provides an alternate material specification compatable with the original specification and does not degrade the quality of the installed equipment.

PDCR 2-124-79 RBCCW Heat Exchanger Service Water discharge flange changeout. This change does not affect the safe operation of the system and exceeds the original equipment requirements to promote reliability.

PDCR 2-129-79 Replacement of the vital load center transformer for bus 22E (G.E. type) with an ITE transformer as part of a program to replace and refurbish all Unit 2 transformers. This change does not adversely affect the operation of any safety system and will enhance reliability.

PDCR 2-137-79 Replacement of / ital load center transformer 22F with a transformer,

which has a higher temperature rise characteristic. The design change is part of a program to replace and refurbish all Unit 2 transformers and will result in a more reliable transformer installation that exceeds the original design specifications.

PDCR 2-138-79 Replacenent of load center transformer 22B with a transformer that has a higher temperature rise characteristic. This change is part of a program to replace and refurbish all Unit 2 transformers and will result in a more reliable load center transformer installation. The change does not affect any safety system.

Page 2/6 PDCR 2-139-79 Modification to hanger baseplate on the Main Steam line to the turbine driven Auxiliary Feedwater Pump.

This change does not affect the operation of the safety system involved and provides additional support that excee's the original design.

PDCR 2-140-79 Rerouting of Channel 'C' Pressurizer Pressure Loop Wiring.

This design change does not change the function or operability of the circuit and restores the loop to its as-built condition following a grounded lead problem in Containment.

PDCR 2-150-79 Modification to Service Water piping restraint.

This change provided for the use of 4 inch channel, vice 3 inch, for material availability.

The design exceeds the original specifications and does not affect the operation of the safety system.

PDCR 2-159-79 Cable installation for Power Operated Relief Valve and Safety Valve Position Indication. This change installed wiring that is separate from and Joes not affect the operation of any safety system.

This is a NUREG 0578 related installation that will provide indication of these valves to the Control Room operators.

PDCR 2-161-79 Installation of Containment equipment for Power Operated Relief Valve

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and Safety Valve Position Indication. This change provides the Control Room operators with position indication for these valves as response to HUREG 0578.

There is no adverse affect on any safety system as all components are seismically supported.

PDCR 2-179-79 Replacement of Pressurizer Spray Mechanical Snubber.

This change replaced a snubber damaged during maintenance with one that exceeded the require-ments of the original design.

Page 2/7 PDCR 2-181-79 Mounting of a Site Support Center Television Camera in the Control Room ceiling.

This change confirms the seismic adequacy of the installation and does not affect any safety system.

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Page 2/8 PROCEDURE CHANGES The following summarizes safety analyses for procedure changes as listed in the FSAR in accordance with the provisions of Title 10, Code of Federal Regulations, Section 50.59.

Procedure Change 2502 - Emergency Shutdown These four procedures were all changed 2506 - Loss of Coolant Incident to incorporate new requirements to 2509 - Steam Line Rupture verify and correct inadequate core 2515 - Steam Generator Tube cooling.

In addition, the procedures Rupture all require tripping of all four reactor coolant pumps on a low pressure safety injection actuation. The procedure steps result in improved emergency response and are consistent witn safety margins contained in the safety analysis.

2521 - Loss of Feedwater/

This procedure was developed to Steam Generator incorporate guidance from several other procedures to address emergency actions for loss of feedwater. The actions are consistent with system design, previously existing emergency and operating procedures and NSSS vendor's accident response guidance.

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Page 2/9 TESTS The following list by test number summarizes those tests performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59.

A summary of the safety analysis is included for each test. None of the tests were evaluated as an unreviewed safety question.

T78-31 Auxiliary Steam Radiation Monitor Sample System Operability - This test was conducted to verify proper operation of the radiation monitor after the performance of several modifications to put the monitor in operation.

Tne monitor is non-safety related.

T78-35 CEA 42 Position Indication System Full Travel Comparison Test - The test was conducted to verify Technical Specification requirements were met for reed switch position indication following installation of a modified position transmitter. The system was operated in accordance with approved system operating procedures and as permitted by the Technical Specifications.

T78-36 CEA 63 Position Indication System Full Travel Comparison Test - This test was identical to that performed for CEA 42, T78-35.

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Acceptance Test Procedure for Pyrotronics Fire Detection Panels - This test verified tne acceptability of a backfit fire detection panel installed as part of the upgrade of the Millstone Fire Protection System.

Testing was performed in accordance with manufacturer's specifications and system design details.

T79-2 ICI Response Tire Test - This test was performed to obtain response data on two new type incore detectors installed for test purposes during the last refueling. The test involved data taking only and did not affect any safety related systems.

Page 2/10 T79-3 MSI Output Relay Test - The test was parformed to verisy proper actuation relay perfomance for main steam isolation following changeout of the relays. The test was perfomed out of mode requiring MSI operability.

T79-5 Reactor Coolant Pump Speed Sensing System Instrument Preoperational Test-Phase I - The test was conducted to verify equip.aent functionability following installation of the RCP underspeed sensing system and involved testing installed underspeed components only.

T79-6 Pressure Test of Containment Nitrogen Piping - The test was conducted in order to allow pressure upgrading of the containment high pressure nitrogen system and was performed in a mode not requiring containment integri ty.

T79-7 Diesel Load Rejection Test - The test was performed to verify emergency diesel generator load reject capability to satisfy surveillance require-ments.

The test was conducted in accordance with Technical Specification and normal operating procedure requirements.

T79-8 RCP Speed Sensing System Preoperational Test-Phase Two - This test was performed as the final calibration and setup of the reactor coolant pump underspeed trip system and ensured tne proper system response and trip settings in accordance with Technical Specification requirements.

T,79-9 Criticality / Low Power Physics Test, Cycle 3 - This test was part of the Cycle 3 reload startup test program perfomed to verify proper core load and nuclear characteristics. The testing was conducted in accordance with nomal operating procedures and the Technical Specifications, including applicable Special Test exceptions.

Page 2/11 T79-10 Power Ascension Testing, Cycle 3 - The power ascension test was performed as part of the Cycle 3 reload startup test program.

The testing verified proper power range plant per formance, including power dependent nuclear characteri stics.

The power ascension test was conducted in accordance with normal plant operating procedures and Technical Specification requi rements.

T79-ll Neutron Shield Temperature Measurements - This test was performed to determine the acceptability of the shield water temperature following a design change which installed a reactor cavity neutron shield.

The test was for the purpose of data collection only and did not involve interface with any other plant systems or equipment.

T79-12 Startup Radiation Survey Procedure - This test procedure was performed to measure containment radiation levels following installation of the reactor cavity neutron shield.

The test involved data taking only.

T79-13 Turbine Cycle Performance - This test involved sodium tracer injection in various parts of the secondary cycle to determine secondary cycle flow characteristics.

The testing obtained secondary system performance information only and all releases were monitored per plant Environmental Technical Specification requirements.

T79-14 Full Power Operation with Two Condensate Pumps - This test was performed to monitor condensate system performance during power escalation to determine the feasibility of running with one condensate pump in standby.

The system was operated per design and in accordance with normal cperating procedures.

T79-16 Power Ascension to Stretch Power, Cycle 3 - This test was performed to follow power ascension to the new strctch power limit of 2700 MWt and to verify proper nuclear and plant performance. All operations and testing were performed in accordance with Technical Specification and operating procedure requirements.

Page 2/12 T79-17 Ultrasonic Examination of Charging Pump Housing - This testing was implemented to control an ultrasonic examination program of charging pump housing to determine cracking characteristics and degradation. The testing was performed per Technical Specification requirements and did not affect minimum pump operability requirements.

T79-18 Charging Pump Block Hydrostatic Test - This test was performed as a pre-installation test for a replacement charging pump.

The test verified acceptability and was performed in accordance with design and code requi rements.

The following tests were all performed as a part of IE Bulletin 79-17 to complete a visual inspection of piping systems containing stagnant borated water. The systems were operated per normal plant operating procedures and testing involved system inspection and data collection only.

T79 Shutdown Cooling System Leak Test T79 HPSI System Inspection T79 Containment Spray Syster.i Inspection T79 Boric Acid and CVCS Leak Test T79 Spent Fuel Pool Cooling System Leak Test The following tests were performed to identify leakage in ESF systems outside the containment.

Testing involved inspection and data collection only while the systems were operated normally.

In the case of T79-32, the ronitoring system isolation valves remained closed during helium tracer testing.

T79 HPSI and Containment Spray Systems Leak Rate Test

Page 2/13 T79 Shutdown Cooling System Leak Rate Test T79 Leak Test of Containment Air Monitoring and Post Incident Hydrogen Monitoring System T79-25 Collection of Auxiliary System Load Data for IE Bulletin 79 This test was performed to verify the adequacy of calculations used to predict auxiliary system voltages in response to IE Bulletin 79-04. The testing involved data collection only while operating the plant in accordance with normal operating procedures.

T79-27 and T79-28 Diesel Generator 120 and Diesel Generator 13U Field Exciter Power Trans-former Rating - These tests were conducted to verity the diesel generator field exciter power transformers were operating within their nameplate ratings in response to IE Bulletin 79-23. The testing was conducted during normal diesel operation, involved data taking only and did not affect diesel generator ope.rability.

T79-31 ESAS Output for Auxiliary Feedwater Auto Start Circuit Functional Test - This test was conducted to verify correct engineered safeguards actuation system output following completion of a modificction to provide actuation for auto initiation of auxiliary feedwater. The test did not affect the normal performance of the ESAS and was performed in Mode 5.

T79-33 Subcooled Margin Monitor Functional Test - This test was performed to functionally check the subcooled margin monitor installation and verify the adequacy of the installation.

The test did not affect any safety related systems as the monitor provides indication only.

Page 2/14 STEAM GEllERATOR TUBE INSERVICE INSPECTI0ft RESULTS This section provides a summary of the steam generator inservice inspection results for the steam generator tubes, in accordance with Technical Speci fication 4.4.5.1.5.b.

In response to the specific requests for data per Technical Specification 4.4.5.1.5.b:

1.

The number and extent of tubes inspected are included in the summary.

Inspection results for tube denting inspections, in addition to the inservice inspection requirements, are also sumarized.

2.

There were no tubes identifi d with an indication of an imperfection affecting wall-thickness pt utration.

3.

The identification of and reason for the tubes plugged is included in the inspection results.

GENERAL DISCUSSION An Inservice Eddy Current Inspection was performed on Steam Generators Nos. I and 2 at Millstrne Point 2 during the period from March 12, 1979 through April 3,1979. The inspection was performed L'y Combustion Engineering Power Systems Group, Systems Integrity Services personnel.

The inspection was conducted in accordance with Combustion Engineering Test Procedures No. 00000-NLE-082-08, 00000-ESS-073-01 and 00000-ESS-070-02 and satisfied the requirements of the Nuclear Regulatory Commission Guide 1.83 Revision 1 (July 1975) and the Plant Technical Specifications as revised April 14, 1979.

The Eddy Current Test Program was an Inservice Eddy Current Inspection and Dent Assessment Program with the testing frequency set at 400kHz and the test instrument sensitivity adjusted to allow calibration of the test equipment to standards containing known defects and dents.

The inspection pattern, as a minimum consisted of 99% of accest.ible tubes in the tube support plate regions for hot and cold legs of both steam generators, 3% of egg crate tubes for hot and cold legs of both steam generators, exposed peripheral tubes in Steam Generator No.1 and sludge profile in Steam Generators Nos.1 and 2.

All tubes in the support plate regions were tested with an A540F probe, the remaining tubes with an A560SF probe.

Page 2/15 During the conduct of the inspection the signal generated by eddy current probe corresponding to dents was split off of the remainder of the signal by an electronic device called a Peak Reader (DB-1). The output of the peak reader was displayed on one strip chart and the normal high gain defect detection data was recorded on the other strip chart.

Magnetic tape recordings were made of the defect data only.

Both dent and defect data was obtained simultaneously with one pass of the probe from the hot side over the bend to the top two support plates on the cold side. Cold side data consisted of straight length runs.

Dent data analysis was accomplished by comparison to dent standard signals for dents of known dimension for which 1 volt corresponds to 1 mil reduction in radius for an axisymmetric dent.

The defect data was reviewed for anomalous signal:. Such anomalies were evaluated by phase analysis techniques for determination of the cause of the signal.

RESULTS OF THE INSPECTION Steam Generator No. 1 No tube wall flaws were observed in either the hot or cold side in any tube inspected. A total of 2,928 tubes were inspected from hot side and 514 tubes were inspected from the cold side.

This total accounts for 36% of hot side tubes and 6% of cold side tubes.

The Ant assessment results are sunnarized in Tables I and II for Steam Generator No. I hot and cold sides, respectively. The overall a/erages of dent size data for each elevation is presented for the 1978 and 1979 inspection results. These results will be broken down into three categories:

tube sheet indications, egg crate indications and support plate indications.

The tube sheet (Elevation 0) dent-like indications show an increase in occurrence when comparing the 1978 and 1979 data. This increase in frequency occurs in the hot side.

Note that the average dent size changes very little, approximately 0.6 mils for the hot side.

These dent-like signals typically occur just above the start of the tube

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expa.1sion at the secondary side face of the tube sheet.

The suntiary tables indicate an increase in tube sheet dent-like indications.

It is difficult to draw any definitive conclusion concerning changes at the cold side tube sheet. The low incidence of denting and difference in sample size makes comparison with the previous inspection impractical.

The 1979 egg crate dent signals for the nine elevations on the hot side show little or no change in overall frequency or size in comparison with the 1978 data (36.3% vs. 36.7% and 1.06 vs. 1.09 mils).

However, the No. 1 Egg Crate on the hot side does show an increase in frequency of occurrence, 35% in 1978 going to 45% in 1979.

The average size of the observed dents at this elevation does not show any significant change (1.02 vs. 1.07 mils).

Page 2/16 The tube support plate dents at elevations 10 and 11 on the hot and col _d side show no change in size.

The sumary tables for the hot and cold sides of Steam Generator No.1 indicate a decrease in the average dent size. This reduction is due primarily to the fact that many of the largest dents observed in 1978 were plugged.

The sludge pile on the tube sheet face on the hot side was measured.

The maximum sludge height observed was 10.5" which is the same as in the previous inspection (1977).

In addition to the typical magnetite sludge, this inspection indicated bands of high conductivity sludge.

This condition did not exist during the previous sludge inspection (May and November 1977).

The following tubes were plugged as a result of this inspection:

Line Row Reason 50 92 Blocked at No.10 Support Plate, Cold Side 85 83 Cold Side Previously Plugged, Hot Side Open Steam Generator No. 2 A total of 2,261 tubes or 28% were tested on the hot side and 261 tubes or 3% on the cold side.

Data analysis indicates no tube wall flaws in any tube tested.

Tables III and IV summarize dent assessment results for Steam Generator No. 2 Hot and Cold sides respectively. These results are broken down into tube sheet indications, egg crate indications and support plate indications.

Tube Sheet (Elevation 0) dent-like indications show an increase in occurrence when comparing the 197" and 1979 data. This increase in frequer.cy occurs in the hot side and the average size essentially remains the same (+0.08 mils).

These dent-like signals typically occur just above the start of the tube expansion at the secondary side face of the tube sheet.

The summary tables indicate an increase in tube sheet dent-like indications.

It is difficult to draw any definitive conclusion from the summary tables concerning changes at the cold side tube sheet.

The low incidence of denting and difference in sample size make comparison with the previous inspection impractical.

Egg crate dent signals for the nine elevations cr +.he hot side show an increase in overall frequency and size when compz.. ing 1979 results with 1978 data (28.9% vs. 21.1% and 0.88 mils vs.1.19 mils). The most pronounced growth and increase in occurrence is in egg crate elevations 1 through 3.

Page 2/17 For Egg Crate No.1, the frequency of occurrence ciianges from 30% in 1978 to 48% in 1979; Egg Crate No. 2, 64% in 1978 to 80% in 1979; and Egg crate No. 3, 53% in 1978 to 73% in 1979.

Increase in average size of dent signals for Egg Crate No. I was 0.97 mils in 1978 to 1.28 mils in 1979; Egg Crate No. 2, 0.91 mils to 1.19 mils; and Egg Crate No. 3, 0.86 mils to 1.29 mils.

Tube support plate dents at elevations 10 and 11 on the hot and cold side show an increase in size. Summary tables indicate an increase in average dent size of 1.72 mils on the hot side and 1.54 mils on the cold side.

Sludge measurements on the hot side indicate a maximum of 6.2" in 1979 as compared to 6.0" in 1977. Cold side measurements indicate 6.9" in 1979 and 6.0" in 1977.

In addition to the typical magnetite sludge, this inspection indicated bands of high conductivity sludge. This condition did not exist during previous sludge inspection.

The following tubes were plugged as a result of this inspection:

Line Row Reason 80 92 Blocked at No.10 Support Plate, Hot Side 34 100 No Entry at Tube Sheet on Hot Side 40 78 Suspect Tube Sheet Leak 9

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TAllLE I MILLSTONE POINT UNIT 2 STEAM GENERATOR NO. I !!OT SIDE 1978 1979_

No.

1 Avg.

Std.

No.

t Avg.

Std.

Elevation Tested Dented Dent Dev.

Tested Dented Dent Dev.

T.S. 0 2475 22.3 1.51 0.88 2927 40.2 2.13 1.05 E.C. 1 2475 35.0 1.02 0.57 2928 45.7 1.07 0.56 E.C. 2 2475 80.4 1.08 0.54 2928 74.1 1.09 0.51 E.C. 3 2475 86.4 1.40 0.78 2928 75.9 1.24 0.62 U.C. 4 24 7:

57.7 0.88 0.47 2928 52.6 0.98 0.49 E.C. 5 2475 47.4 0.96 0.57 2919 42.9 0.95 0.46 E.C. 6 2475 15.4 0 74 0.36 2911 14.0 0.81 0.39 E.C. 7 2475 4.9 0.81 0.,44 2903 9.6 0.79 0.38 E.C. 8 2435 0.8 0.76 0.78 2745 2.4 0.68 0.37 E.C. 9 2392 0.0 0.00 0.00 2394 0.4 0.60 0.09 TSPL 10 2133 100.0 9.08 2.63 2067 100.0 7.86 2.40 TSPL 11 716 100.0 8.71 2.94 729 100.0 6.96 2.37 E.C. 1-9 22152 36.7 1.09 0.64 25584 36.3 1.06 0.54 TSPL 10-11 2849 100.0 8.9) 2.71 2796 100.0 7.64 2.42 5

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TABLE H MIL 1500NE POINT UNIT.2. STEAM GENERKIDR NO.100LD SIDE 1978 1979 No.

Avg.

Std.

No.

t Avg.

Std.

Elevation Tested Dented Dent Dev.

Tested Dented Dent Dev.

T.S. 0 2996 1.0 2.10 1.14 514 15.6 2.10 0.95 E.C. 1 2996 0.0 0.00 0.00 514 0.0 0.00 0.00 E.C. 2 2996 0.2 0.58 0.20-514 0.0 0.00 0.00 B.C. 3 2996 0.1 0.50 0.00 514 0.4 0.70 0.00 li.C. 4 2996 0.2 0.80 0.73 514 0.2 0.50 0.00 E.C. 5 2996 0.6 0.70 0.34 514 1.4 0.63 0.16 E.C. 6 2996 0.7 0.76 0.40 51 4 1.6 0.86 0.36 E.C. 7 2996 0.2 2.30 0.45 514 1.6 1.34 0.65 E..C.

8 2882 0.0 0.00 0.00 464 0.0 0.00 0.00 E.C. 9 2681 0.0 0.00 0.00 366 0.0 0.00 0.00 TSPL 10 2166 100.0 12.94 2.9.3 2064 100.0 11.19 3.02 TSPL 11 756 100.0 12.83 3.28 727 100.0 10.15 2.99 li.C. 1-9 26535 0.2 0.85 0.59 4428 n.6 0.91 0.49 TSPL 10-11 2952 100.0 12.91 3.02 2791 100.0 10.94 3.04 y

Y t'

TABLE III MILISIVNE POIN1' UNIT 2 S'lT:MI GENEltKlDil NO. 2 Il0T SIDE 1979 1978 No.

Avg.

Std.

No.

1 Avg.

Std.

Elevation Tested _

Dented Dent Dev.

Tested Dented Dent Dev.

T.S. 0 2947 13.50 2.09 1.23 2261 30.6 2.17 1.22 E.C. 1 2947 30.60 0.97 0.59 2261 48.1 1.28 0.84 E.C. 2 2947 64.10 0.91 0.48 2261 79'.8 1.19 0.64 E.C. 3 2947 52.80 0.86 0.48 2750 72.6 1,29 0.66 E.C. 4 2947 22.00 0.75 0.39 2260 28.6 0.92 0.49 E.C. 5 2947 11.40 0.73 0,.44 2260 15.9 0.99 0.55 E.C. 6 2947 2.50 0.70 0.47 2260 4.8 0.83 0.45 E.C. 7 2947 3.20 1.05 0.79 2260 5.6 1.17 0.85 E.C. 8 2801 0.03 1.00 0.00 2135 0.1 0.67 0.29 E.C. 9 2595 0.00 0.00 0.00 2044 0.1 0.50 0.00 TSPL 10 2141 100.00 7.15 2.04 2010 100.0 8.95 2.65 TSPL 11 754 100.00 6.36 2.24 732 100.0 7.80 2.23 x3f

!!.C. 1-9 26025.

21.10 0.88 0.50 20001 28.9 1.19 0.68

'ISPL 10-11 2895 100.00 6.95 2.12 2742 100.0 8.67 2.60 t

O T

TABLE IV MILLSENE POINT UNIT 2 STI%M GENERA'IDR NO. 2 COLD SIDE 1978 1979 No.

Avg.

Std.

No.

Avg.

Std.

Elevation Tested Dented Dent Dev.

Tested Dented Dent Dev.,

T.S. 0 2728.

14.3 0.98 0.58 261 14.6 2.54 1.21 E.C. 1 2728 0.0 0.00 0.00 261 0.00 0.0 0.00 E.C. 2 2728 0.0 0.00 0.00 261

'O.00 0.0 0.00 E.C. 3 2728 0.0 0.50 0.00 261 0.00 0.0 0.00 E.C. 4 2728 0.1 0.50 0.00 261 0.00 0.0 0.00 E.C. 5 2728 0.4 0.75 0.35 261 0.00 0.0 0.00 E.C. 6 2728 0.0 0.00 0.00 261 0.8 1.25 0.35 E.C. 7 2728 0.0 0.00 0.00 261 11.5 1.58 0.63 E.C. 8 2614 0.0 0.00

.0.00 211 0.0 0.00 0.00 E.C. 9 2414 0.0 0'.00 0.00 118 0.0 0.00 0.00 TSPL 10 2139 100.0 9.42 2.10 2008 100.0 11.05 2.83 TSPL 11 770 100,0 8.73 2.03 732 100.0 9.95 2.30 E.C. 1-9 24124 0.1-0.68 0.31 2156 1.5 1.56 0.60 TSPL 10-11 2909 100.0 9.24 2.10 2740 100.0 10.78 2.75 N

R S

, d,

.m.

RADIATION EXPOSURE REGULATORY CUIDE 1.16 REPORT FOR 1979 NORTHFAST NUCLFAR ENERGY Co. t' NIT 2 NW8hR.pF_ PERSONNEL (>100 MREM) _

TOT A L_ MAN,-R F11 WORK & JOS FUNCTION STATION UTILIIY OTHER STATION UTILITY OTHER thPIDYEES FNPl.OYEES EMPIDYFFC FMPIDYEFS mPl5 SEES FMPLOYEES REACTOR OPERATIONS & SURVEILIANCE MAINTENANCE PERSONNEL 4

0 2

1,82 0.52 2.4 '

OPERATING PERSONNEL 31 0

9 12.56 0.10 3.59 HEALTH PHTSICS PERSONNEL I4 3

3) 9.06 4.74 17.41 SUPERVISORY PERSONNEL 0

0 0

0.!!

D. (H) 0.01 ENGINEERING PERSONNEL 4

2 Il 1,77 1.07 6.62 ROUTINE MAINTENANCE MAINTENANCE PERSONNEL 7

1 2

1.76 1.22 1.14 OPERATING PERSONNEL 0

0 0

0.20 0.00 0.01 HEALTH PHYSICS PERSONNEL 0

0 0

0.09 0.00 0.00 SUPERVISORY PERSONNEL 0

0 0

0.01 0.00 0.00 ENGINEERING PERSONNEL 0

0 0

0.00 0.10 0.04 INSERVICE INSPECTION MAINTENANCE PERSONNEL 0

0 1

0.21 0.00 0.77 OPERATING PERSONNEL 0

0 0

0.03 0.00 0.02 HEALTH PHYSICS PERSONNEL 0

0 0

0.05 0.00 0.06 SUPERVISORY PERSONNEL 0

0 0

0.01 0.00 0.00 ENGINEERING PERSONNEL 2

2 26 0.49 1.09 19.51 SPECIAL MAINTENANCE M4tNTENANCE PERSONNEL 39 22 238 39.34 15.43 104.47 OPERATING PERSONNEL 29 0

41 7.80 0.02 14.89 HEALTH PHYSICS PERSONNEL 9

2 12 3.87 1.36 3.72 SUPERVISORY PERSONNEL 0

0 6

0.07 0.00 2.55 ENGINEERING PERSONNEL 9

11 82 2.69 8.92 42.14 WASTE PROCESSING MAINTENANCE PERSONNEL 2

0 0

1.70 0.00 0.06 OPERATING PERSONNEL 3

0 1

1.67 0.00 1.10 HEALTH PHYSICS PERSONNEL 1

1 0

0.42 0.39 0.05 SUFFJtVISORY PERSONNEL 0

0 0

0.00 0.00 0.00 ENGINEERING PERSONNEL 0

0 0

0.10 0.08 0.1)

MAINTENANCE PERSONNEL 19 0

48 5.77 0.13 14.17 REFUELING OPFRATING PERSONNEL 4

0 0

1.54 0.00 0.19 HEALTH PHYSICS PERSONNEL 0

0 0

0.01 0.00 0.01 SUPERVISORY PERSONNFL 0

0 0

0.00 0.00 0.00 ENGINEERING PERSONNEL 0

0 15 0.21 0.21 1.H9 M

D MAINTENANCE PERSONNFL 71 23 291 50.60 17.50 121.21 TOTAL OPERATING PERSONNEL 67 0

51 23.82 0.12 19.MO HEALTH PHYSICS PERSONNEL 24 6

45 13.52 6.49 11.75 SUPERVISORY PERSONNFL 0

0 6

0.21 0.00 2.58 N

)

ENGINEERING PERSONNEL 15 15 134 3.2R l l.4 7 71.l1

~

GRAND TOTAL 177 44 527 93.44 17.sr 239.11