ML20012A133

1989 Annual Operating Rept on Safety Relief Valve Challenges & Design Changes & Tests for 1989. W/900227 Ltr

ML20012A133
Person / Time
Site:	Hatch
Issue date:	12/31/1989
From:	Hairston W GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-981, NUDOCS 9003080355
Download: ML20012A133 (71)

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HL-981 i 000286  ?

February 27, 1990 i U.S. Nuclear Regulatory Commission  !

ATTN: Document Control Desk ,

Washington, D.C. 20555 l

l PLANT HATCH - UNITS 1, 2

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NRC DOCKETS 50-321, 50-366 -

OPERATING LICENSES DPR-57, NPF-5 ,

ANNUAL OPERATING REPORT FOR 1989 'l-4 Gentlemen:

l Enclosed is the Annual Operating Report for Plant Hatch Unit 1. Docket i Number 50-321, and for Plant Hatch Unit 2, Docket Number 50-366. This ,

report is submitted in accordance with the requirements of Technical Specifications Sections 6.9.1.4 and 6.9.1.5. .

If you have any questions in this regard, please call this office at '

any time.

Sincerely,

- .v. 4 Apt-

. G. Hairstc4., III L JKB/eb L ,

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Enclosure:

Annual Operating Report for 1989 i i

c: (See next page.)

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Georgialbwer m( ,

i U S. Nuclear Regulatory Commission February 27, 1990 Page Two l

i c: S praia Power Company a Mr. H. C. Nix, General Manager - Nuclear Plant  ;

Mr. J. D. Heidt, Manager Engineering and Licensing - Hatch  ;

GO-NORMS .,

U.S. Nuclear Reaulatory Commission. Washinaton. D.C.

Mr. L. P. Crocker, Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaion II J Mr. S. D. Ebneter, Regional Administrator i Mr. J. E. Menning, Senior Resident Inspector - Hatch (

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000286

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J ENCLOSURE PLANT EDWIN I. HATCH - UNITS 1 AND 2  ;

NRC DOCKETS 50-321 and 50-366 '

OPERATING LICENSES DPR-57 and NPF-5 ANNUAL OPERATING REPORT FOR 1989 TABLE OF CONTENTS l

PAGE Introduction 1 Safety Relief Valve Chali.enges for 1989 2 Design Changes and Tests or Experiments 3 l

Unit 1 Design Changes (Safety Related) 4 Unit 2 Design Changes (Safety Related) 28 Test or Experiment Requests 53 Data Tabulations and Unique Reporting Requirements Occupational Personnel Radiation Exposure 54 t Reactor Coolant Chemistry 58 L

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P INTRODUCTION Edwin I. Hatch Nuclear Plant is a two unit facility located approximately 11 miles north of Baxley, Georgia on U. S.

Highway 1. The plant consists of two Light Water Reactors licensed to operate at a power level of 2436 Megawatts Thermal each. The Maximum Dependable Capacities for 1989 were 756.6 Net Megawatts Electric for Unit 1 and 767.7 Net Megawatts Electric for Unit 2. General Electric Company furnished the Boiling Water Reactor, Nuclear Steam Supply System, the Turbine and the Generator. The plant was designed by Southern company Services, Inc. with assistance provided by Bechtel Power Corporation. The condenser cooling method used employs induced draft cooling towers and circulating water systems with normal makeup supplies drawn form the Altamaha River. '

The plant is a co-owned facility with ownership delegated as follows:

Georgia Power Company 50.1%

Oolethorpe Electric Memberchip Corporation 30.0%

Municipal Electrical Authority of Georgia 17.7%

City of Dalton, Georgia 2.2%

Licensing information for the units is as follows:

Unit 1 Unit 2 Docket Number 50-321 50-366 ,

License Issued 08/06/74 06/13/78 (DPR-57) (NPF-5)

Initial Criticality 09/12/74 07/04/78 Initial Synchronization 11/11/74 09/22/78 Commercial Operation 12/31/75 09/05/79 Georgia Power Company has .3cle responsibility for overall .

planning, design, construction, operation, maintenance and disposal of this plant.

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Safety Relief Valve Challenges for 1989 i

UNIT 1 i

No Challenges 1 UNIT 2 No challenges s

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i DESIGN CHANGE REQUESTS and TEST or EXPERIMENT REQUESTS -

Pursuant to 10CFR50.59, the following is a brief description and '

surrary of the safety evaluation for each change made to Safety Rel,:ted systems and ccaponents, and each test or experiment performed during 1989. The safety evaluation summaries address the three criteria used to determine if a proposed change, test or experiment involves an unreviewed safety question, i.e.:

1. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report may be increased.

2. If the possibility for an accident or malfunction of a different type than any previously evcluated in the Final Safety Analysis Report may be created.

3. If the margin of safety as defined in the basis for any Technical Specification is reduced.

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! i UNIT 1 DESIGN CHANGE REQUESTS l 81-058 Install, modify, or delete pipe supports por I. E. Bulletin Rev. 2 79-014 design requirements. Revision 2 incorporates all <

79-014 work for Unit 1.

1. Modifications made under this request reduce the actual I stresses associated with safety related system piping and )

, pipe supports to meet the commitments in the Final Safety j Analysis Report.

2. Modifications made under this request introduce no new modes j of failure. j

3. Modifications made under this request increare the margin of safety as defined in the basis for the applicable Technical Specifications.81-174 Provide transient monitoring equipment / instrumentation on the High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems to record system parameters and responses ,

during startups.

1. The installation of performance monitoring equipment increases the operability and availability of the affected systems by ninimizing and improving troubleshooting and maintenance operations. This equipment is provided for l

system monitoring only and does not impact the operation of i the affected systems or their controls.

2. The function of the existing control circuits is not changed and the new design meets the applicable requirements of IEEE 279 and IEEE 308.

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3. The margin of safety as defined in the basis of the applicable Technical Specifications is not reduced.

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82-014 Replace the thermocouple on the Reactor Vessel Bottom Head Drain.

1. The theraccouple provides no control function; only temperature indication, operation of the Reactor Water Cleanup System is not affected.

2. There are no changes to system operating or response functions. No new modes of failure are introduced.

. 3. This modification does not change any control functions.

Therefore, system operation, design base analyses and the unit Technical Specifications are unaffected.83-276 Replace the existing General Electric supplied sling assembly for the refueling floor service platform with an assembly which meets the standards of ANSI B30.9.

1. The replacement of the existing sling assembly with a new qualified sling assembly improves the factor of safety associated with the equipment.

2. The overall dimensions and type of the replacement sling assembly is identical to the existing assembly.

3. The higher safety factor associated with the new sling assembly increases the margin of safety in a lifting operation.84-011 Install a new fire detection system inside the power block Rev. 2 which will ensure cumpliance with the fire codes and 10CFR50, Appendix R.

1. This modification was implemented to comply with the response to Appendix R requirements to upgrade plant safety.

2. The structures have been analyzed to ensure the additional loads are within the applicable seismic criteria.

3. System operation and design as defined in the basis of the applicable Technical Specifications are not affected.

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)84-012 Install a new fire detection system in the Control Building Rev. 2 which will ensure compliance with the fire codes and 10CFR50, Appendix R.

1. This modification was implemented to ensure compliance with the response to Appendix R requirements to upgrade plant safety.  ;

2. The affected structures have been analyzed to ensure the methods of support used for the additional loads satisfy the ,

applicable seismic criteria.

3. System operation and design as defined in the applicable Technical Specifications is not impacted.84-014 Remove the old fire detection system after the new fire Rev. 1 detection system is installed.

1. The old system is superseded by the new system and is no ,

longer necessary. The old system was removed in stages as the corresponding areas of the new system were brought into

, service. The new system satisfies the surveillance requirements of the Fire Hazard Analysis in a manner meeting or exceeding the old system.

2. The new system performs the same functions as the old system and does not result in any different types of accidents or malfunctions than those previously evaluated in the Final ,

Safety Analysis Report.

3. The new system meets the requirements of the Fire Hazard Analysis, which is referenced by the applicable Technical Specifications.84-018 Install a new fire detection system in areas outside the Rev. 1 power block which will ensure compliance with the fire codes and 10CFR50, Appendix R. '

1. This modification was implemented to comply with the response to Appendix R requirements to upgrade plant safety.

2. The supports were seismically designed even though the system itself is not safety related.

3. System operation and design as defined in the basis of the applicable Technical Specificatier.s is not affectea.

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84-019 Install the Multiplex System master panel and printer in the l Control Room and install slave panels and printers in the carbon dioxide tank area on the 147 foot elevation of the '

Control Building. Interconnect the master and slave printers and connect the fire detection systems.

1. This modification was implemented to ensure compliance with  :

the response to Appendix R requirements to upgrade plant safety.

2. The affected structures have been analyzed to ensure the  ;

methods of support used for the additional loads satisfy the applicable seismic criteria.

3. System operation and design as defined in the applicable Technical Specifications is not impacted.85-048 Replace the existing Leeds & Northrup Temperature Recorder on the Control Rod Drive System with a new recorder as the existing recorder is in poor condition, has high maintenance requirements and replacement parts are no longer available.

1. The temperature recorder does not perform any safety-related functions; therefore, its postulated failure does not create an unsafe plant condition.

2. The temperature recorder is not safety related and does not impact systems that are safety related; therefore no new modes of failure are introduced.

3. There is no margin of safety specified in the basis for any Technical Specifications for the measurement of the Ccatrol t Rod Drive System's hydraulic drive water temperature.85-139 Replace the pressure control switch For the Cardox System in Rev. I the Diesel Generator and Control Buildings.

1. The replacement was recommended by the vendor and restores the system to its original condition as the present switch is not of a make supplied by the vendor.

2. This modification introduces no new modes of failure.

3. There is no change to any safety system setting.

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86-097 Remove temporary instrumentation installed on the Main Steam Isolation Valves for monitoring stem vibration.

1. The strain gages that were attached to the exposed part of the valve stem were located such that they did not interfere with normal closure of the Main Steam Isolation Valves. Test data demonstrated that the welds used for attachment had an insignificant affect on the integrity of the valve stem.

2. The removal of this modification adds no new modes of

• operation or failure.  ;

3. This modification does not affect the basis of any Technical -

Specification.86-166 Build up the lower valve guide rib of the B Residual Heat Removal Torus Suction Valve to allow for proper seating of >

the valve disc.

l. This modification restores the alignment of the valve disc to .

the original condition.

2. This modification does not affect the operability of the affected valve but rather ensures the valve integrity, r

3. This modification ensures compliance with the applicable i Technical Specifications.86-268 Provide isolation capability for the Plant Service Water System so as to allow maintenance to be performed on one division while keeping the other division in an operable status.

1. The replacement valves are considered as an icyrade in both design and operation and serve the same funt ion as the -

existing valves.

2. This modification does not alter the operation or ability of the system to perform its safety function. As the replacement valves are heavier than the existing valves, ,

seismic / stress analyses were performed to ensure no new modes of failure were introduced.

3. It was necessary to perform part of this modification in an  !

outage condition due to the system operability requirements.

However, in all cases the applicable Technical Specifications and associated action statements were complied with. The result is enhanced system availability through improved  ;

divisional isolation capability.

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86-210 Upgrade the existing fission product monitoring system with a f replacement (General Electric's NUMAC System) which is compatible with existing equipment connections.

1. The NUMAC log-count rate meter is designed by General Electric as a replacer.ent for the existing log-count rate meters. However, operation of the replacement meters is >

different due to the microprocessor based design which  ;

enhances the system's operation. Interfaces of the t replacement meters with related equipment are the same as the existing meters. This equipment is classified as non-safety related and, as before, the new equipment will not perform any control function. Additionally, a seismic hazard evaluation was performed to assure the integrity of the other ;

equipment in and near the subject panel is not affected.

2. Esplacing the existing equipment with the General Electric designed replacement does not introduce any new failure mechanism or modes. The Fission Product Monitoring System will continue to function as before.

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3. Due to the self-diagnostic capability of the new equipment, maintenance or repair on a unit can be completed within a shorter period of timer approximately three hours.

Replacement with a spare unit will require even less time.

These enhanced design features preclude having to proceed with a plant shutdown as required by the applicable Technical Specifications when the instrument becomes inoperable.86-226 Modify structural supports in the affected areas of the Reactor Building to ensure the Residual Heat Removal System Heat Exchangers are supported in accordance with the Final  !

Safety Analysis Report design commitments.

1. The addition of horizontal restraints or stiffeners to existing stee) beams has no effect on the probability or consequences of any accident. The proposed modifications ensure the beams meet seismic category I requirements. *

2. This modification does not introduce any failure mechanisms u which were previously unanticipated or any events which are l not bounded by the Final Safety Analysis Report.

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3. This modification does nor 1:fect any margin of safety as defined in the basis of the applicable Technical Specifications.

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86-236 Provide normal and emergency lighting to the Analog Transient Rev. 1 Trip System area of the Main Control Room and make necessary modifications to the lighting power sources, suspension mechanisms and affected ventilation ductwork to support the lighting installation.

1. The applicable Final Safety Analysis Report requirements have been determined to be unnecessarily restrictive as they impose requirements for normal lighting levels that are abnormally high. Therefore in the performance of this design, the requirements for emergency lighting levels are used as a reference and normal lighting levels were established at double the emergency levels which are sufficient to provide adequate visibility. The lighting and ventilation ductwork suspension system modifications have been seismically analyzed.

2. There is no change to system operation as this is a restoration to the original design.

3. This modification does not affect the basis of the applicable Technical Specifications.86-375 Remove code Hol or equivalent cables from the General i

Rev. 1 Electric supplied Generator Frequency Monitor to the Main control Room.

1. The penetrations made in the control room wall are sealed in accordance with approved plant procedures. The cables removed were temporary with all connections made to non-safety related systems.

2. The cables were supported per design specifications, therefore their removal does not pose additional concerns. '

This modification does not affect any safety-related systems. All fire barriers were maintained.

3. The affected penetrations were sealed to preclude air flow to the Turbine Building and maintain the positive pressure in the Main control Room.

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1 87-007 Replace the High Pressure Coolant Injection System's turbine <

steam inlet and exhaust drain pot level switches with thermal actuated level switches.

1. The existing level switch was replaced by a current design level switch with the same operational capability. New conduits were added to the Reactor Building. These conduits are supported by new and existing supports. The new supports. ,

were designed as Seismic Category I structures.

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2. The new switch performs the same function as the existing switch, which is annunciation and opening of the bypass )

valve. A loss of power to the electronics will drive the l switch contacts to the alarm condition.  !

3. The operation of these level switches is not addressed in the ;

Technical Specifications and is not critical to the operation of the affected system. t 87-009 Replace the Reactor Core Isolation Cooling System turbine steam inlet and exhaust drain pot level switches with thermal actuated level switches.

1. The existing level switch was replaced by a current design level owitch with the same operational capability. New conduits were added to the Reactor Building. These conduits '

are supported by new and existing supports which are designed as Seismic Category I structures.

2. The new switch performs the same function as the existing switch, which is annunciation and opening of the bypass valve. A loss of power to the electronics will drive the switch contacts to the alarm condition.

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3. The operation of these level switches is not addressed in the Technical Specifications and is not critical to the operation -

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87-061- Provide an opening on the Diesel Generators' bearing oil housing to allow personnel to check the generator bearing insulation as recommended by the vendor.

1. The addition on an inspection plate does not degrade or adversely affect the ability of the Diesel Generators to perform as designed.

, 2. This modification does not change any of the system's l operational or response characteristics.

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! 3. The Technical Specifications are not impacted.

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87-125 Add hard piping and a sight glass to the Reactor Core l Isolation cooling System's high point vent valve to provide i personnel protection when venting the system.

1. The components added meet the design standards of the  :

material to which they are attached. The function of the I injection system and the Reactor Building's drain system have not been changed. This modification reduces the possibility of personnel injury.

2. The ability to mitigate an accident is not reduced below any previously analyzed and there is no function or operational change in the system.  ;

3. This modification does not affect the system's operability or function as defined in the Technical Specifications. It does enhance personnel safety and provide an improved surveillance method for pump operability verification.87-131 Replace the existing overcurrent trip device on the identified 600 volt switchgear with a solid state device.

1. The new trip device improves the breaker's reliability, selectivity, and improves coordination with downstream loads thereby improving the breaker's performance. ,

2. The new trip device has the same function as the device being replaced. The new device incorporates improved design and ,

current technology for improved reliability.

3. The installation of the new device does not change the safety classification of the affected switchgear.

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1. No credit is taken for this radiation monitor in the prevention or mitigation of any accident scenario. The system functions identically as before. A seismic hazard review has been completed to preclude the inclusion of any new hazards.

2. The system functions identically as before.

3. During implementation of this modification, the effluent line was isolated to preclude an undetected release of liquid radioactive waste. The ability to monitor liquid radioactive waste discharged to the environment is enhanced by this modification through the use of more reliable equipment.88-020 Replace the current chlorination system with a liquid sodium hypochlorite addition system.

1. The removal of the_present system eliminates the possibility of a hazardous chemical release. The storage of liquid sodium hypochlorite on site cannot produce sufficient amounts of lethal chlorine gas. Additionally, personnel are not required to rely on the chlorine detectors to detect hazardous conditions.

2. The liquid sodium hypochlorite tanks are located in the existing chlorination structure within a safety dike which is designed to contain the entire inventory. In case of a tank rupture, the release of toxic vapors is negligible. The potential for inadvertent release of chlorine from chemical reactions of stored sodium hypochlorite and acid is eliminated by the use of noncompatible connections and fittings. Administrative controls ensure chemical tank trucks are sampled prior to filling a tank to ensure the right chemical is put into the right tank.

3. The basis requiring isolation of the Main control Room in the event of a chlorine release is superseded by the commitment that no gaseous chlorine will be stored on site.

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88-112 Provide a design for permanent isolation of the reactor )

Rev. 1 building ventilation system from the Hot Machine Shop and the i I

Railway Airlock Ventilation System.

i 1, The fans and ductwork affected by this modification are not I safety related nor are they required to function after an )

accident. Removal of the affected equipment enhances the ability of the existing Heating, Ventilation and Air Conditioning System and Standby Gas Treatment System to l maintain secondary containment.

2. The fans have power removed and the ductwork is isolated by blank flanges and gaskets. No failure modes different from those previously analyzed are introduced by this i modification. ,

3. This modification enhances secondary containment requirements defined by applicable Technical Specifications by reducing air flow from the Reactor Building into the Railway Airlock. ,80-196 Relocate the sensing probe and associated tubing for the refueling floor to outside air differential pressure measurement to a location which will eliminate environmental effects which cause false indications of the outside air pressure.

1. The tubing extension and the new test line does not degrade or affect the ability of the system to perform its intended function.

2. This modification does not degrade the operability of any equipment or prevent any system from functioning as stated in the Final Safety Analysis Report. .

3. The function of the affected instrumentation is not changed and is not addressed in the basis of the unit's Technical Specifications.

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1 88-203 Change the setpoint for the underfrequency relays associated I with the Unit 1 Diesel Generators from 59.5 Hertz to 58.8 Hertz.

1. Closure of the output breakers at 58.8 Hertz does not.

adversely affect the operability of the Diesel Generators or i degrade the ability of the Diesel Generators to supply the ,

required loads during an emergency.

2. This modification reduces the probability of the output breakers failing to close by allowing for reasonable calibration drift.

3. This modification improves the ability of the output breakers i to close thereby enhancing the ability of the Diesel 1 Generators to supply the required loads during an emergency.

1 88-232 Remove and replace any equipment and structures necessary to support the required ten (10) year maintenance of the Residual Heat Removal and Core Spray Systems motors. After the required maintenance is complete, restore the affected equipment and structures to the original design or equivalent conditions. This modification includes, but is not limited to, the addition of eyes for rigging, junction boxes, supports for the motor rotors or any other modification which proves necessary to support the removal and reinstallation of the affected equipment and structures.

1. Equipment and structures that interferred with the required motor maintenance were removed. Restoration was completed to restore, to the extent possible, the original design of the affected systems.

2. Interfaces removed were functionally' tested and nondestructively examined, as required, following restoration.

3. This modification does not affect the basis of any applicable Technical Specifications.

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.- i 88-270 Improve the performance of the Drywell cooling System by Rev.'2 replacing the cooling coils of the existing units and adding isolation valves to the associated Plant Service Water piping

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to improve Local Leak Rate Testing capabilities of the ,

affected Primary Containment Isolation Valves.

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1. The drywell coolers are not safety related; however, the coils are part of the Plant Service Water closed loop boundary inside the primary containment. The coils are l seismically qualified to ensure the structural integrity of )

the pressure boundary. The availability and performance of i the Plant Service Water System during an emergency is not i affected. i

2. Seismic analysis of the new piping configuration assures that

, the affected primary containment pressure boundary is maintained and there are no new modes of failure introduced.

3. Although a containment isolation valve was moved upstream in the piping, the leak rate limits specified in the Technical Specifications remain unchanged. The availability of the Plant Service Water System during normal and accident ,

conditions is not affected. The surveillance requirements  :

for the service water pumps and the containment isolation  :

valves are also unaffected.88-294 Replace the existing overcurrent trip devices in the 600V switchgear with solid state trip devices for improved selective coordination and reduced maintenance.

1. The new device improves selective coordination between the switchgear and associated downstream equipment. They provide more accurate and reliable protection from overcurrent faults.

2. This modification does not alter the function or operation of the affected equipment and results in no change to system

. logic. '

3. There is no change to system operation, function or

! coordination. Safety margins are not affected.

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88-315 Replace the existing isolation check valve for the Standby Liquid Control System to improve the valve's reliability.

1. The design aquirements of the replacement valve meet or exceed those of the previous valve. The design of the Standby Liquid Control System is not degraded by this modification.

2. The pipe stress analysis on the system and the replacement '

valve's seisuic qualification have been reviewed and are found to meet the original system design requirements. )

3. The leakage rates specified in the Technical Specifications remain unchanged.88-343 Improve the supports for the sprinkler system in the Reactor Building by adding stiffeners as required.

1. This modification was implemented to ensure compliance with the response to Appendix k requirements to upgrade plant safety. The addition of stiffeners reduce the stresses on the structural beams.

2. The affected structures have been analyzed to ensure the methods of support used the applicable seismic criteria.

3. System operation and design as defined in the applicable Technical Specifications is not impacted.88-347 Modify the control switches and control panels for the Diesel Generators so as to comply with the guidelines for control room design set forth in NUREG 0700.

1. 'This modification allows personnel to manually control and monitor the operation of the Diesel Generators. It has no affect on the original design requirements or the operability of the Diesel Generators.

2. This modification corrects identified Human Engineering Discrepancies and results in improved operator response if-y manual operation of the Diesel Generators is required during F uccident' conditions. The modifications have been reviewed

! and determined to have no impact on applicable seismic goalifications.

3. Modification were performed on only one diesel at a time and L were governed by system. operability requirements as specified L in the Technical Specifications. The performance of the L Diesel Generators is not affected.

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1. The replacement of the existing starter breaker with a larger size breaker provides coordinated selective protection to the motor circuit which ensures more reliable' operation of the affected Plant Service Water System's strainer motor. The I replacement breaker is qualified to the same requirements as the previous breaker. )

2. Replacement of the breaker does not increase the probability ,

of strainer failure.

3. As the required safety-related loads could have been adequately supplied by one division'of plant service water for the period of time required.to implement this modification, the basis for the Technical Specifications remained unaffected.88-354 Change duplicated circuit numbers on cables that feed the Residual Heat Removal System's heat exchanger outlet valve motor operator and the inboard containment spray valve motor '

operator.

l. This was a documentation change only and involved no physicel change to plant equipment.

2. The retagging of cables introduco no new modes of operation 1 or failure.

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3. This documentation change does not affect the basis of the rpplicable Technical Specifications.

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v  ;88-363 Replace the existing Offgas Post Treatment Radiation Monitors with current design radiation monitors. l i

1. The affected monitors are not safety related, therefore no l credit is taken for the operation of these monitors in the I prevention or mitigation of accident conditions, i L

_2. The function of the system remains unchanged. A seismic hazard analysis was performed to preclude the introduction of new hazards.

3. The new equipment performs the same system-functions and meets the operability requirements of the old equipment.

During implementation of this modification, the offgas line was isolated to preclude any undetected release of >

radioactive gases. The implementation activities were governed by the requirements of the applicable Technical Specifications.

l 89-001 Correct identified Human Engineering Discrepancies by grouping the refueling floor ventilation control switches ,

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together and placing the offgas pretreat chamber purge switch next to the offgas pretreat radiation recorder. This will make this panel identical to the associated panel on Unit 2.

1. This modification is a physical location change only. There is no change to the function, control logic, or operation of-any component involved.

2. This modification does not alter any equipment important to L safety and does not change any system operation or response characteristics.

3. No functional changes were made. The basis of the Technical Specifications are unaffected.

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!89-002 Relocate the Residual Heat Removal Service Water System Rev. 1 pressure indicators to comply with human factors guidelines.

1. The indicators perform the same non safety-related function

l. as before and the system operation and response characteristics remain unchanged.

2. The subject indicators do not perform any safety function.

The relocation of these indicators allow one operator to monitor'and control the system from a single location.

[ 3.- The instruments affected by this modification are not l

l, required by the Technical Specifications and there is no change to any of the functional aspects of the system.

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89-003 Relocate the control switches, indicating lights and flow indicators of.the Feedwater System so as to make the affected panel layout identical to Unit 2.

1. This modification does not change the operation or response l characteristics of the system.

2. This modification improves operator response ability as the orientation and appearance of'the system controls are identical to the other unit.

3. There is no change to any operational or response characteristic of the affected system. Therefore this modification does not affect the basis of the Technical Specifications.09-007 Regroup the controls of the Primary Containment Inerting and l Purging System so as to make Unit 1 identical to Unit 2. l

1. This modification involves no change to function, operation or control logic of any of the affected equipment. It is a physical loce. tion change only. i

2. There is no change to any system operation or response characteristics. An evaluation was performed to verify the i seismic integrity of the panels affected. 1

3. As the equipment function and operability are unaffected, there is no impact on the basis of any of the applicable l Technical Specifications.

L L 89-018 Replace the Reactor Recirculation System's seal flow check valves with nuclear-grade valves of a more current manufacture as the present manufacturer no longer makes i nuclear grade components.

l l 1. The replacement of the subject valves with like kind l-r0 placements does not reduce the availability of water to the pump seals. The results of a pump trip will not be affected  ;

by the seal water supply. The replacement valves meet or .

exceed the original design requirements.  ;

i 2.- The replacement valves meet the service requirements specified for the original valves. Loss of seal water does not result in an accident different than previously  ;

je evaluated.

3. The Local Leak Rate Testing requirements are not affected by this modification.

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I 09-048 . Change the control power source for the path 3 shutdown instrument panel so that the power circuit will not be routed  ;

through Path 3 Fire Area 0024. 1 l

1. This modification changes the power source for non j safety-related instruments only. It has no effect on any  ;

safety-related instruments in the affected panel, or any safety-related equipment function or operation considered in

-the' Final Safety Analysis Report.

2. The change of power source does not affect the operation or response characteristics of existing equipment, ,

3. The change of power source does not affect design setpoints or limits as defined in the Technical Specifications.89-049 Protect identified circuits with an appropriate fire barrier material to ensure compliance with Appendix R requirements.

1. Protection of the identified circuits with appropriate fire barrier material enhances the ability of the affected components / systems to perfora as designed when required.

2. Protecting the identified circuits with a fire barrier ,

material does not introduce any new modes of failure.

3. This modification does not affect the unit's Technical Specifications.

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c 9- i 89-059 Remove'and replace any equipment and structures necessary to ,

- Rev. 1 support the ten (10) year maintenance of the Reactor Recirculation System's pumps and motors. After the '

maintenance is complete, restore the affected equipment-and j structures to the original design or equivalent conditions. )

This modification includes, but is not limited to the 4 addition of' eyes for rigging, junction boxes, supports for

! the motor rotors or any other modification which proves necessary to support the removal and reinstallation of the -

affected equipment and structures. >

1. Equipment and structures that interferred with the required motor maintenance were removed. Restoration was completed to restore, to the extent possible, the original design of the affected systems.

2. Interfaces removed were functionally tested and nondestructively examined, as required, following restoration.

3. This modification does not affect the basis of any applicable Technical Specifications as all components were returned to service in accordance with the original design specifications.

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'89-063 Disconnect the terminal block and perform environmentally qualified inline splices for component wiring in the limit switch compartment of the Residual Heat Removal System's inboard shutdown cooling suction valve.

1. The changes to the cable terminations make the affected circuits more reliable. Environmentally qualified splice components maintain the integrity of the equipment. This modification does not change the design intent, logic or functional operability and capability of the valve.

2. This modification was performed when the unit was in the shutdown / refueling mode with the affected valve in the required (open) position. The seismic qualifications of the valve are not impacted.

3. This modification was performed when the unit was in the-E shutdown / refueling mode with all. fuel removed from the vessel

! and the affected valve in the required (open) position. The -

Technical Specifications are unaffected.

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89-082 Revise the design to allow the Pretreatment/ Flux Tilt Radiation Monitor to be utilized as a backup for the current 1 Pretreatment. Radiation Monitor. Also replace the current i Pretreatment/ Flux Tilt Radiation-Monitor with one of a more l current design. l I

1. The affected system is not a safety-related system and its i function and operation are not.a factor in any accident analysis. The affected components perform no control functions. The incorporation of a backup system improves the  ;

reliability of the system.

2. The function of the affected equipment is to monitor plant conditions and alert control room personnel of abnormal operating conditions. There is no interface with any

-safety-related equipment. ,

3. This modification allows an additional monitor to serve as a backup for the existing monitor. The applicable Technical Specifications are not affected.

I 89-083 Replace the existing Plant Service Water Radiation Monitor with a monitor.of more current design. j

1. The current monitor is not safety related and does not interface with any safety-related systems. The monitor f L

providos control room personnel with information on the ,

radioactivity of the Plant Service Water system effluent.  ?

2. This modification makes no changes to the operation or response characteristics of the affected system._ A seismic l_ hazard review has been completedito preclude the inclusion of p anygnew hazards.

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3. This modification enhances the ability to monitor the Plant

Service Water effluent through the use of more reliable l equipment. The new equipment performs the same function and .

meets the performance requirements of the old equipment.

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(89-085 Replace _the' existing Reactor Building Closed Cooling Water-Radiation Monitor with a monitor of more current design.

l '. The current monitor is not safety related and does not ir.terface with any safety-related systems. The monitor provides control room personnel with information on the radioactivity of the Reactor Building Closed Cooling Water System.

2. This modification makes no changes to the operation or response characteristics of the affected system. A seismic hazard review has been completed to preclude the inclusion of any new hazards.

3. The radiation monitoring ability of the Reactor. Building closed Cooling Water System is not addressed by the Technical Specifications, i 89-098 Upgrade the Main Control Room panel fuses to Class 1E fuses Rev. 1 to facilitate panel uniformity and improve system reliability for fuses other than high fault current Appendix R fuses.

1. The replacement fuses maintain the intended design functions ,

of circuit protection. They are similar in weight and overall dimensions and are Class 1E qualified. The design function and response of systems associated with the affected fuses are unchanged by this modification.

2. This modification improves.the system reliability and

, protection for the affected systems. During implementation, a fuse jumper kit was installed. The installation was evaluated on an individual fuse. replacement basis to ensure the integrity of the system involved is maintained.

3. The use of the jumper kit maintained the margin of safety as defined in the basis of the Technical Specifications of_each system modified.89-100 Provide temporary power and sign &l cables for the operation of the Crack Arrest Verification System.

1. This modification does not affect the operation or reliability of any of the systems involved in the accident L

. scenarios. defined in=the Final Safety Analysis Report. There L is no effect on any safety-related system.

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2. This modification does not change any aspects of system logic or operation which could pose the possibility of a new mode r

of failure.

'3. This change has no effect on the Technical Specifications.

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89-105 Change the time dial settings for the undervoltage relays on the F and G Station Service Busses so as to provide a wider margin between motor starting voltage and the pickup point of the undervoltage relays.

1. Overall operation of the relays on the essential basses are unchanged.

2. This modification'provided a wider margin for voltage drop for large motor loads during the time to start and reach steady state operation thereby increasing operational reliability in the event of an accident.

3. There is no change to the permissible voltage level and time delay for disconnecting offsite power sources defined in the Technical Specifications.89-106 Install test switches which block the actuation signal to the Alternate Rod Insertion System's solenoid valves so as'to allow testing of the system while at power. Replace the trip

• units and modify the logic for the Reactor Recirculation Pump Trip to ensure compliance with 10CFR50.62.

1. This trip unit modification satisfies the requirement for diversity between the Reactor Protection System and the Alternate Rod Insertion System. The modification to the.

Reactor Recirculation Pump Trip logic to a two out of two scheme on the receipt of a high reactor pressure or low moderator. level signal minimizes the potential for inadvertent actuation and provides the capability to perform maintenance, testing or calibration of the_ system while at power. As a result, the operability and reliability of the system is enhanced.

H 2. Any interfaces with safety-related systems utilize the appropriate isolation devices. The affected non safety-related trip system is electrically independent and physically separated from the safety-related Reactor

? Protection System. ,

3. The-setpoints chosen for the Reactor Recirculation Pump Trip are the same as those for the Alternate Rod Insertion System.

These setpoints are within the trip settings of the e applicable Technical Specifications. The logic change from a l

one out of two trip to a two out of two trip requirement i minimizes the possibility of inadvertent ~ actuation and L enhances the ability to maintain the system.

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l I89-118 Provide the ability to reevalutte and recommend setpoint changes for identified high temperature alarms associated 1 with the Heating, Ventilation and Air Conditioning System so as to represent actual climatic conditions and reduce the number of nuisance annunciators.

1. The affected temperature alarms have no effect on analyzed accident scenarios. These alarms serve as notification to control room personnel of temperatures indicative of steam or condensate leaks. The alarms continue to serve this function as they activate whenever temperatures exceed predetermined limits. There le no effect on the qualified life or operability of safety-related equipment.

2. There is no hardware modification associated with this change. The change involves only the periodic reevaluation of-limits which are subject to climatic changes. The-operation and response characteristics of safety-re]ated equipment.ramain unchanged as the actual operating temperatures are unaffected and current administrative controls rely on monitoring and recording of such temperatures rather than annunciation. ,

3. None of tbt affected setpoints are identified in the applicable Technical Specifications. Indicator surveillance and not alarm' response is used to assure compliance.89-162 Provide permanent connection points to the Plant Service Water System such that a temporary cooling system may be installad as the need arises.

1. The branch connections added to the system consists of a gate valve-and a blind flange. A check valve was placed between the gate valve and the blind flange for the~ connection to the portion which supplies the Reactor Building.

l 2. Any failure of the new branch connections or valves would result in the same effects as if another branch or valve of similar size failed. This mode of failure has been addressed in previous analysis.

3. This modification does not- affect the function of the Plant Service Water System or the basis of the Technical L Specifications.

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89-163 Provide temporary cooling to the Plant Service Water System  ;

by adding a temporary chiller system. q 4

l. Any loss of the temporary chiller system discharge into the l safety-related piping will not impact the safety function of 1

-the Plant Service Water System. A check valve and manually l operated gate valve are svailable to prevent loss of Plant Service Water System inventory. Any malfunction of the temporary chiller system would result in temperatures no greater than those presently analyzed.

2. The temporary chillor system is designed such that it cannot fail in a manner which could cause a loss of Plant Service Water inventory. Any missiles generated by the addition of the temporary chiller is bounded by existing analyses.

3. The net effect of this modification is to lower the temperature of the Plant Service Water to the Reactor Building. No margins of safety are defined in the Technical Specifications for minimum Plant Service Water temperature. ,89-227 Install suitable air release valves on the Residual Heat Removal Service Water System as a temporary replacement for the current valves, which have a cast iron body that is susceptible to brittle fracture, until steel bodied air release valves are procured.

1. The majority of the modificaticn is built to the same standards as the adjoining system piping. An exception taken to normal nuclear quality standards is to permit the use of a commercial grade manual valve. The' integrity of the valve pressure boundary was demonstrated with a hydrostatic test prior to placing the valve in service and valve operability will be demonstrated at quarterly intervals. The valve service conditions will be consistent with its service qualifications and it will not be installed in a harsh environment or have any electrical connections. The valve location is such that it will be accessible for manual

-operation.

2. This modification is fabricated to the same standards as the adjoining piping using either qualified parts or parts which have been dedicated to safety-related service using approved plant procedures. The result is the original design basis of the system is maintained and the ability of the system to mitigate postulated accidents is not reduced.

3. This modification does not impair the operability of any item required by the Technical Specifications. Modifications to the system operating procedures wi::1 ensure the system is operable and properly maintained.

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LWIT 2 DESIGN CHANGE REQUESTS 1

79-321 Alter the scaling factor from 0.8 to 0.67 on the return flow input to the Reactor Water Cleanup System's leak detection i summer.

1. This modification makes the system trip at lower leakage i rates.

2. There is no hardware change associated with this modification as only the trip setpoint wrs lowered.

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3. Safety margins are increased due to the lower setpoint.81-174 Provide transient monitoring equipment / instrumentation on the High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems which is capable of recording system parameters and responses during startups.

1. The installation of performance monitoring equipment increases the operability and availability of the affected p' systems by minimizing and improving troubleshooting and maintenance operations. This equipment is provided for system monitoring only, and will not impact the operation of the affected systems or their controls.

2. The function of the existing control circuits is not changed and the new design meets the applicable requirements of IEEE 279 and IEEE 308.

3. The margin of safety as defined in the basis of the Technical Specifications is not reduced.

1-177 Replace valve motors and actuators which are not Rev. 3 environmentally qualified with motors and actuators St.- -

environmentally qualified.

! 1~. Tha replacement and actuators are qualified for all l postulated events.

2. The operatio.. of the replacement motor and actuators is not l different from the previous equipment.

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3. The replacement equipment is subject to the same Technical Specifications as the previous equipment.

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A ,83-089 Make the skimmer surge tank gate cover for the reactor cavity rigid so that'it will not warp when bolted in place and allow water to leak into the dry cavity. .

1. This modification replaced the existing skimn.er surge tank-gate cover with a more rigid cover plate assembly.

2. No new modes of failure are introduced.

3. This modification provides a higher margin of safety by the addition of a new leak tight cover plate assembly.84-014 Remove the old fire detection system after the new fire detection-system is installed.

l.- The old system is superseded by the new system and is no longer necessary. The old system was removed in stages as the corresponding areas of the new system are brought into service. The new system satisfies the surveillance requirements of the Fire Hazard Analysis in a manner meeting or exceeding the old system.

2. The new system performs the same functions as the old system and does not result in any different types of accidents or malfunctions than those previously evaluated in the Final Safety Analysis Report.

3. The new system meets the requirements of the Fire Hazard Analysis, which is referenced by the Technical Specif.ications.84-019 Install the Multiplex System master panel and printer in the Control Room and install slave panels and printers in the carbon dioxide tank area on the 147 foot elevation of the y Control Building. Interconnect the master and slave printers L and connect the fire detection systems.

1. This modification was implemented to ensure compliance with the response to Appendix R requirements to ' upgrade plant safety, p 2. The affected structures have been analyzed to ensure the L methods of support used for the additional loads satisfy the applicable seismic criteria.

l. 3.. System operation and design as defined in the Technical l Specifications is not impacted. The increase of loads on the E affected~ structure due to the system is negligible.

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84-020 Install a new fire detection system inside the power block Rev. 3. which will ensure compliance with the fire codes and 10CFR50, Appendix R.

1. This modification was implemented to comply with the response to Appendix R requirements to upgrade plant safety.

2. The supports were seismically designed even though the system itself is not safety related.

3. System operation and design as defiried in the basis of the applicable Technical Specifications is not affected.

i 84-082 Provide compatible replacements for the identified temperature switches which are no longer available..

1. The replacement equipment provides the same function and is equivalent in configuration and performance to the old equipment.

2. The replacement switches reduce the potential for mercury contamination and do not degrade.the integrity of the affected systems.

! 3. The Technical Specifications are unaffected by this modification.84-138 Incorporate the necessary setpoint changes identified as a l Rev. 1 result of the issuance of Regulatory Guide 1.105.

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1. The design basis uses analytical limits described in the Final Safety Analysis Report as limits from which a setpoint is determined. Once the analytical / design basis limit is -

identified, the allowable value was determined using the L instrument accuracy. Once'the allowable value is determined L the setpoint was established using the drift and calibration tolerance.

2. This modification does not change the purpose or performance j of the affected systems.

3. The Technical Specifications were the basis for some of the L modifications. The new setpoints are more conservative and increase the margin of safety. ,

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85-162 Remove the clutch trippers from the motor operators on the I L

Residual Heat Removal System's test line outboard isolation I valve and the Reactor Recirculation System's pump discharge valve in order to prevent worm shaft gear failures due to l high speed gear ratios. I I

1. The removal of the clutch trippers from the cotor operators '

L of the affected valves does not alter the electrical operation of the valves or modify the performance of the existing motor operators. The only effect is that manual operation of the valves will require holding the clutch lever in the engaged position while manipulating the handwheel.

2. As this modification does not affect the electrical operation or modify the performance of the subject valves, it does not alter the performance of their intended safety functions.

, 3. This modification does not alter the electrical operation or L

response characteristics of the affected valves.

!85-188 Modify the piping to and around the Radioactive Waste System's leak detection switches such that system leakage will.not bypass the switches.

g 1. This is a non safety-related system and has no effect in any L

postulated accident. This modification enhances the performance of the leak detection capabilities, u 2. The enhanced performance of this system does not pose any new I

failure modes.

3. This modification has no effect on the margin of safety as defined in the basis of the Technical Specifications.86-203 Modify structural cupports in the affected areas of the  ;

L: Reactor Building to ensure the Residual Heat Removal System L Heat Exchangers are supported in accordance.with the Final l Safety Analysis Report design commitments.

1. The addition of horizontal restraints or stiffeners to l existing steel beams has no effect on the probability or consequences of any accident. The modifications ensure p compliance with Seismic Category I requirements.

2. This modification does not introduce any failure mechanisms which were previously unanticipated or any events which are

, not bounded by the Final Safety Analysis Report.

3. This modification does not affect any margin of safety as defined in the basis of the Technical Specifications.

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86-286 Revise design to allow the Pretreatment/ Flux Tilt Radiation Rev. 1 Monitor to be utilized-as a backup for the current Pretreatment Radiation Monitor. . Replace the current Pretreatment/ Flux Tilt Radiation Monitor.

1. The affected system is not a safety system and its function and operation is not a factor in any accident analysis. The L affected components perform no control functions. The incorporation of a backup system improves the reliability of the system.

2. The function of the affected equipment is to monitor plant conditions and alert control room personnel of abnormal operating conditions. There is no interface with any safety-related equipment.

3. This modification allows an additional monitor to serve as a backup for the existing monitor. The-basis of applicable Technical Specifications are not affected.86-400 Allow. Stellite 6 valve surfacing material to be used interchangeably with Stellite 21.

1. This modification does not alter the original design intent of the component and it does not affect the function of the component.

2. This modification does not affect the operability of the valves.

3. This modification restored the valve disc hardfacing to satisfy the Local Leak Rate Testing requirements of the Technical Specifications.

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86-431 Replace the existing Posttreatment Radiation Monitors with ones of more current design for improved maintainability and reliability.

1. .The replacement monitors were designed as replacemente for the previous monitors and are compatible with the existing connections and systems. There is no alteration to any syster function or operation.

2. This 74. placement is compatible with the prev.ious equipment and has no adverse impact on the function or operability of any other system. The new monitors are seismically qualified and their installation does not affect the seismic qualification of the panel in which they are to be located.

3. The new equipment was designed as an exact replacement for i the old, and maintains the function and operation of the' system.87-008 Replace the High Pressure Coolant Injection System's turbine j steam inlet and exhaust drain pot level switches with thermal actuated level switches.

1. The old level switch was replaced by a current design level l switch with the same operational capability. New conduits  ;

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.were added to the Reactor Building. These conduits are  !

supported by new and existing supports. The new supports  !

were designed as Seismic Category I structures.

L 2. The new switch performs the same function as the old switch, ,

L which-is annunciation and opening of the bypass valve. A l loss of power to the electronics will drive the switch I contacts to the alarm condition.

3. The operation of these level switches is not addressed in the Technical Specifications and is not critical to the operation of the affected system.

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87-010 Replace the Reactor Core Isolation Cooling System turbine steam inlet and exhaust drain pot level switches with thermal actuated level switches.

1. The old level switch was replaced by a current design level switch with the same operational capability. New conduits were added to the Reactor Building. These conduits are supported by new and existing supports which are designed as Seismic Category I structures.

2. The new switch performs the same function as the old switch, which is annunciation and opening of the bypass valve. A loss of power to the electronics will drive the switch contacts to the alarm condition.

3. The operation of these level switches is not addressed in the Technical Specifications and is not critical to the operation of the affected system.87-071 Replace the electronic controls on the condenser water Rev. 2 control valves of the Turbine Building and Drywell Chilled Water Systems with pneumatic actuator / positioner controllers.

1. This modification increases the control capability of the chilled water systems. Some of the subject valves are safety related. The qualification was maintained to ensure the integrity of the associated piping. As the valves serve no active safety function, the actuators are not required to be safety related.  !

2. Previous accident analysis are unaffected by this modification.

3. The affected systems and components are not discussed in the Technical Specifications.

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. , , l 87-117 Modify the -atpoints, control logic and annunciation of the Standby Gra Tseatment System to prevent incdvertent trips of =

the heateth at? filter train due to residual heat deenergillnq ti.e control relay, and to ensure the system will automatically restart upon a loss of off-site power.

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1. This modification restores the system to its original design basis by a suring automatic initiation during certain accident scenarios. The setpoint increase preserves tha ability of the secondary and fire detection switches to detect excessive heat or fire while eliminating spurious fan and-heater trips due to normal residual-or radiative heat.

The addition of annunciation in the control room enhances the -'

operator's ability to recognize a filter train trip and restore the affected train to operable status and/or start the standby train.

2. The setpoint change.and added annunciation do not introduce any new modes of failure. The replacement temperature switches have similar design and system interfaces and interactions as the previous switches. The reconfiguration of the permissive circuit to provide automatic start capability after power restoration upon loss of off-site power is limited to reversing the normal state of the control  !

relay from energized to deenergized. This could result in a failure to trip the heaters and fans should the relay fail.

However, the fire detection-temperature trip would still be operable and would shutdown the heaters and the fans before the design basis is exceeded.

3. This modification will increase the availability of the system following certain design basis accidents. ,87-152 Upgrade the DOR Qualified High Pressure Coolant Injection system's turbine hydraulic controller with a turbine hydraulic controller qualified to 10CFR50.49.

1. This modification provided a like kind r.3 placement and does not affect the function of the equipment involved. .The l equipment is qualified to the appropriate standards.

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2. This modification provided a like kind replacement and does not affect the function of the equipment involved. The qualification program assures no new modes of failure are introduced.

L 3. This modification is a like kind replacement and introduces no change to the operation or function of the affected system.

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87-181 Replace the existing Radwaste Effluent Radiation Monitor with a more current design in order to improve-its reliability and availability.

'1. No credit is taken for this radiation monitor in the prevention or mitigation of any accident scenario. The operation of the system is unchanged.

2. The system functions identically as before. A seismic hazard review has been completed to preclude the incittion of any ,

new hazards.

3. During implementation of this modification, the effluent.line was isolated to preclude an undetected release of liquid radioactive waste. The ability to monitor liquid radioactive waste discharged to the environment is enhanced through the use of more reliable equipment.87-210 Remove the fire detection system from the drywell.

1. The operation of the fire protection system is not impacted as the equipment in the drywell has been made obsolete by the installation of a nitrogen purge system for primary containment.

2. This modification does not affect primary containment fire protection as defined in the Fire Hazards Analysis.

3. The plant fire detection and proteculon systems remain in compliance with the Fire Hazards Analysis which are referenced by the Technical Specifications.

1,88-018 Modify the location of Core Spray System wiring in the

. control room panels as necessary to meet the divisional ,

L separation requirements of Appendix R.

1. This modification consisted of a change in location of certain circuits and does not involve a change to the operation or response functions of the Core Spray System.

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2. This modification brought the system into compliance with all applicable des 17n criteria and does not affect the operation L or response functions.

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[ 3. This modification does not affect the operability of the system as defined .'n the Technical Specifications.

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88-033 Add a spacer to the Reactor Water Cleanup System's return check valve to-allow proper disc seating.

1. This modification ensures the valve functions as designed thereby maintaining the system's integrity.

2. The function of the valve and system interfaces remain unchanged.

3. This modification ensures proper valve seating and prevents the possibility of reverse flow when the High Pressure Coolant Injection System is operating.

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88-073 Add seal-in circuits to the Source Range Neutron Monitoring i and Intermediate Range Neutron Monitoring controls.

1. Installation of a seal-in circuit increases the probability of detector insertion after a scram by eliminating the need j to continually depress a pushbutton to drive in the '

detectors. This increases the operator's ability to perform i scram recovery functions.  !

2. This modification increases the probability of detector insertion and enhance scram recovery activities.

3. This modification does not adversely affect the basis of any applicable Technical Specifications.88-101 Alter the alarm setpoints of the Drywell Temperature Recorders in order to eliminate unnecessary alarms. I

1. This modification only affected the annunciation of drywell  ;

high temperature and does not prevent the system from --

performing its intended functions.  !

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2. This modification does not degrade any equipment or prevent i any equipment from performing as designed.

3. All temperatures required for accident mitigation will- I continue to be recorded, as required by the Technical Specifications; only the annunciation function is affected.

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~ 88-132 Remove the manual Reactor Vent Valves.

1. This modification removes valves which are redundant to the air operated Reactor Vent Valves. Their removal eliminated loads from the piping system thereby reducing the probability of a piping system failure. The operation of the plant under normal or accident conditions is not~affected.

2. The only types of accidents associated with this modification are a small break loss of coolant and the associated jet which are conditions that have been previously analyzed.

3. This modification was carried out in accordance with the Techniccl Specifications. The modification was included in the plant's Inservice Inspection Program.88-145 Add hard piping and a sight glass to the Reactor Core Isolation Cooling System's high point vent valve to provide personnel protection when venting the system.

1. The components added meet the design standards of the material to which they are attached. The function of the ,

injection system and the Reactor Building's drain system have not been changed. This modification reduces.the possibility of personnel injury.

2. The ability to mitigate an accident is not reduced below that previously analyzed, and there is no function or operational change in the system.

3. This modification does not affect the system's operability or function as defined in the Technical Specifications. It does enhance personnel safety and provides an improved surveillance method for pump operability verification.

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'88-146 Provide cooling to the Low Pressure Coolant Injection Rev. 2 System's Inverter Room ~and the Vital AC Rooms.

l. .This modification added a new Control Building Chilled Water System which is not safety related and does not adversely impact any safety-related system. It also modified the Control Building Heating, Ventilation and Air Conditioning System, the nonessential portion of Instrument Air, and the  !

Demineralized Water System. None of these systems are safety related. As cooling units were installed in each room ventilation, air no longer serves a cooling function and ventilation rates are reduced. However, as batteries are located in the room, the lower ventilation' rate is adequate to prevent hydrogen accumulation above two (2) percent.- The required emergency ventilation rates are not affected by this modification. The onif safety-related equipment affected by this modification is the Plant Service Water System, which supplied the previous cooler.. The connections to this system l were properly isolated and blanked to ensure the integrity of the system's pressure boundary.

2. The modification to the Plant Service Water System does not j

degrade the system's pressure boundary and the removal of "

ductwork does not affect the associated safety-related ductwork supports.

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3. The addition of a system and the modification to the Plant  !

Service Water System and safety-related ductwork supports do )

not impact the basis of the Technical Specifications.  !88-176 Install a support for the D Main Steam Isolation Valve i L solenoid valve junction box, and tag the supports for the other Main Steam Isolation Valve solenoid valve junction  ;

boxes.

1. This modification improved an isolated structural system in .

order-to satisfy current design requirements. I

2. Modification to this support system introduced no new modes  !

of failure.

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3. The Technical Specifications are unaffected by this change. j L i i

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J 88-204 Change the setpoint for the underfrequency relays associated with the Unit 2 Diesel Generators from 59.5 Hertz to 58.8

-Hertz.

l.- Closure of the output breakers at 58.8 Hertz does not

~ I adversely affect.the operability of the Diesel Generators or  !

degrade the ability of the Diesel Generator to supply the required loads during an emergency.

2. This modification reduces the probability of the output breakers failing to close by allowing for reasonable calibration drift. ,

3. This modification improves the ability of the output breakers to close thereby enhancing the ability of the Diesel Generators to supply the required loads during an emergency.88-214 Replace the 0-ring on the Main Steam Isolation Valves' two way air control valves with a viton gasket.

1. The gasket provides the same-sealing function as the 0-ring but will not be able to be dislodged during installation thereby preventing a condition which, in the worst case, could cause closure of the Main Steam Isolation Valve with no means of reopening the valve. The gasket maintains the same seal integrity as the 0-ring.

2. This modification increases the reliability of the sealing function of the 0-ring with no adverse affect on valve operation. The gasket is of the same material as the 0-ring and is qualified for the same environment.

3. This modification does not affect the Main Steam Isolation Valve closure time or seat leakage characteristics.

(: 88-239 Replace.the existing terminal block for the Main Generator Metering Current Transformer with a terminal. block which has sliding links in order to allow the installation of test equipment while the unit is-at power.

1. The affected equipment is not safety related and has no function important to safety.

2. The replacement of the subject terminal block does not affect the seismic qualification of the affected panel.

3. This modification does not impact the basis of any Technical Specification.

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l88-242 ' Relocate.the' Refueling Floor Vent Switches to correct-identified Human Discrepancies.

1. This modification poses no change to control logic of any sort; only the location of the source of control was changed.

2. This modification was a physical relocation of control switches and indicating lights. This relocation reduces the possibility of malfunction by reducing the possibility of operator error.

3. The subject control switches and indication are not addressed in the_ Technical Specifications.88-243 Relocate the Core Spray Reset Switches to resolve identified' Human Discrepancies.

1. The function of the-affected switches will not change; only their location.

2. During implementation the subject reset switches were electrically disconnected from the Core Spray start circuits thereby rendoring the automatic start capability temporarily inoperable. However, the pumps were capable of being manually started and stopped with the normal control switches. The modified system functions the same as before but with reduced possibility of operator error.

3. The Technical Specifications allow the Core Spray System to be inoperable under certain conditions.- By implementing this modification during outage conditions, the basis of the applicable Technical Specifications wab gnot impacted.

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244 Relocate the Residual Heat Removal Service Water System's

.Rev. 2 pump discharge ~ pressure indicators to correct identified Human Discrepancies. Change the scale on the subject ,

-indicators such that the scale divisions are at 10 pound intervals.

1. 'This modification only affects the location and readability 1 of the subject indicators. The is no change to the system's i logic or function. .I l

2. The system's availability during any accident scenario is not I affected. The relocation and scaling change reduce the  ;

possibility of operator error, i

3. The. instruments involved are not required by the Technical Specifications. The pump and system operability is unaffected, f

88-245 Relocate the Drywell Pressure and Torus Water Level Recorder ,

to correct identified Human Discrepancies.

1. The relocation of this recorder improves the operator's ability to monitor the subject values effectively.

2. The relocation of this equipment poses no additional consequences.

3. The function and operability of the affected equipment remain unchanged.

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88-246 Relocate the the Control Rod Drive' System Flow Indicator and the Reactor Feedwater System Flow Indicators within the panel so that instruments of similar function are grouped together.

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1. .This modification improved the operator interface.with plant instrumentation. There is no change to equipment function.

2. There is no change to any equipment function, operation or response.

3. There'is no affect to the Technical Specifications.

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s- i 88-248' Relocate and' add.the identified control switenes to make ,

Rev. l' Unit 1 and Unit 2' Turbine and Balance of Plant Control Panels I as similar as possible~thereby correcting identified Human i Engineering-Discrepancies.

l '. This modification enhanced the panel design in the Main control. Room by regrouping the controls in a more logical manner.

2. No change is being made to any system operation or response characteristics. Only the physical location of the controls ,

was changed. s

3. There are no changes being made to any system operation or response as defined in the Technical Specifications.88-249 Relocate the B Diesel Generator's main tank fuel level indicator and day tank fuel level indicator to correct identified' Human Engineering Discrepancies.

1. There is no affect on equipment important to safety. The .

automatic operation of the diesel remains as previous. Level indication-available locally and at the Unit 1 panel will ensure sufficient fuel is available for diesel operation.

l l 2. No change is being made to the system's operation or functional capability; only the location of the fuel level ,

indication was changed. The level indicators function exactly as prior to this modification.

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! 3. As the fuel transfer pumps will remain operable and alternate l means of level indication are available, the Technical Specifications are unaffected.

1 L 88-250 Reverse the high/ low direction of the Diesel Generator's Rev. 3 speed and voltage control switches, reverse the location of I the start and stop switches at the local panels, and add start and stop switches to the panel in-the Main Control Room so as to correct identified Human' Engineering Discrepancies.

1. This modification does not alter the operation of the diesel.

2. This modification improves operator response should manual operation of the diesel be required. There-is no change to the functional capability of the diesels. The new switches meet all applicable criteria for safety-related equipment.

3. Modifications were performed on only one diesel at a time leaving two diesels operable at any given time. The applicable Technical Specifications are not affected.

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'88-255- Replace the junction boxes for the Drywell High Range j Radiation Monitors with junction boxes which are H environmentally qualified.

1. The installation of new pull box assemblies eliminates the need for additional environmental seals required for the l previous boxes and'provides an enhanced, environmentally l qualified configuration. l

2. The affected monitors are safety related in that they provide

, a post accident monitoring function and a trip for high I drywell radiation. These functions and the associated setpoints are not altered but ensured by this replacement.

3. This modification improves the reliability of the affected monitors without altering their operation. 1 88-296 Replace the existing overcurrent trip devices in the 600V -

switchgear with solid state trip devices for improved selective coordination and reduced maintenance.

1. The new device improves selective coordination between the ,

switchgear and associated downstream equipment. They provide more accurate and reliable protection from overcurrent faults.

2. This modification does not alter the function or operation of the affected equipment and results in no change to system logic.

3. There is no change to system operation, function or coordination. Safety margins are.not affected.88-316 Standardize the outboard Main Steam Isolation Valves' limit switches to a common model.

1. The replacement limit switches are qualified for operation in a worst case scenario environment and perform the exact same  ;

function as the previous limit switches.

2. The limit switch function .is unchanged.

3. Operability requirements as specified in the Technical Specifications are maintained.

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^ ,88-362 ' Relocate the Refueling Range Reactor Pressure Vessel Water Level Transmitter so as to eliminate the reference leg riser between the associated drywell- penetration and the transmitter, i

1. The'new location of the transmitter does not change any of its functions and the seismic qualification of the transmitter is unaffected. The tubing routing is not in a high energy line break area.

2. The new location of the transmitter allows for shorter tubing runs which maintains the proper slope thereby eliminating the possibility of forming air pockets in the sensing lines. The functions of the transmitter remain unchanged.

3. The subject transmitter is not addressed in the Technical l Specifications. However the tubing header modifications downstream of the drywell penetrations affect other i instruments addressed in the Technical Specifications. Due l to.these limitations, the header modifications were done with the plant in a shutdown condition and fuel removed from the j Vessel. I 88-364 Replace the existing Containment Radiation-Monitors with a more reliable model.

1. The replacement monitors were designed as replacements for the previous monitors and are compatible with the existing +

connections and systems. There is no alteration to any system function or operation.

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2. This replacement is compatible with the previous equipment and will have no' adverse impact on the function or operability of any other system. The new monitors are seismically qualified and their installation does not affect l

the seismic qualification of the. panel in which they are i located.

3. The new equipment is designed as an exact replacement for the old and improves the function and operability of the system.

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'89-046 Replace the cable supplying power to the Residual' Heat Removal System's shutdown cooling suction valve with a cable i which will provide adequate current carrying capacity. 1

1. The design and material standards were not decreased from the I original design. .The current carrying capacity is increased and the voltage drop from the power source to the valve  ;

decreased. This ensures the valve is capable of performing l its intended design-function. J

2. The increased cable size allows for more reliable operation of the valve. The existing protective overloads will prevent l

.the additional current capacity from exceeding the i capabilities of the motor operator. ,

3. There are no logic changes associated with this replacement and the functional ability of the valve to perform as designed remains unaltered.89-054' Remove and replace any equipment and structures necessary to support the required ten (10) year maintenance of the Residual Heat Removal System and' Core Spray System motors.

After the required. maintenance is complete restore the affected equipment and structures to the original design or equivalent conditions. This modification may include but is not limited to the' addition of eyes for' rigging, junction boxes,. supports for the motor rotors or any other i modification which proves necessary to support the removal and reinstallation of the affected equipment and structures.

1.

. Equipment and structures that interferred with the required motor maintenance were removed. Restoration was completed to restore, to the extent possible, the original design of the affected systems.

2. Interfaces removed were functionally tested and nondestructively examined, as required, following-restoration.

3. This modification does not affect the basis of any applicable Technical Specifications as all components were returned to L service in accordance with the original design L specifications.

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89-057 . Install a temporary compressor system-and temporary instrumentation necessary for:the performance of an-Integrated Leak Rate Test. Remove the temporary equipment after.the test is completed.

1. This modification was in place while the reactor was in a cold shutdown condition. All changes to the plant were r removed or restored prior to leaving cold shutdown.

2. This modification was to support a test described in the Final Safety Analysis Report and required ~by the Technical Specifications. The affected plant equipment was properly tested before being returned to service.

1-

3. The Technical Specifications are unaffected as the equipment altered by this modification is not required with the plant in the cold shutdown condition.89-099 Upgrade the Main Control Room panel fuses to Class 1E' fuses Rev. 1 to facilitate panel uniformity and improve system reliability for fuses other than high fault current Appendix R fuses.

-1. The replacement fuses maintain the intended design functions of circuit protection. They are similar in weight and overall dimensions and are Class lE qualified. The design function and response of systems associated with the affected fuses are unchanged by this modification.

2. This modification improves the system reliability and y l protection for the affected systems. During implementation a y' fuse jumper kit was installed. The installation was evaluated on an individual fuse replacement basis to ensure the integrity of the affected system was maintained.

3. The use of the jumper kit maintains the margin of safety as defined in the basis of the applicable Technical Specifications of each system modified.

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f 89-122 Upgrade the' Class 1E fuses outside the Main Control Room panel 1 fuses to new Class 1E fuses to facilitate panel ,

uniformity and improve system reliability. l J

1. _The~ replacement fuses fulfill the intended design functions  !

of circuit protection and are qualified to the requirements of the applicable harsh environments. They are similar in I weight and overall dimensions and are Class 1E qualified. The design function and response of systems associated with the affected fuses are unchanged.

2. This modification will. improve the system reliability and protection for the affected systems. The installation was evaluated on an individual fuse replacement basis to ensure the integrity of the affected system is maintained. ,

3. The ratings or types.of fuses are not addressed by the Technical Specifications. There will be no change to equipment' operation.89-185 Remove the internals of the Plant Service Water System's condenser discharge check valves on the A and B Drywell chillers.

1. Removal of the valve internals eliminates the possibility of valve malfunction. The current accident-analyses are unaffected as the valves are required to be in the open position for-system operation.

L 2. The subject valves previously provided an unnecessary l function as identified by INPO in SOER 86-003. The pressure L

of the Plant Service Water System is maintained at higher

. values-than the chilled water thereby ensuring no. ,

uncontrolled radioactive discharges..

L 3. The function previously provided by these valves is not required. The Technical Specifications are unaffected.

48

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, 's . l 89-193 Relocate identified circuits from motor control centers which do not have environmental qualifications commensurate with I the requirements of 10CFR50.49 and I. E. Bulletin 79-OlB to motor control centers which do.

I

1. The new source of power provided by this modification ensures I that equipment can perform as intended during postulated I accident scenarios. .The new sources of power are Class 1E and Seismic Category I qualified. Existing divisionalized

-raceways'were used and cable fills were kept within acceptable. limits. Divisional separation requirements were maintained.

2 . The additional loading of the new power sources is negligible when compared to the capacity of the transformer which feeds them. Voltage drops in the new cables are within acceptable limits thereby assuring equipment operation in the worst case conditions. Electrical protection coordination is not affected by this modification.

l

3. This modification was performed while the unit was in the cold shutdown or refueling mode when equipment affected was ,

not required. There is no change to the control logic, function, or response characteristics.

l 89-199 Replace the A Feedwater Penetration Bellows to improve Local Leak Rate Testing characteristics.

1. The replacement bellows are designed and fabricated in accordance with the applicable ASME Codes and meet the standards applicable to the system and equipment replaced. A 1-Local Leak Rate Test and an Integrated Leak Rate. Test was performed after the installation thereby ensuring the integrity of the drywell pressure boundary.

2. The replacement bellows are fabricated from a material which l

is superior to that of the original. The installation was E done in accordance with the applicable ASME Codes. .The

! functional and operating characteristics of the primary containment and feedwater line remain unaltered.

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3. Primary containment is not required when the unit is in the cold shutdown or refueling modes. This modification was l performed during these modes only.

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89-217 Add seal welds as necessary to the Residual Heat Removal.

System's full 1 flow test valve in order to restore the integrity of the valve body and seat ring surface.

1. This modification does not change the function or operation of the valve.

2. This modification restored the valve to a condition whereby ,

it is capable of fulfilling its intended design functions.

3. As there is no change to the valve performance or function.

The Technical Specifications are unaffected.89-222 Restore the identified terminal block assembly to its  !

original design configuration.

f

1. This modification involved the deletion of a non safety-related raceway and the rerouting of an annunciation '

cable for the Reactor Core Isolation Cooling System's steam injection and steam isolation valves. The operation of the system and annunciation is unaffected.

2. The only equipment'affected by this modification which is safety related is tne electrical penetration assembly which I was restored to its original design configuration. A seismic hazard review has been completed to ensure the integrity of the assembly.

3. The affected annunciation and the electrical penetration are not addressed in the Technical Specifications.89-226 Provide the valves and piping necessary to allow for individual isolation and testing of the Instrument Air System's accumulator check valves while on the unit is on line.

1. This installation does not alter the design intent or system operation. The new piping and valves meet the same requirements as the original equipment and are seismically i qualified.

~

2. This modification allows for the testing of the subject check valves without isolating the header. This new capability will help to ensure the system integrity. System leakage requirements were changed as a result of this modification.

3. The Instrument Air System is not addressed in the Technical Specifications. This modification improves the ability to ensure the integrity of the containment isolation valves serviced by the instrument air lines.

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'19-233' Install suitable air release valves on the Residual Heat Removal Service Water System as a temporary replacement for l the current valves,.which have a cast iron body that is j susceptible to brittle fracture, until steel bodied air j release valves areLprocured. .

1

1. .The majority of the modification was built to the same I standards as the adjoining system piping. An exception was taken to normal nuclear quality standards to permit the use ,

of a commercial grade manual valve. The integrity of the 'l valve pressure boundary was demonstrated with a hydrostatic test prior to placing the valve in service and valve operability will be demonstrated at quarterly intervals. The valve service conditions"are consistent with its service qualifications and it will not be installed in a= harsh environment and does not have any electrical connections.

The valve location is such that it will be accessible for manual operation.

2. This modification was fabricated to the same standards as the adjoining piping using either qualified parts or parts which have been dedicated to safety-related service using approved plant procedures. The result is the original design basis of-the system was maintained and the ability of the system to mitigate postulated accidents is not reduced.

3. This modification does not impair the operability of any item required by the Technical Specifications. Modifications to the system operating procedures ensures the system is operable and properly maintained.89-234 Repair the penetration expansion bellows for the reactor condensate drain piping by adding an additional bellows assembly in order to improve Local Leak Rate Testing characteristics.

1. The new bellows are designed and fabricated in accordance with the applicable ASME Codes and meet the standards applicable to the system and equipment being replaced. A Local Leak Rate Test was performed after the installation thereby ensuring the integrity of the drywell pressure

. boundary.

2. The new bellows are fabricated from a material which is superior to that of the original. The installation was done in accordance with the applicable ASME Codes. The functional and operating characteristics of the primary containment and condensate drain line remain unaltered.

3. Primary containment is not required when the unit is in the cold shutdown or refueling modes. This modification was performed during these modes only.

51

j 89-238 ~ Install suitable air release valves on the Residual Heat Removal Service Water and Plant Service Water Systems as a replacement for the current valves, which have a cast iron s '. . body that is susceptible to brittle fracture.

~

1. This modification allows valves manufactured by another vendor to replace the previous valves. The function and

. operation of the.affected systems are unchanged.

2. As this modification does not involve any electrical or -

control logic changes, the operation of the affected systems is not altered.

3. The function of the new valves is the same as the previous valves. Implementation of the modification was performed in accordance with the restraints and requirements imposed by the Technical Specifications of the affected systems.89-250 Replace the accumulator check-valves on the Diesel Generators Air Start Accumulators with new check valves of improved design to facilitate leakage testing of the valves.

1. This modification provides testing capabilities without altering the design intent or system operation and meets Seismic Category I requirements.

2. The capability to leak test the new valves ensures the affected equipment functions as designed. The original system design criteria and function remain unchanged.

3. The added testing capability improves the system relitbility of the system.

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TEST or EXPERIMENT REQUESTS No tests or experiments were performed in 1989.

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, OCCUPATIONAL PERSONNEL RADIATION EXPOSURE FOR 1989 This section has been compiled to satisfy the requirement of E. I. Hatch Unit 1 and II Technical Specifications Section 6.9.1.5 and to assure compliance with the Code of Federal Regulations as set forth in pertinent sections of Title 10.

Special attention was afforded to the methode prescribed by the-Commission in Regulatory Guide 1.16 in order that the intent as well as the letter of these laws might be fulfilled with providing meaningful information as to the degree and circumstances of all exposure of personnel at this facility. An indication of the effectiveness of the plant radiation program may be inferred from the large number of individuals with no measurable exposure or minimal dose.

The time period covered by this tabulation extended from January i

l. 1, 1989 through December 31, 1989. All monitored personnel were l included in summary as provided under 10CFR20.407. (a) (2) . I Individual exposures as indicated by self-reading pocket ion l

chambers were recorded daily with the use of an ALARA Computer j

System. These exposures were tabulated and printed in hard copy on a weekly basis and when required, along with the difference between these readings and the most restrictive exposure limit.

The corresponding ion chamber results as recorded on the disc dosimetry files were supplanted by thermo-luminescent dosimeter measurements made over a period of approximately one month as the data became available from a vendor.

Each person listed in the dosimetry disc files was assigned a usual job category based on his daily activities. There are six job categories of this nature and they are identified in the following table. Running totals of dose acquired in each of these categories were maintained for each person in his dosimetry

! file. Each dosimeter reading, in addition to being retained for L individual exposure records, is added for individual exposure L records, and is then added to the total representing the cumulative dose in the appropriate job category.

l The implicit assumption involved in this method of accounting for exposure in different tasks is that all exposure acquired in job categories other than the usual will be documented by a radiation work permit. This circumstance should prevail in all significant cases.

54 I

A _ . _ _ - - - - . - - - - - _ _ - _ - - _ - - _ - _ - - - . _ _ _ . - - -- --

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Further delineation to the number of persons and amount of exposure of-people in different job categories by various ,

personnel categories is indicated by the standard reporting format of Regulatory Guide 1.16. Each personnel dosimetry disc file contains the personnel category information required to accomplish this completion. The individual running dose totals for each job were used by_the ALARA Computer to compute the number of= man-rem indicated in each group. Backup disc files were maintained for redundancy in the case of destruction-of temporary inaccessibility suffered by the main files. Hard copy records as printed by the ALARA Computer were also L maintained. -

By the use of the ALARA Computer System dosimetry information has been compiled, retained and tabulated in such a manner as to satisfy the pertinent Federal Regulations and Plant Technical Specifications. The system has been organized to l

provide this information in the format specified by these t- requirements and the suggestions of the Regulatory Guides.

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uSummary of personnel ~ monitoring ending Dec"31, 1989 GEORGIA POWER COMPANY - NUCLEAR GENERATION W T E.I.-HATCH

.O. BOX 439 BAXLEY' GA A1513-DPR-57 2

.NPF <

t

-i Estimated whole body exposure range Nymher of

• (rem) individuals eac in

..... ........... ..... _........... .....----....... --.......h range .

i No measurable exposure .............................. 893 Naasurable exposure less than 0.1 .................. 443 0.1-to 0.25 ........................................ 313' f

0.25 to 0.5 ........................................ 225 0.5 to 0.75 ........................................ 120 '

O.75 to.1.0 ........................................ 108 1.0 to 2.0 .......................................... 128 2.0 to 3.0 .......................................... 12 1

3.0 to 4.0 .......................................... 1

. 4.0 to 5.0 .......................................... 0

.5.0 to 6.0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0 6.0 to 7.0 .......................................... 0 7.0 to 8.0 .......................................... 0 18.0 to-9.0 .......................................... 0-o L 9.0 to 10.0 ........................................ 0 10.0 to 11.0 ........................................ 0 11.0-to 12.0 ........................................ 0 .

12 + ................................................. 0 Total number of personel monitored 2243

, This report-is submitted in accordance with paragraph (a) (2) of 10CFR20.407. l u  !

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'i GEORGIA POWER COMPANY - NUCLEAR GENERATION PLANT E.I. HATCH P.O. BOX 439 BAXLEY, GA $1513 LI81s!I!BF": 5 5 ,

REGULATORY GUIDE '1.16 INFORMATION END OF YEAR REPORT 1989

1. PERSONNEL (>100 MREM) TOTAL MAN-REM WORK & JOB FUNCTION STATION UTILITY CONTRCT STATION U

_____.__ .... _...__._____ ._ ______... .... __...............___TILITY CONTRCT R3 actor Operations & Survgillance Maintenance operations

& Construction 85 3 130 32.573 .778 59.588 53 0 1 20.150 .040 .581 Health P sics & Mb 47 1 33 26.348 .177 16.185 Supervis & Office Staff 23 0 6 11.233 .413 2.860 Engineer Staff 20 1 10 8.364 5.218 g

.__..._____ .______........___........._____.__._____........ .... 397 . . . . . . . _ _ . .

Reuttne Plant Maintenance Mahntenance operations

& Construction 114 0 141 41.825 .126 57.029 22 0 0 9.389 0.000 0.000 Health P sics & Mb 33 0 33 18.329 0.000: 21.144 Supervis & Office Staff 11 0 6 4.660 .419 2.127 Engineer g Staff 9 0 21 3.384

..____ .___....____... ______________..____..___....__ ..........__.186 7.379 Inservice Inspection

-c Maintenance operations

& Construction 16 0 83 6.316 .083 33.651 0 0 0 .528 0.000 0.000 Health Physics & ub 13 0 5 6.600 0.000 1.655 Supervisory & Office Staff Engineering Staff 1 0 1 .612 .231 .559 l 2 1 11 .677 3.922

........_____ ...____._________________..._______.. _______. ..... 186 . . _ _ _ _ _ _ . . . . .

SpecLal Plant Maintenance Manntenance & Construction 105 1. 227 44.081 .526 -89.630 ODerations 41 0 0 15.271 0.000 0.000 Superviso sics & Lab MbalthPhby&OfficeStaff 29 1 34 15.158 .877 18.331 8 2 6 3.024 1.208 1.412 Engineering Staff 8 0 16 3.00

__..________ ....____..._______..______ .______. _______.. ..1 . _ . _ _ _5.193

_ _ .041 '.....

Waste Processing Maintenance operations

& Construction 20 0 32 5.965 .083 13.782 4 0 0 2.148 0.000 0.000 Health P vsics & Mb 9 0 6 4.821 0.000 2.539 Supervis W & Office Staff 0 0 0 .354 .231 .189 l Engineer ng Staff 0 0 0 0.000 .354 l .____ ...._________________________. ___________________.. 135 .

RSfueling Maintenance Operations

& Construction 21 0 41 5.514 .083 16.432 6 0 0 2.500 0.000 0.000 Health Physics & Mb 5 0 10 1.163 0.000 -4.341 Supervisory & Office Staff 2 0 0 .616 .231 .189 l Engineering Staff 1 0 6 .327 0.000 2.382 Totals Maintenance & Construction 361 4 654 136.274 1.678 270.112 Operations 126 - 0 1 49.986 .040 .581 Health. Physics & Lab 136 2 121 72.418 1.054 64.195

l. Supervisgry & Office Staff 45 2 19 20.500 2.734 7.335

, Engineering Staff 40 2 64 15.888 .809 24.447 Grand Totals 708 10 859 295.066 6.315 366.670 l

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, .C REACTOR COOLANT CHEMISTRY  ;

Tabulations on a monthly basis of values of Steam Jet Air Ejector Isotopics and Reactor Coolant-parameters as required by Section 4.6.F.1 of the Unit I Technical Specifications are found in the following tables. Unit II values are also shown i although it is not required that they be reported. Isotopic values which are listed as "0" are less than the low level density limit of the counting system.

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. . 1 Unit I I 1989

'SJAE Isotopics i uCi/SEC

(

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. DATE 1989 lMWP IXe-133 IXe-135 IXe-138 lKr-85m lKr-87 lKr-88 l f6

_________ _ _ p _____ p_______l_ _l________l_ _ ____ p______ p_______l Jan. 2 2435 l3.73E0 15.00El 11.75E3 1.32E1 l1.26E2 l6.06El 12.00E3

____________ _ p_______ p __ p____ ___ ___ l _ _ _ _ p__ _____ p_____ __

Feb. 2 l2436 l3.81E0 15.24El l1.31E3 l1.22E1 l1.05E2 l4.79El 11.53E3

_l ___ p_______l ____p_______l ____p_______p______p_______

Mar. 2 l2436 12.76E0 l4.63El l 1.44E3 l1.49El 11.10E2 14.56El l1.66E3 l _ p ______ p_____ l _ __ __ _ p_______ p __ p______ p________

Apr. 3 l2436 l3.60E0 l4.87El i1.32E3 l1.09El l9.99El 14.36El l1.53E3

_ _ __- __ p______ p_______ p_____ p____ ___ l - __p_______p_______p________

May 8 2434 l3.62E0 14.65El l1.38E3 11.16El 19.88El 13.01El l1.57E3

__ __ l_ __ p__ __ p______ l ____ p ___l - ___ p __l ____

Jun. 5 12436 l3.32E0 14.83E1 l1.23E3 11.04E1 l 9.15El l3.85El. 11.43E3 l- ___ p _____ l ___ l ____ p ____p_______p_______p________

Jul. 3 1 2436 13.77E0 14.07El ll.21E3 l1.40E1 18.54E1 14.40E1 11.40E3

_____________l_ _ ___ p _______ p __ p ____ ___ p_______ p____ ___ p___ - p________ .

Aug. 7 12436 12.14E0 14.89El 11.40E3 11.27El l1.03E2 14.04El -l1.61E3

_____________p. _ p_______ p -p_______p_______p _ ____p_______ p _ _____

Sep. 6 l2436 13.47E0 14.75El 11.25E3 l1.04E1 19.79El 14.20E1 11.45E3

___ p _ p______ p______ p___ _ _ _ p_______ p_______ p _p ____

oct. 4 12436 17.42E1 12.38E2 11.36E3 14.48El l1.80E2 l1.37E2 12.03E3 u ______ _ ____ p______ p_______ p_______ p_______ p_______p ______ p_______ p________

Nov. 2 l2436 l2.26E2 16.16E2 l3.74E3 l1.29E2 l5.08E2 l3.51E2 15.57E3

_ __ _______ p___ ___ p______ _ p __ __ __ p__ _ _ _ _ _ p _____ _ _ p ___ p______ p ____

Dec. 4 l2436 l1.33E2 l3.70E2 l3.4BE3 l6.95El I3.62E2 l2.39E2 l4.66E3

, ____________ p _ ___p_______ p_______ p ______ p ______p_______p_______ p________

REACIOR CHEMISTRY IODINES uC1/ttd DATE 1989 IMWT II-131 lI-132 11-133 II-134 II-135 l DEI-131 L ____ _ ___ _____ l , _.__ ___ p__ ___ p_ _ ____ p_ __ _ ___ p. p_______p__________

i Jan. 2 l2435 16.55E-5 ll.88E-3 l9.75E-4 17.09E-3 12.49E-3 l7.25E-4 l _ ____ _ __i _ p___ _ p _ p_ __ p_____ p___ _p_________

L Feb. 2 l2436 15.15E-5 12.08E-3 19.62E-4 16.68E-3 12.37E-3 l6.98E-4 l

___________p- _p . p_____ p___ l p___ .p ___ ___ _

Mar. 2 '2436 14.52E-5 l2.45E-3 11.15E-3 l7.53E-3 12.37E-3 17.70E-4 -

_ _ _ _ _ _ _ _ _ _ _ _ _ . .___ _ _ p__ _ p_____ _ p___ l ____ p___ p _________

l- Apr. 3 2436 16.13E-5 12.02E-3 l9.74E-4 l6.72E-3 12.48E-3 17.19E-4 l __________ _ p_____p_______ p______ p_______p ______p_______ p_________

l May 8 12434 l5.07E-5 11.97E-3 l8.78E-4 l6.20E-3 12.34E-3 16.60E-4

____________l_ ____ p____ p_______ p______ p_______ p_ - _ p__________

l- Jun. 5 12436 15.70E-5 l2.03E-3 l 3.83E-4 16.03E-3 12.13E-3 15.14E-4

_________p__ p__ _ p_______p_______p_______p______p__________

Jul. 3 12436 l4.10E-5 ll.98E-3 18.55E-4 l5.70E-3 12.21E-3 16.25E-4

____ ___ _ ___ p_ _ __ _ p __ __ __ _ p __ _____ p _ _ _ __ _ p _ _ __ _ p___ _ _ p __ _ _ _ _ _

l Aug. 7 l2436 16.91E-5 12.42E-3 l4.20E-4 l6.36E-3 12.12E-3 15.55E-4

____________p______p_______p_______p_______p___- p ______ p__________

Sep. 6 12436 15.29E-5 11.81E-3 l9.39E-4 15.86E-3 12.17E-3 l6.53E-4

___________p__ _ p____ _ p___ _ _p_______p_______p_______l__________

Oct. 4 12436 ll.09E-4 12.58E-3 ll,28E-3 18.41E-3 13.13E-3 19.52E-4

_____________ p_____ p______ p_______ p_______ p_______p_______ p__________

Nov. 2 12436 l2.50E-3 11.93E-2 19.66E-3 l1.71E-2 11.27E-2 !7.16E-3

_____________ _______ p_______ p_______1________ p_____..p_______l__________

Dec. 4 l2436 11.12E-3 11.07E-2 4.40E-3 1.30E-2 16.33E-3 13.43E-3

_____________1______ p_______ p_______ p_______ ________ p_______ p__________

r

, 3

.c Unit II 1989 SJAE Isotopics

(. uCi/SEC DATE 1989 lMRT IXe-133 lXe-135 lXe-138 lKr-85m (Kr-87 lKr-88 l g 6

__l_ _l ___l-- __l_______l__ __l_ ___l________l Jan. 3 l2436 17.31E0 11.04E2 l2.77E3 l2.16El 11.87E2 17.43E1 ;3.16E3

____ _____j. _l________l________l________l___ __l________l. ___l _____

Feb. 3- 12436 l1.12E1 11.55E2 l 3.98E3 l3.37El l2.74E2 11.13E2 l4.57E3 l_ _l_ -__ _ l _ ___ _ ___ 1.

___l_______l ___l________l_ -____

Mar. 3 l2436 l7.54E0 ll.15E2 1 3.09E3 l2.98E1 l2.20E2 l8.25El l3.55E3

_____________l l________l________________l- _____l_ ___l____ l_ ___

Apr. 4 l2436 l2.22El l2.69E2 l5.77E3 15.21El 14.39E2 l1.69E2 16.72E3 >

_____l l ___ ___ __ l _ __ _ __ __ l - ____l- ___l________l________l________

May 9  : 2352 11.41El l1.46E2 13.65E3 l2.87El l2.72E2 11.00E2 l4.21E3

i. _l___ _ __l________l________l_ ___l_______l_ __l_ - ___

Jun. 6 l2184 l7.20E0 l1.31E2 l2.87E3 12.80E0 l2.05E2 l7.95El 13.32E3

- l _. l________l_ l_______l_ ___ l - ___l- __l___ __

Jul. 4 l2058 l7.28E0 11.02E2 l2.36E3 l1.73E1 11.70E2 l6.42E1 12.72E3

__ __ _ l_

l________l_ _ _ _ l _ _____l ___l ___l________l_ - ___

Aug. 8 l1834 l 6.58E0 l1.09E2 l2.58E3 l2.00E1 l1.66E2 16.62El 12.94E3 l__ _

l ________ l - ____l_______l_ ___l________l________l_ _ _____

sep. 5 0 1 0.00E0 l0.00E0 10.00E0 10.00E0 l0.00E0 10.00E0 10.00E0

_ _ ___l________l_______l________l_ __l________l________l_________

oct. 4 1 0 10.00E0 10.00E0 10.00E0 l0.00E0 10.00E0 l0.00E0 10.00E0 1_ -

_l________l________l________l ____l___ _ __l________l________

Nov. 2 0 l0.00E0 l0.00E0 10.00E0 l0.00E0 10.00E0 10.00E0 l0.00E0 l _____ l ___ ____ l __l________l_ ___ l _ ______ l _ - l_________

Dec. 4 0 l0.00E0 l0.00E0 10.00E0 10.00E0 10.00E0 0.00E0 10.00E0

___ _ ____- lll ___ -

________ l _ ___ __ __ l ___ __ ___ l __ ___l_ _ __ _____

REACTOR CHEMISTRY IODINES UCi/ml DATE 1989 (Mr II-131 lI-132 II-133 -lI-134 II-135 l DEI-131

_______l_______l_______l__ l_____- l________l____ _ l _ _ _ _

Jan. 3 ~l2436 l6.04E-5 l3.44E-3 l1.50E-3 11.06E-2 l3.81E-3 11.09E-3 l____ _ l_____ _ l _____l____ l___ l ____ l __ __ _ _

l2436 l1.35E-4 l3.74E-3 !1.61E-3 11.07E-2 l3.85E-3 l1.21E-3 Feb. _3___ _ l l- l_____ :________l _ _____l_____ _ l Mar. 3 12436 l4.50E-5 l1.42E-3 16.65E-4 13.24E-3 ll.76E-3 14.78E-4

_____________l____ _ l_______l________l_ _l__ _l________l___________

Apr. 4 l2436 ll.28E-4 l2.33E-3 l1.40E-3 17.77E-3 13.32E-3 19.99E-4

__________ _ l_______l____ _ [____ l_____ l________l_____ _ l _ ________

May 9 l2352 18.72E-5 l3.18E-3 l1.20E-3 19.21E-3 12.95E-3 l9.28E-4

__________ _ l____ _ l_______l____ _ l________j________l_ l_______ _ _

Jun. 6 l2184 ll.64E-4 13.62E-3 l1.27E-3 l8.53E-3 l3.19E-3 ll.05E-3

________l____ _ l________l________l________l_______l________l__________

Jul. 4 12058 l5.33E-5 13.59E-3 l5.57E-4 l8.35E-3 l2.88E-3 17.16E-4

____ _ l________l________l________l ____l_ _l _ ____ _ _

Aug. 8 !1834 l8.55E-5 l3.03E-3 17.20E-4 17.06E-3 12.83E-3 l7.46E-4

____________l _l________l___ l________l________l________l__________

sep. 5 1 0 10.00E-0 l0.00E-0 10.00E-0 10.00E-0 l0.00E-0 l0.00E-0

__ _ ________l____ _ l________l________l________l________l________l___________

oct. 4 l 0 l0.00E-0 l0.00E-0 10.00E-0 10.00E-0 10.00E-0 l____ l____ _ _l____ _ l________l________l________l10.00E __________

nov. 2 l 0 10.00E-0 10.00E-0 10.00E-0 10.00E-0 10.00E-0

____________l____ l_____ _ l________l_____ __l_______l________l10.00E ___________

Jemd_ _ AA___lA&R-Mama _fua mma la aseca m emen-- - - -

n zm m - . -

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@ Doao. Equiv. I-131 Percent of Rated y L-

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. _ _ _ m_ . _ - _ _ _

k < j l

DRILY CHEMIt7RY DATA - 1989- l DRTE MWT(1) . DOSE _E0 !!31(1) MWTC2)-DOSE

................... ...... ............... .. .. . EO,1121(2) i i 01-01'-89' 2436 7.9020E-04' 2435 1.2420E-03 1 01-02-89 2435 7.6000E-04 2436 1.1500E-03 j 01-03-49 2436 .7.1100E-04 2436 9.1400E-04 01-04-89 2434 8.7300E-04 2432 1.2400E-03 01-05-89 2436 7.7160E-04 2434 1.1484E-03 J 01-06 2436 8.3000E-04 2435- 1.3400E-03

' 01-07-89 2436 8.0400E-04: 2436 1.0200E-03 01-08 2436 6.8800E-04 2436 9.2200E-04 i 01-09-49 2436 5.2200E-04 2436- 9.8400E-04 '

01-10-89 2436 7.4000E-04 2436 7.9300E 01-11-89 2435 1.9789E-03 2436- 1.2276E-03 '

01-12-89 2435 8.3800E-04 2436 1.1900E-03 ,

01-13-89 2436 8.4508E-04 -2436- 1.2188E-03 01-14-89 2435 8.8800E-04 2436 8.1300E-04 '

01-15 2436 7.9000E-04 2436. 9.0400E-04 01-16-89 2436 7.1300E-04 2436 9.3500E-04 01-17-89 2436 6.8000E-04 2436 9.1400E-04 ~'

01-18-89 2436 7.9000E-04 2435 1.1300E 01-19 2435 7.9100E-04 2435 1.2300E-03 01-20-89 2436 7.9052E-04 2436 1.1500E-03 01-21-89 -2436 8.4700E-04 2050 1.1500E-03 01-22-89 2436 8.7100E-04 2436 1.3900E-03 01-23-49 2436 7.3150E-04 2436 1.3070E-03 01-24 2433 8.4200E-04 2436 1.2900E-03 01-25-89 2432 6.1800E-04 2436 8,2400E-04 01-26-89 2436 7.5900E-04 2436 9.4600E-04 u

01-27-89 2436 7.1600E-04 2436' 8.7800E-04 '

01-28-89 2436 7.9100E-04 1491 5.5700E-04 01-29-89 2436 6.8300E-04 2436- 9.2600E-04 01-30 2436 7.6000E-04 2436 9.5900E-04 2436 1

01-31-89 7.2500E-04 2436- 1.2200E-03 02-01-89 2436 7.8100E-04 2436 1.0900E-03 0 02-02-89 2436 7.0300E-04 2425 9.0500E-04 02-03-89 2436 7.5600E-04 2436 9.7500E-04 02-04-89 2436 1.1419E-03 2433 1.5709E-03 02-05-89 2436 8.1900E-04 2423 -1.2470E-03 02-06-89 2436 8.6400E-04~ 2430 1.5200E-03 02-07-89 2436 8.5600E-04 2436 1.2200E-03 02-08 2436 8.8500E-04 2436 9.2700E-04 02-09-89 2436 -1,3100E-03 2436 1 2500E-03 02-10-89 2436 7.9600E-04 2436- 1.1300E-03 02-11-89 '2435 8.3000E-04 2436 1.2200E-03 J 02-12 2436 9.9700E-04 2436' -1.2500E-03 02-13-89 2433 8.0720E-04 2429. 9.0930E-04  ;

02-14-89 2435 7.9600E-04 2432 1.2800E-03 02-15-89 2433 8.6200E-04 2434- E.7000E -

02-16-89 2436 8.2000E-04 2436 1.8100E-03 02-17-89 2436 5.0000E-04 2436 6.9200E-04 02-18-89 2436 7.5100E 2436- 1.0900E-03 -

02-19-89 2425 7.6100E-04 2434 1.0400E 02-20-89 2435 7.2100E-04 2436 7.8100E-04 '

02-21-89 2435 7.4700E-04 2436 1.0500E-03 02-22-89 2436 8.0990E 2436 1.1600E-03 02-23-89 2434 8.1100E-04 2434 1.2400E-03 02-24-89 2436 7.7900E-04 2436 1.2200E-03 02-25-89 1566 7.0400E-04 2433 8.1100E-04 i

E 02-26-89 2315 7.8765E-04 2103 1.0942E-03 02-27-89 2436 8.1900E-04 2434 1.0600E-03 02-28-89 2436 9.4080E-04 2433 1.5260E-03 L

pl

. + _, .- _ . - - . -

~. . . - . . .- --- -. --_ -,_ ---,

A-

. (

o .l

, DRILY CHEN!!TRY DATR - 1999 9

D ATE MWT M ) ~ DOSE.,EO,! ! 31 ( 1 ) MWTC2) DOSE,E0_!!31(2)

. -p ................... . . . . . .......... .. . ...... ...............

03-01-89 2434 8.4504E-04 2436 1.1154E-03 03-02-89 2436 7.9817E-04 2432 1.1857E-03 D

03-03 2436 7.9300E-04 2436 9.7739E-04 9.7600E-04 2436 03-04-89 2436- 7.8500E-04 1 b- 03-05-89 2436 9.0300E-04 2436 1.0900E-03 1 D 03-06-89 2436 7.6100E-04 2436 -

1.0000E-03 03-07-89 2435 8.Ot00E 2436 6.6600E-04 J 03-08-89 2430 8.6110E-04 2426 4.1508E-04 1 03-09-89 2436 7.6740E-04 2433 6.9180E-04 03-10-89 2435 8.8494E-06 1990 8.3734E-04 1 03-11-89 4436' 7.7400E-04 2431 9.3500E-04 i 03-12-89 2436 8.6600E-04 2436 6.5300E-04 03-13-89 2436 8.0600E-04 2436 7 .110 0E- 04 03-14-89 2434 7.4000E-04 2436 7.e.000E-04--

03-15-89 2436 7.6300E-04 2436 8.2400E-04 03-16-89 2436 8.0209E-04 2436 7.5tSOE-04 03-17-89 2434 7.80T4E-04 2434 8.5788E-04 03-18-89 2436 7.9200E-04 2436 8.8300E-04 03-19-89 2436 -7.3900E-04 2d35 6.3900E-04 1 03-20-89 2436 7.3500E-04 2436 7.2900E-04  ;

03-21-89 2436 7.3900E-C4 243E 8.2500E-04 03-22-89 2435 7.5200E-04 2436 8.2500E-04 03-23-49 2436 7.4854E-04 2436 7.9246E-04 03-24-89 2435 6.0600E-04 2436 '7.1500E-04  !

03-25-89 2436 7.3700E-04 2158 6.2200E-04 03-26-89 2436 7.0000E-04 2436 5.1200E-04 03-27-89 2436 7.1800E-04 2436 5.8800E-04 03-28-89 2436 7.4300E-04 2436 7.7000E-04 03-29-89 2435 6.9636E-04 3436 5.3718E-04 03-30-89 2436 5.7100E-04 2436- 6.2700E-04 03-31-89 2435- 5.4000E-04 2436 9.0400E-04

'4 04-01-89 2431- 7.0851E-04 2436 1.0175E-03 04-02-89 2436 5.-3320E-04 2436 1.1639E-03 04-03 2105- 8.1290E-04 2436 1.1791E-03 04-04-89 2436 7.3659E-04 2436 7.8917E-04 04-05'89 2436 Y.2700E-04 3436 1.0800E-03 04-06-89 2436 7.6800E-04 2436 9.0900E-04 04-07 2140 7.3500E-04 2436- 1. 0 6 0 0E-0 3 -

04-08-89 2434 8.6400E-04 2436 8.2500E-04 04-09-89 2436 7.0400E-04 2070 9.3500E-04 04-10-89 2436 6.8000E-04 760 4.8900E-04 04-11-89 2436 7.9600E 747 4.5800E-04  ;

04-12-89 2436 7.7900E-04 695 '4.5400E-04 .

04-13-89 2433- 6.7900E-04 697- 4 6200E-04  ;

04-14-89 2436 6.5800E-04 1183 3.3600E-04 4 04-15-89 2436 4.0300E-04 2436 1.4100E-03 04-16-89 2436 -7.2970E-04 2435 1.1440E-03 04-17-89 2436 7.7200E 2436 9.7600E-04 04-18-89 2436 7.2677E-04 2436 8.0296E-04 04-19-89 2432 1.0500E-04 2433 1.0700E-03 04-20-89 2433 4,8000E-04 2436- 1.0800E-03 04-21-89 2436 7.0900E-04 2435 1.0100E-03 04-22-89 2435 7.0700E-04 2435 9.9700E-04 04-23-89 2436 7.7300E-04' 2436 9.4600E-04 04-24-89 2436 8.4000E-04 2403 1.0800E-03 04-25-89 2436 7.1200E-04 2413 .7.7500E-04 04-26-89 2436 7.2100E-04 2402 9.9400E-04 04-27-89 2436 7.4900E-04 2400 1.0500E-03

'04-28-89 1989 7.3800E-04~ 2394 9.1000E-04 04-29-89 750 3.1060E-04 2400 1.0590E-03 L 04-30-89 777 3.8067E-04 2394 1.0095E-03 l .

l

j 4 j I

!i : ]

DRILY CHEMIE1RY DATR - 1989 DATE MWT(1) DOSE _EQ _  !!31(1) MWTC2) DOSE _E0_!131(2)

( '

05-01-89 830 3.4030E-04 2380 1.1230E-03 0$-02-89 950 3.3300E-04 2368 1.2900E-03 05-03-89 2430 6.9300E-04 2374 1.3900E-03 i 05-04-89 2433 1.1000E-03 2364 6.2500E-04 '

05-05 2436. 6.7610E-04 2360 9.8297E-04 05-06-89 2434 6.8900E-04 2358 1.0100E-03 05-07-89 2435 4.6670E-04 2354 7.9540E-04 I 05-08-49 2436 7.8100E 2345 9.4500E-04 05-09-89 2436 7.8200E-04 2345 9.0800E-04 ~1 05-10-89 2432 7.3400E-04 2333 1.1440E-03 05-11-89 2472 6.7500E-04 2335. 9.3700E-04 J 05-12-89 2435 6.9100E-04 2310 8.6600E-04 E '

05-13-89 2436 7.0800E-04 2312 8.4500E-04

~05-14-89 2436 8.3000E-04 2327 8.2200E-04

• 05-'15-89 2436 6.6500E-04 2288 9.2900E-04

! 05-16-89 2435 6.7975E-04 2286 1.2304E-03 l'

05-17-89 2434 7.3550E-04 2305 1.0006E-03 05-18-89 2434 7.0519E-04 2300 9.6256E-04' 05-19-89 2435 8.8000E-04 2286 6.0100E-04 05-20-89 2433 6.5800E-04 2270 9.8700E-04  ;

05-21-89 2435 6.4900E-04 2279 1.0600E-03 .

05-22-89 2436 7.0700E-04 2275 1.1000E-03 05-23-89. 2436 5.8600E-04 2280 1.1500E-03 05-24-89 2436 6.3300E-04 2261 9.6900E-04.

05-25-89 2436 6.5500E-04 2259 1.0600E-03 05-26-89 2436 6.7900E-04 2254 8.0000E-04 05-27-89 2436 4.4200E-04 2256 9.0200E-04 05-28-89 2436 4.7500E-04 2241 9.7900E-04' 05-29-89 2436 6.5500E-04 2244- 1,0200E-03 l

05-30-89 2436 6.2500E-04 2227 8.2000E-04 L~ 05-31-89 2435 6.5600E-04 2230 7.8300E-04 06-01-89 2436 2.0100E-04 2233 8.4000E ,1 06-02-89 2436 6.9400E-04 2210 8.9100E l 06-03-89 2436 6.9700E-04 1350 5.0400E-04 L 06-04-89 2436 5.9200E-04 2353 8.0400E-04 4 06-05-89 2436 5.1100E-04 2188 1.1700E-03~

06-06-89 2436 6.3500E-04 2195- 9.51~00E-04 06-07-89 2436 7.3200E-04 2164 9.2200E-04 06-08-89 2436 6.0600E-04 2182- 8.9600E-04 06-09-89 2434 5.6100E-04 2164' 9.7100E-04 06-10-89. 2436 7.7530E-04 2171 1.4340E-03

' 06-11-89 2436 7.6981E-04 2167 1.2605E-03 06-12-89 2436 6.0800E-04 2161 1.2100E-03 06-13-89 2436 6.4100E-04 2154 9.2000E-04 06-14-89 2436 6.7400E-04 2150 '9.2600E-04 l 15-09 2436 6.2100E-04 2150 8.2800E-04 06-16-89 2432- 6.9360E-04 2140 1.0650E-03 l 06-17-89 2433 6.9000E-04 2076 9.9800E-04 L 06-18-89 2434 6.4600E-04 2135 6.7000E-04 06-19-89 2436 6.6800E-04 2129 9.4200E-04, 06-20-89 2436 5,5600E-04 2115 1.1300E-03 ^

06-21-89 2436 4.8200E-04 2111 7.5400E-04 06-22-89 2436 5.8000E-04 2100 1.1400E-03 06-23 2436 6.0300E-04 2098 9.4200E-04 06-24-89 2436 8.3700E-04 2090 1.0300E-03 06-25-89 2434 1.1100E-03 2090 8.6800E-04 l 06-26-89 2434 6.3000E-04 2083 9.9500E-04 06-27-89 2436 6.7400E-04 2080 1.1880E-03 06-28-89 2436 7.2300E-04 2063 1.0230E-03 06-29-89 2432 7.8800E-04 2072 1.0300E-03 06-30-89 2436 6.1400E-04 2052 1.0200E-03

~ , . - . . .- _ . - ~ -. -- - -..

s DRILY CHEM!$7RY DATR - 1989 DATE MWT(1) D0tt ,E0_!!31(1) MWTC2) DOSE.E0_!!31(2)

('- .-

07-01-89 2436 6.3600E-04 2057 9.2000E-04 07-02-99 2350 6.7800E-04 2056 1.0300E-03 2436 07-03-s9 6.6500E-04 2048 1.3400E-03 07-04-89 2436 7.7100E-04 2049 4.7100E-04 07-05-89 2436 '6.9000E-04 2035 7.2300E-04 07-06-89 2436 6.9900E-04 2032- 1.0200E-03 07-07-89 2436 6.4794E-04 2024 4.7814E-04 07-08-89 2436- 6.5000E-04 2018 9.3700E-04 -

07-09-89 2436 4.7300E-04 2018 7.4300E-04 07-10 2436 6.6500E-04 2010 7.4800E 07-11-89 2436 5.5900E-04 1996 7.7000E-04 07-12-89 2436 7.2600E-04 2002 8.8000E-04 i 07-13-89 2436 6.3100E-04 1988 9.8400E-04 07-14-89 2436 7.9000E-04' 1985 9.1400E-04 07-15-89 2436 6.7700E-04 1976 1-3400E-03 07-16-89 2436 4.7655E-04 1976 1.1225E 07-17-89 2436 6.2400E-04 1968 8.7500E-04 07-18-89 2434 6.1200E-04 19C2 8.3700E-04 07-19-89 2436 6.7500E-04 1956 8.1900E-04 07-20-89 2436 4.9100E-04 1936 1.0600E-03 4 07-21-89 .2436 6.5700E-04 1925 8.4100E-04  ;

07-22-89 2436 6.0800E-04 1945 8.2000E-04 07-23-89 2436 6.5000E-04 1957 8.2300E-04 07-24-89 2436 6.1600E-04 1935 8.9300E-04 07-25-89 2435 7.0100E-04 1915 8.1000E 07-26-89 2436 7.2470E-04 1918 8.7030E-04 07-27-89 2435 5.1350E-04 1917 8.9510E-04' 07-28 2436 7.4800E-04 1900 8.3900E-04 07-29-89 2436 6.0600E-04 1897' 9.4900E 30-89 2436 6.8100E-04 1902 8.2000E-04 31 2436 6.7300E-04 1871 8.7200E-04 +

08-01-89 2434 8.4051E-04 1889 1.1425E-03 L

04-02-89 2435 8.0980E-04 1878 6.1520E-04 .

l 08-03-89 2434 6.4100E-04 .1869 9.5900E-04 f h 08-04-89 2436 6.6800E-04 1875 8.7900E-04 L 08-05-89 2436 6.4800E-04 '1875 6.2500E-04 -

1 08-06-89 2436 6.8700E-04 1850 7.8000E-04 1 08-07-89 2436 6.9800E-04 1850 9.5500E-04 L"

08-08-89 2436 6.4100E-04 1839 8.5200E-04 08-09-89 2436 7.5000E-04 1838- 8.8500E-04 -- i 08-10-89 2436 6.0100E-04 1837 4.0800E-04 L

08-11-89 2436 4.7000E-04 1850 9.1100E-04 08-12-89 2436 4.7100E-04 1813 9.0900E-04 08-13-89 2436 5.8200E-04 1813 8.6100E-04 L 08-14-89 2436 5.0800E-04 1796 6.8400E-04 '

I 08-15-89 2435 6.6300E-04 1805. 8.7700E-04 L 08-16-89 2435 6.2200E-04 1801 1.0300E-03 L 08-17-89 2432 6.3400E-04 $795 8.2900E-04 08-18 2435 7.0500E-04 1800 8.8700E-04 08-19-89 2436 6.6600E 1800 8.6400E-04 08-20 2436 6.3400E-04 1800 7.8200E-04 L 08-21-89 2436 5.3800E-04 1771 8.6400E-04 1 08-22-89 2435 6.3400E-04 1759 8.5700E-04 1 08-23-89 2430 6.9200E-04 1757 6.1600E-04 08-24-89 2436 6.2900E-04 1739 8.1700E-04 '

08-25-89 2436 8.3900E-04 1750 8.0800E-04 08-26-89 2436 7.5600E-04 1750 1.0900E-03 08-27-89 2436 7.5000E-04 1750 7.7000E-04 08-28-89 2436 7.6100E-04 1750 1.0100E-03 08-29-89 2433 6.0400E-04 1720 9.0300E-04 08-30-89 2432 6.8100E-04 1425 5.02002-04 08-31-89 2433 4.4400E-04 1765 8.1100E-04 l

l i.

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m DRILY CHEMIE7RY'DR7R - 1989 DR7E MW7(1) ........,1131(1)

DOS E.,,E Q MW7(2) DOSE

................... ...... ........ ...... ... E0_!!31(2) 09-01-89 2436 -7.1300E-04 1700 8.7100E-04  :

09-02-89 1600 6.3000E-04 1696 9.8500E-04 '

09-03-89 920 3.5300E-04 1706 1.1230E-03 2436 09-04 6.9800E-04 0 1.1000E 4 09-05-89 2436 9.6S00E-04 0 4.4700E-05 09-06-89 2436 7.5700E-04 0 4.3400E-05 09-07-89 2436 9.0800E-04 0 2.4500E-05 09-08-89 2436 8.6100E-04 0 1'.6600E-05 1 09-89 2436 6.5400E-04 0 4.8700E-06 09-10-89 2432' 8.0400E-04 0 3.6100E-06 09-11-89 2431 6.7900E-04 0 2.2100E-06 09-12-89 2435 8.3700E-04 0 1.9500E-06 09-13-89 2435 8.0900E-04 0 1.8200E-06 09-14-89 2435 4.2800E-04 0 1.7300E-06 09-15-49 1900 7.1000E-04 0 8.3000E-07 09-16-89 845 3.2000E-04 0 8.1000E-07 '

09-17-89 840 3.6730E-04' 0 0.0000E+00 09-18-89 2434 9.2800E-04 0 0.0000E+00 09-19-89 2425 8.4900E-04 0 0.0000E+00 09-20-89 2436 7.6400E-04 0 0.0000E+00 09-21-89. 2436 8.2600E-04 -0 0.0000E+00 09-22-49 2436 6.4100E-04 0~ 0.0000E+00 >

09-k3-89 2436 6.5300E-04 0 0.0000E+00 09-24-89 2435 6.9600E-04 0 0.0000E+00 09-25-89 2436 4.0600E-04 0 0.0000E+00-09-26-89 2436 6.4000E-04 0 0.0000E+00 09-27 8436 8.0300E-04 0 0.0000E+0n 09-28-89 2436 6.1600E-04 0 0.0000E+00 09-29-89 2435 6.5400E-04 0 Oc0000E+00 09-30-89 2436 0.0000E+00 0 0.0000E+00

• 10-01-89 2436 0,0000E+00 0 0.0000E+00 10-02-89 2436 0.0000E+00 0 0.0000E+00' t 10-03-89 2436 5.7000E-04 0 0.0000E+00 10-04-89 2433' 5.1002E-04 0 .0.0000E+00 10-05-89 2434 6.6217E-04 0 0.0000E+00 2436 0.0000E+00 10-06-89 6.6800E 0 10-07-89 1663 6.2200E-04 0 0.0000E+00 10-08-89 2434 6.9900E-04 0 0,0000E+00 10-09-89 2432 7.4500E-04 0 0.0000E+00.

10-10-89 2436 5.5050E-04 0 0.0000E+00 L 10-11-89 2436 0.0000E+00 0 0.0000E+00-10-12-89 2434 7.5289E-04 0 0.0000E+00, 10-13-89 2433- 7.5300E-04 0 0.0000E+00 10-14-89 2420- 5.3200E-03 0 0.0000E+00-10-15-89 2436 5.9900E-03 0 0.0000E+00 10-16-89 2436 5.8500E-03 0 0.0000E+00.

10-17-89 2434 6.9800E-03 0 0.0000E+00 10-18-89 2436 1.6500E-02 0 0.0000E+00 10-19-89 2436 7.5000E-03 0 0.0000E+00 10-20-89 2436 6.4900E-03 0 0.0000E+00 10-21-89 2436 7.2900E-03 0- 0.0000E+00 10-22-89 2436 7.5800E 0 0.0000E+00 10-23-89 2436 8.2300E-03 0 0.0000E+00 10-24-89 2436 8.6900E-03 0 0.0000E+00 '

10-25-89 2436 9.1100E-03 0 0.0000E+00 -

10-26-89 2436 8.3900E-03 0 0.0000E+00 10-27-89 2436 8.1500E-03 0 0.0000E+00 10-28-89 2434 7.5500E-03 0 0.0000E+00 10-29-89 2430 7.1100E-03 0 0.0000E+00 10-30-89 2436 7.1800E-03 0 0.0000E+00 10-31-89 2436 6.6700E-03 0 0.0000E+00 i

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DRILY, CHEM!$7RY DR7R - 1999 DATE MWT(1) DOSE _EO,1131(1) MWT(t) DOSE

................... ...... ............... ...... ... E0_!!31(2) ...........

O ~

11-01-89 2436 7.7100E 0 0,0000E+00 i 11-03-49 2436 7.1800E-03 0- 0'.0000E+00 11-03-89 2435 7 2000E-03 0' O.0000E+00 11-04-89 2051 9.8500E-03 0 0.0000E+00 '

11-05-89 2435 6.7300E-03 0 .0,0000E+00 11-06-89 '2435 6.7300E-03 0 0.0000E+00 11-07-89 2436 5.8300E-03 0 0.0000E+00 11-08-89 2434 5.8200E-03 0 0~.0000E+00 11-09-89 2433 5.9200E-03 0 0.0000E+00 11-10-89 2436 6.4290E-03 0 0.0000E+00 11-11-89 2436 6:4150E-03 0 0.0000E+00 .)

- 11-12-89 2436 7.0400E-03 0 0.0000E+00 11-13-89 2436 1.4400E 0 0.0000E+00  :

11-14-89 2436 1. 2 0 0 0E-02 ' 0 0.0000E+00 6 11-15-89 2435- 5.3700E-03 0 0,0000E+00 11-16-89 2436 4.8900E-03 0 0.0000E+00 11-17-89 2436 4.9900E-03 0 0.0000E+00. t 11-18-89 2436 5.1820E-03 0 0.0000E+00 11-19-89 2436 4.7900E-03 0 0.0000E+00 11-20-89 8436 4.6870E-03 0 0,0000E+00 11-21-89 2436 5.3900E-03 0 0.0000E+00 ,

11-22-89 2436 4.5200E-03 0 0.0000E+00 ]

11-23-89 2436 4.2500E-03 0 0.0000E+00 11-24-89 2436 3.8360E-03 0 0.0000E+00-11-25-49 2436 4.4200E-03 0 040000E+00 11-86-89 2436 4.4600E-03 0 0.0000E+00 11-27-89 2435 4.0400E-03 0 0.0000E+00 1 11-28-89 2436 3.7512E-03 0 0.0000E+00 1 2430 11-89-89 4.6100E-03 0 0.0000E+00 11-30-89 12-01-89 2436 2436 3.7400E-03 4.0300E-03 0

0 0.0000E+00 0.0000E+00 ] i 12-02-89 '2436 4.1100E-03 0 0.0000E+00 1

-1 12-03-89 2420 3.7800E-03 0 0.0000E+00 i

12-04-89 2436 3.7500E-03 0 0.0000E+00 12-05-89 2436 3.6700E-03 0 0.0000E+00 12-06-89 2436 3.4500E-03 0 0.0000E+00 i 12-07 2436 3.6700E-03 4 0.0000E+00 12-08-89 2436 3.5300E-03 9 6.9600E-06 12-09-89 2436 3.9800E-03 122 6.7300E-06 12-10-89 2436 3.9800E-03 95 1.3500E-05 '3 12-11 2436 5.9500E-03 90 '1.0100E-05 t 12-12-89 2433 4.8200E-03 315 5.0500E-05 12-13-89 2435 3.6700E-03 515 1.0400E-04 '

12-14-89 2433 4.1450E-03 326 5.5860E-05 12-15-8F 2436 1. 2 TV 0 E'0 3 275 5.5000E-05 12-16-89 2435 4.6900E-03 430 5.0100E-05 .

12-17-89 2431 4.5500E-03 580 8.9200E-05 12-18-89 2432 4.1600E-03 170 2.1500E-05 i 12-19-89 2435 4.2100E-03 25 7.3500E-06 -

12-20-89 2436 3.9500E-03 '457 1.0300E-04  !

12-21-89 2436 3,9100E-03 560 8.6300E-05 18-22-89 2436 4.1400E-03 867 1.6400E-04 4 12-23-89 2434 4.2100E-03 1445 2.4500E-04 o'

<=

12-24-89 2433 4.1660E-03 2408 4.1300E-04 12-25-89 2429 4.0600E-03 2436 3.8000E-04 18-26-89 2436 4.9800E-03 2407 5.2200E-04 ,

12-27-89 2436 7.6100E-03 2424 4.5200E-04 12-28-89 2435 4.7500E-03 2436 5.2600E-04 12-29-89 2436 3.8500E-03 2433 4.3200E-04

. 12-30-89 2436 3.6500E-03 2420 3.9700E-04 l 12-31-89 2436 3.9700E-03 2432 3.E300E-04 l-L 1

- - . -- -