ML20235M444
| ML20235M444 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/31/1988 |
| From: | Hairston W GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 0590I, 590I, HL-317, NUDOCS 8902280257 | |
| Download: ML20235M444 (94) | |
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ENCLOSURE PLANT EDWIN 1.
HATCH - UNITS 1 AND 2 NRC DOCKETS 50-321 and 50-366 OPERATING LICENSES DPR-57 and NPF-5 ANNUAL OPERATING REPORT FOR 1988 l
TABLE OF CONTENTS PAGE Introduction 1
SRV Challenges for 1988 2
Design Changes and Tests or Experiments Plant Design Changes (Safety Related) 4 Test or Experiment Requests 74 Data Tabulations and Unique Reporting Requirements Occupational Personnel Radiation Exposure 76 Reactor Coolant Chemistry Summary 80 s
6 8902280257 881231 gDR ADOCK 05000321 PNU 1
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e-i 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 1988 were 755.6 Net Megawatts Electric for Unit 1.and 768.3 Net Megawatts Electric for Unit 2.
General Electric Company furnished the Bolling 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 from the Altamaha River.
The plant is a co-owned facility with ownership delegated as follows:
Georgia Power Company 50.1%
Oglethorpe Electric Membership 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 i
Commercial Operation 12/31/75 09/05/79
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Georgia Power Company has sole responsibility for overall planning, design, construction, operation, maintenance and disposal of the plant.
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I SRV Challenges for 1988 g
UNIT 1 i
'DATE VALVES ACTUATED 2/26/88 1821-F013A, C,
F, G,
H, L
4/19/88 1B21-F013A, C,
D, E ',
F, G, H,
J, K, L UNIT 2 p
No Challenges i
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DESIGN CHANGE REQUESTS and TEST or EXPERIMENT REQUESTS Pursuant to 10CFR50.59, the following is a brief description and summary of the safety evaluation for each change made to Safety Related systems and components, and each test performed during 1988.
The safety evaluation summaries address the three criteria used to determine if a proposed change or test involves an unreviewed safety question, i.e.:
1.
If the probabliity 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 evaluated 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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'e DESIGN CHANGE REQUESTS78-256 Provide more reliable timers to be used for load shedding U1, U2 and load sequencing'during a Loss of Coolant Accident or.a Loss,of Off Site Power emergency.
1.
The principles of the. load shedding and load sequencing during a. Loss of Coolant Accident or a Loss of Off Site j
Power emergency remains unchanged.
2 There are no new modes of failure introduced as a result
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of this design change, i
3.
None of the safety limits are affected by this change.80-053 Provide the Main Stack Radiation Monitoring System with U1, U2 he high range effluent radiation monitoring capability for j
Post' Accident conditions as specified in NUREG 0578 (2.1.8.B.1).
1.
This modification does not change the purpose or.the performance of the system.
2.
The system response after the modification is exactly the same as before the modification.
3.
The system setpoints remain unchanged.80-455 Replace the existing air solenoids, which require an oil Rev. t supply, with fall close air solenoids which do not require U2 oil supply in. order to make the Refueling Bridge Main Holst Air Brake fall safe.
1.
The replacement switches are preferable for this application and are of the same quality as the original equipment.
2.
There are no logic changes involved.
3.
There are no logic and/or setpoint changes involved therefore there is no effect on the margin of safety.
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81-008
-Provide additional' cooling capacity in the drywell.
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.The quality of the new equipment installed In'the drywell is basically the:same as the existing design.
The design-and Installation of the:new equipment meet the criteria of 10CFR50 GDC-57.
The new units do not. function following an' accident.
The equipment is seismically quallfled to lEEE 344, 1975 but does not meet IEEE 323, 1974 and is not l
safety related.;
2.
No new modes of failure are introduced because the units and the associated ductwork are' seismically mounted where they cannot fall on other equipment.
3.
The new coolers provide additional cooling capacity for the drywell thereby removing hot spots at lower elevations and maintaining the average temperature below the maximum design temperature.
The pipe work, ductwork, cooling coil unit and fan unit are seismically mounted and the pressure boundary of the cooling coll is built to ASME Code requirements.
The margin of safety as defined in the Technical Specifications has been evaluated and accounted for.81-058 Install, modify, or delete pipe supports per lEB 79-014 Rev. 2 design requirements.
Revision 2 incorporates all 79-014' l
U1 work for Unit'1.
f 1.
.This change reduces the actual stresses to meet commitments in the Final Safety Analysis Report.
2.
No new modes of failure have been introduced.
3.
This change increases the margin of safety.
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81-174 Provide transient monitoring. Instrumentation on the High U1, U2 Pressure Coolant injection and Reactor Core isolation Cooling Systems capable of recording system parameters and responses during system start-ups.
1.
This will not alter the logic or affect the operation of j
the High Pressure Coolant injection or Reactor Core
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Isolation Cooling Systems as described in the Final. Safety Analysis Report.
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2.
The new circuit design will not introduce any new modes of failure.
3.
This will increase the operability and availability of the J
High Pressure Coolant injection and Reactor Core Isolation Cooling Systems by minimizing and improv.ing trouble shooting and maintenance activities.82-252 Provide protecton to keep voltage transients such as those U1 caused by motor start-ups from tripping the undervoltage
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or overvoltage monitors in the Reactor Protection System panels.
1.
The safety design basis of the Reactor Protection System remains unchanged.
The addition of a line voltage regulator Will prevent voltage trar.clents, such as those caused by a motor start-up, from tripping the undervoltage or overvoltage monitors in the Reactor Protection System panels.
2.
The addition of the line voltage regulator will increase the reliability of the alternate power supply for the Reactor Protection System.
The regulator, which is non-Class 1E, will service the alternate power supply which is also non-Class 1E.
The regulator will be separated from the Class 1E Reactor Protection System panels by the Class 1E supply breakers and will not impact the operation of these breakers.
3.
The Reactor Protection System scram functions and trip settings are unaffected.
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83-059 Delete the High Pressure Coolant' Injection System's Oli U1 Tank High-Level Alarm by disconnecting the high level alarm switch.
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The High Pressure Coolant injection Turbine is provided l
with adequate lubrication with the oil tank level at-the l
low level alarm setpoint; therefore, elimination of the high level alarm'is Justified.
The turbine manufacturer has concurred with this change.
2.
Nulsance annunciations will be eliminated.
No new modes I
of failure are introduced.
3.
This change does not affect equipment addressed in the Technical Specifications.
i 83-089 Make the Reactor. Cavity Skimmer Surge Tank Gate Cover U2 rigid so that it will not warp when bolted in place and l
allow water to leak into the dry cavity.
1.
This modification replaces the existing skimmer surge tank gate cover with a more rigid cover plate assembly.
2.
No new modes of failure are created due to this modification.
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This modification provides a higher safety factor by l
adding a new leak tight cover plate assembly.
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83-264 install heaters in the Station Battery Rooms to maintain a U1 temperature of 77.F during normal operation, a hydrogen monitoring system to prevent development of explosive conditions and install two hour time delays on the battery room fans.
This will increase battery life and capacity to adequately handle the addition of ATTS loads and prevent rapid temperature fluctuations in the station battery rooms in the critical two hour period following a Loss of Off Site Power event.
1.
The addition of the unit heater and the hydrogen monitoring systems will not inhibit or interfere with the operation of any safety system or component within the battery rooms or elsewhere in the plant and have no bearing on the probability of an accident as the station batterles are used only In a Loss of Off Site Power event.
The delay of the start of the emergency exhaust fans until two hours following such an event does not increase the probability of an accident or malfunction as this modification prohibits a cooling drawdown of the battery rooms and prevents the electrolyte in the batterles from cooling beyond the temperature which would prevent their maximum discharge capability.
2.
This modification will not degrade any equipment or prevent any system from functioning as stated in the Final Safety Analysis Report.
3.
The basis for the applicable Technical Specification requires station battery operability.
This modification will enhance battery operability; especially in cold weather.84-011 Install a new fire detection system in the Unit 1 Power Rev. 2 Block to comply with the fire codes and 10CFR50, Appendix U1 "R".
1.
This modification is being implemented to comply with the response to Appendix "R"
to upgrade plant safety.
2.
The structures have been analyzed to ensure that the margin of safety will not be decreased by the additional loads.
3.
System operation and design as defined in the Technical Specifications is not impacted.
The increase of loads on the structure due to the system is negligible.
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842012
' Install' a new fire detection system In the Control Rev'. 2L Building'to comply.wlth the' fire codes and 10CFR50, U1 Appendix "R".
1.
This modification is being Implemented to comply with the response to Appendix "R"
to upgrade plant safety.
2.
The structures have been analyzed-to ensure that the margin of safety will not be decreased by the additional loads.
3.
System operation and design-as defined-In the Technical' Specifications is.not impacted.84-014
. Remove the existing fire detection after the new Rev. 1 Pyrotronics system is Installed and functional.
U1, U2 1.
The.new system will satisfy surveillance requirements.of the Fire Hazard Analysis in a manner meeting or exceeding the old system.
2.
The new-Pyrotronics system performs the same function as the old system and cannot result in any different types of q
accidents or malfunctions than previously evaluated.
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The new system will meet the requirements of the Fire Hazard Analysis, which is referenced by the Technical
' Specifications.
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84-017 Provide fire detection for the Diesel Generator Building
'U1, U2 to comply with 10CFR50, Appendix'"R".
.1.
The modification is being implemented to comply with the response to Appendix "R"
to upgrade plant safety..
1 2.
Though the Fire Protection System is not safety related, I
the supports were seismically designed.
3.
System operation and design as defined in the Technical
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Specifications is not impacted.
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84-018.
Install a new fire detection-system in' areas outside the.
Rev.1 Power' Block-to comply with the fire codes and 10CFR50, U1, U2 -
Appendlx "R".
1.
The modification-is being implemented to comply with the response to Appendix "R"
to upgrade plant safety.
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Though the Fire Protection System is not safety.related, V
the supports were seismically designed.-
3.
System operation and design as defined.in:the Technical I
Specifications is not impacted.
L.84-019 Install the. Multiplex System master panel and printer in U1,.U2 the Control Room and install slave panels and printers in the CO2= tank area on the 147' elevation of the Control 0
Building'.
Interconnect the. master and slave printers and connect the fire detection systems.
1.
The modification Is-being implemented to comply with the response to Appendix'"R" to upgrade plant safety.
2.
The structures have been analyzed to ensure the margin of safety will not be decreased by the. additional loads.
'3..
System operation and design as defined in the Technical Specifications is not impacted.84-020 install a new-fire detection system in the Unit 2 Power Rev.
3-Block to comply with the fire codes and 10CFR50, Appendix U2 "R".
1.
The modification is being implemented to comply w'ith the response to Appendix'"R" to upgrade plant. safety.
'2.
The supports are seismically designed.
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System operation and design as defined in the Technical Specifications is not impacted.
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i!84-142 Add manual keylock switches to the Loss of' Coolant Rev..1, Accident signal that trips the drywell cooling units 1and U2 return' air fans ;t.o allow for operation of specified units.
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- In a: post accident situation.
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The operation of the Primary' Containment Cool'ing System i
~does not actively interact with any equipment or systems important to safety'except for the Diesel Generators.
A Diesel; Generator loading calculation was perforged to show that-the Emergency Diesel Generators have suffholent capacity to support the operation of the drywell cooling.
3 units.and return air fans in addition.to the " norma'l"
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emergency loads.
It was determined that overriding the trip signal for this equipment during a Loss of Coolant-Accident followed by a Loss of Off Site Power event would not cause the Emergency Diesel Generators to be overloaded.
2.
The new override control relays are Class 1E, seismically l
quallfled, and divisionally separate.
The addition of the control relays does not affect the seismic qualification of the control' panels.
.lf the keylock switches or the-control relays should fall "off" during accident conditions-(1.e. a Loss of Coolant Accident with a Loss of Off Site Power event) no new accident is created since the drywell cooling system does not perform a' safety related function.
If the keylock switches or the control relays should fall "on" during accident conditions no new accident is created as the loading calculations show the diesel generator has sufficient capacity to support the operation of the.drywell cooling equipment and the
" normal" emergency loads.
In the event of.a Loss of Coolant Accident concurrent with a Loss of Off Site Power event, the Primary Containment Cooling System may be manually loaded onto the diesel generator using the override switches.
However no new accidents or malfunctions of the diesel generator'is i
created as'the manual keylock switches in combination with I
administrative controls ensure the is no potential for inadvertent operation of the drywell cooling equipment in the override mode.
In the event of a Loss of Coolant Accident followed by a Loss of Off Site Power event all operating equipment will stop and automatically be reloaded onto the diesel generator.
No new accident is created as the loading calculations show the diesel generator has sufficient capacity to support the operation of the drywell cooling equipment and the " normal" emergency loads.
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In the event of a Loss of Coolant Accident concurrent with a Loss of Off_ Site Power event, the components associated with drywell cooling will administratively not be loaded onto the diesel generator until sufficient electrical capacity of the diesel generator has been verifled by the operators.
In the event of a Loss of Coolant Accident followed by a Loss of Off Site Power event the diesel generator loading calculation shows that sufficient capacity exists to support the operation of the drywell I
cooling equipment without overloading the diesel generator.
Therefore the margin of safety as defined in the Technical Specifications is not reduced.84-226 Install a new Refueling Bridge Mast on the Unit 1 U1 Refueling Bridge.
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1Til s mod i f icat ion wi l l not increase the probability of occurrence or the consequences of a refueling accident (bundle drop).
2.
This modification will not create the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report.
3.
This modification will not reduce the margin of' safety as defined in the basis for any technical specification.84-249 Replace the existing Traversing in-core Prob @ System purge U1 line check valve with a safety grade valve to satisfy the containment isolation requirements of NUREG-0737.
1.
The replacement valve design requirements exceed those of the existing valve; therefore, the system's ability to meet the original design conditions will not be altered.
2.
This modification will not affect the operation of this system or the associated purge system.
3.
This modification will not affect the operation of this system as described in the Technical Specifications; therefore, the margin of safety is not reduced.
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U2 line check valve with a safety grade valve to satisfy the containment isolation requirements of NUREG-0737.
1 The replacement valve design requirements exceed those of the existing valve; therefore, the system's ability to meet the original design conditions will not be decreased.
2.
This modification will not affect the operation of this system or the associated purge system.
3.
This modification will not affect the operation of this system as described in the Technical Specifications; therefore, the margin of safety is not reduced.84-297 increase the tangent dimension of the outer clamp from U1 three (3) Inches to approximately four and one quarter (4 and 1/4) inches for the Residual Heat Removal System Service Water Pumps.
1.
The modified pump supports for the Residual Heat Removal System Service Water Pumps will ensure the pumps maintain their Selsmic Category 1 integrity.
2.
No unanalyzed failures are being introduced.
3.
The Selsmic Category I integrity of the pumps is maintained and the pumps remain operable.85-029 Replace the Refueling Bridge Load Cell System with a U2' system which is more reliable.
1.
This modification does not change the operation of the Refueling Bridge Grapple.
It only adds a more reliable load cell system.
2.
No new modes of failure are introduced.
3.
The Technical Specifications remain unaffected by this chsnge.
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85-049 Replace the existing Leeds & Northrup Temperature Recorder U2 on the Control Rod Drive System with a new recorder as the L
. 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 will not create an unsafe plant condition.
2.
No new modes of failure are introduced by this modification.
3.
No margin of safety is specified in the bases for the measurement of the hydraulic drive water temperatures. The recorder is utilized as a means to monitor record and annunciate high hydraulic drive water temperatures only.85-10S Replace the present Fission Product Monitoring System with U2 the General Electric NUMAC System.
1.
The new NUMAC Radiation Monitor is functionally equivalent to the existing monitor.and will perform all process radiation monitoring requirements.
2.
The NUMAC is an upgraded functional replacement and will not change any operational parameters.
3.
The replacement NUMAC monitors will provide the same protection as the existing monitors and will not reduce the margin of safety as defined in the Technical Specifications.
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1 r-85-112 Replace the bad cable conductor.'for the Reactor Water U1
. Clean Up System inboard Isolation valve position Indication with the existing spare cable which is located In the same motor control-center and runs.to the same control' panel'.
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This modification replaces a bad conductor in the existing j
cable with a conductor in a spare cable.
The' spare cable is of the same type and size asrthe existing cable.
2.
There is no change in any system function.
The cable only provides for valve indication-l 3.
There is no change in any system function therefore the margin of safety as defined in the Technical Specifications is not reduced.
i 85-122-Provide hard piping from the High Pressure Coolant l
U1 Injection' System's Pump Discharge Vent Pipe to a Reactor Building floor drain with a sight glass for flow 1
observation.
1.
This modification is designed and fabricated in accordance with the requirements of ASME Section 3, Class 2 and ANSI B31.1.
The:line will be seismically' analyzed to ensure the integrity of.the High Pressure Coolant injection System.
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2.
-The operation and. safety funct'lons of the system remain unchanged.
3.
This modification does not affect any requirements for the High Pressure Coolant Injection System as defined in the Technical Specifications.85-158 Modify the Reactor Water Clean Up System's Plpe Whip
- j U2 Restraint Number 147 in order to reduce the loads.
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The enhancement of this pipe whip restraint by adding shim
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plates does not change the design intent of th? restraint.
2.
No new modes of failure or accidents result from this modification as the design intent and configuration of the restraint are not altered.
3.
The margin of safety as defined in the Technical Specifications is not reduced.
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l 85-168' Replace obsolete and unreliable Offgas Preheater U1, U2 Temperature Controllers.
1.
This Instrument regulates power to the Offgas Preheater and assures start-up when required as stated in-the Final Safety Analysis Report.
'I 2.
The instrument provides no new system of operation not
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analyzed in the Final Safety Analysis Report.
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No Technical Specification limits are reduced,85-178 Spare out the old, permanent, unused integrated Leak Rate Rev. 1 Test cables to provide space for new Integrated Leak Rate i
U1 Test cables.
1 1.
This modification is an upgrade to equipment already existing in the plant and will actually improve the l
Integrated Leak Rate Testing.
2.
This modification is in support of a test described in the-Final Safety Analysis Report and required by the Technical Specifications.
1 3.
This modification is in support of a test specifically
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required by the Technical Specifications.
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86-029 Replace the Refueling Bridge Load Cell System with a U1 system which is more reliable.
1 1.
This modification does not change the operation of the Refueling Bridge Grapple, it only adds a more reliable load cell system.
2.
No new modes of failure are introduced.
3.
The' Technical Specifications remain unaffected by this f
change.
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188-097.
Install' Instrumentation on the Main Steam Isolation 1 Valves U1-
.for'the purpose of monitoring: stem vibration.
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The strain' gauges that are attached to the exposed part of the valve-stem are located such that theyodo not interfere with normal closure of the Main Steam isolation Valves.
Test data demonstrated the welds used for-attachment have Y
an insignificant affect'of'the integrity of the valve stem.
2.
No new modes of operation are added.
3'
'The Technical Specifications are not affected.
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l 86-0991 Replace the'HFB 70A thermal / magnetic breakers supplying U2
.the primary side of the motor control center's: 600-120/208 volt transformers with fused disconnect switches using 100 amp type FRS-R-fuses for the specified motor. control centers.
Replace the HFB 30A thermal / magnetic breakers of the primary _ side of the motor control center's 600-120/208
-voit transformers wlth 40' amp thermal / magnetic breakers for the specified motor control centers.
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. Replacing the 30, 70, and 100 ampere circuit breakers with l
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fused disconnect. switches will improve selectivity and l
coordination with downstream loads for faults of the type postulated in an Appendix "R"
type fire.
Therefore, system performance.during an event of this typeLwould be Llmproved.
The implementation of this modification will not adversely. affect the seismic qualification of the
- subject motor control centers.
.2.
System operation and performance are not affected by<the replacing of circuit breakers'with fused disconnect switches.
3.
.The-replacement of the breakers with fused disconnect switches will not change the safety related function or classification of the subject motor control centers.
Therefore, safety limits are not affected by this change.
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186-100 ReplaceJexisting 100 ampere thermal / magnetic molded case U2 circuit breakers'with a 100 ampere molded case non-automatic breaker, 1.
Replacing the automatic breaker with a non-automatic breaker'and fuses will increase reliability and eliminate the possibility of an over trip caused by two breakers in series tripping'on a fault.
The fault will-be cleared by-fuses so the upstream breakers will remain closed.
Therefore, system performance will have a greater margin of safety.
2.
System operation and performance are not altered as a
-result of this modification.
3.
' Safety margins will be improved because of this modification.86-101 Replace the 125 voit AC and 125 voit DC non-current U2-limiting control, fuses in the specified control room
. panels with' current.llmiting fuses.
1.
-Replacing non-current limiting fuses with current Illmiting fuses will imorove reliability.
2.
System operation and performance are not affected as a result of this change.
3.
Replacing the non-current limiting' fuses with current ilm cing fuses will improve safety margins.
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86-112 Provide replacement controllers for the Hydrogen U2 Recombiner System as the current controllers have become obsolete and replacement parts are no longer available.
1.
The replacement controllers selected are identical in function as the existing controllers and are one-for-one replacements.
They utilize the same process inputs and outputs for. controlling the Hydrogen Recombiner System operation as the existing controllers do.
The response times and sensitivities of the replacement controllers are either the same as, or are better than the existing controllers.
Therefore, flow and heater controls are not affected by this modification.
2.
The new controllers are one-for-one replacements for the existing controllers.
They have the same/similar design intent and system interfaces / Interactions as the existing controllers.
Therefore, no new modes of failure are introduced.by this modification.
3.
The Hydrogen Recombiner System operation or hydrogen removal capability is not affected by this modification; l
nor is the operation or function of any other systems important to safety because the new controllers are one-for-one replacements having the same operating characteristics and qualifications.86-204 Replace the obsolete Residual Heat Removal System Service Rev. 1 Water Cross-tle Pressure Switches with currently U2-manufactured pressure switches.
1.
The replacement pressure switch will not degrade or affect the ability of the system to perform its intended function.
The replacement switch is environmentally and seismically quallfled for this applicat.:on.
2.
The proposed modification will not degrade any equipment or prevent any system from functioning as stated in the Final Safety Analysis Report.
3.
The instrument setpoints will remain unchanged; therefore, system operation is unaffected.
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l 86-210 Upgrade the Flssion Product Monitoring System by replacing U1 the three existing log-count rate meters with one new NUMAC log-count rate meter.
1.
The NUMAC :og-count rate meter is designed by General Electric as a replacement for the existing log-count rate meters.
However, operation of the replacement log-count rate meters is different due to the microprocessor based design which enhances the system operation.
Interfaces of the replacement log-count rate meter with the related equipment will be the same as the existing log-count rate 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 panel.
2.
Replacing the existing equipment with the GE designed replacement does not introduce any new failure mechanism or modes.
The Fission Product Radiation Monitoring System will continue to function as before.
The system design bases remain unchanged.
3.
Based on the applicable Technical Specification, when the NUMAC log-count rate meter is taken out of service the reactor must be in the Hot Shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in the Cold Shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
However, due to the self-diagnostic capability of the new equipment maintenance or repair on the unit can be completed within a three hour time frame.
Replacement with a spare log-count rate meter should require even less time.
These enhanced design features of the new equipment should preclude having to proceed with a plant shutdown as required by the applicable Technical Specifications.
After the maintenance, repair or replacement is performed the operability requirements of the applicable Technical Specifications for leakage detection systems will be met.
Therefore, the margin of safety as defined in the Technical Specifications is not reduced.
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n 86-219 Install insulation on the Reactor Pressure Vessel Water Rev.-1 Level Instrumentation Reference Legs to protect the water 1
l U2 in the reference legs from flashing to steam thereby giving erroneous water level indications when high drywell I
E temperatures exist.
1.
The insulation will only be installed on the instrument l
piping from the temperature equalization column to the drywell penetration.
The operation of the water level column will be unaffected.
'2.
The NUKON insulation System has been evaluated and accepted by the regulatory agencies for its resulting properties following a design basis accident.
3.
The addition of insulation on the water level reference legs will not prevent the actuation of Emergency Core Cooling Systems at the various predetermined water levels.86-223 Add spray shleids over Class 1E electrical equipment other U2 than that required for Appendix "R"
In order to protect the equipment from inadvertent actuations of the existing sprinkler systems in the Reactor, Turbine, Radwaste and Control Buildings.
1.
This modification will protect certain Class 1E electrical equipment (other than 10CFR50 Appendix "R"
equipment) from inadvertent actuation of existing sprinkler systems and will not affect.the probability of occurrence or the consequence of any accident.
The water protective seal applied to the equipment increases equipment reliability in the event of inadvertent sprinkler system actuation and does not have any effect on the ability of the equipment to perform its intended function or to operate during or after a design basis event.
The water sealant and gasket material added to the electrical equipment will not affect cooling or ventilation requirements as the design j
ventilation paths are not blocked by the sealant.
The j
additional weight of the sealing material is insignificant
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and will not affect the seismic qualification of the 4
equipment.
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i 2.
Thia modification will :.ot degrade any equipment or I
prevent any system from functioning as stated in the Final l
Safety Analysis Report.
3.
No margin of safety as specified in the Technical Specifications is affected by this modification.
21
86-235 Install the Alternate Rod Insertion System in order to U2 provide compilance with the Anticipated Transient without a Scram rule 10CFR50.62.
1.
The Alternate Rod Insertion Systen is Independent from any equipment important to safety such'that operation of, or any failures of the Alternate Rod insertion System or equipment will not adversely affect or prevent any equipment important to safety from performing its safety function.
2.
The Alternate Rod insertion System is not designated as safety related and is designed as a backup to the normal scram function.
Any Interfaces with safety related systems utilize the appropriate isolation devices.
3.
The Alternate Rod insertion System is not classified as a safety related system.
The operation of, or any failures of the Alternate Rod insertion System or equipment will not affect the operation of any equipment important to safety as defined in the Technical Specifications.86-237 Replace the pilot disc material of specified Main Steam U1 Safety Rellef Valves from Stellite 6B to PH 13-8 molybdenum steel.
1.
This modification is an improvement to the design of the Safety Rellef Valves and will not degrade the Intended function of the affected valves.
2.
No new modes of failure will be Introduced by this design change.
1 3.
This modification will reduce the setpoint drift of the Safety Relief Valves and will improve their reliability.
l 22
86-243 Install' remote test switches inside the' level switch boxes U2 forethe heater circuit in order to prevent the lifting of the heater ~ wire at theLlevel elements when testing the Scram Discharge Volume Thermal Level Sensors.
1.
A failure of the components in this modification will not l
affectLthe system in any way such that the plant-will be in a condition outside those evaluated in the Final _ Safety Analysis Report as-a failure of one of these test switches
(
would only cause a half scram condition.
2.
This modification does not add any materials which have not been used previously at Plant Hatch.
.lf the components in this modification fall the result would be a half scram condition which has been previously evaluated in the Final Safety Analysis Report.
3.
This modification is being made to increase the durability-of the system and to enhance the performance of the plant surveillance procedure.
This modification meets the margin of safety as defined in the Technical Specifications.86-254 Modify cable trays in the cable spreading room to avoid U1, U2 the need for an-In-tray linear _ thermal detection wire per lthe criterla of Fire Hazard Analysis, Appendix "F".
1.
This modification will meet the Spreading Room fire detection recommendations as outlined in the Fire Hazard' Analysis, Appendix "F".
2.
No new' accident or malfunction types will be introduced as a result of this modification.
3.
This change is part of.an overall effort to upgrade plant fire detection.
1' l
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l 86-260 Provide design for the support of sample piping at the U1 Flssion Product Monitor Panels.
1.
The existing supports are being added to a system which is not safety related.
This modification will not affect the operation of the system or its original design function.
3 2.
The supports have been evaluated for seismic "two over one" criteria to assure that the supports cannot fall and damage any safety related components.
Therefore, no new failure modes are introduced.
)
3.
The addition of the supports will not affect the design intentions of the Fission Product Monitoring System or any margin of safety.86-284 Replace the boron solution in the Standby Liquid Control j
U2 System with enriched boron to meet the Anticipated Transient without a Scram criteria specified in 10CFR50.62.
1.
This modification will not alter the-operation of the Standby Liquid Control System or its conformance to the Bolling Water Reactor Emergency Procedure Guidelines.
The operational response of the Standby Liquid Control System will be Increased by this modification.
2.
The enriched sodium pentaborate will undergo isotopic testing to ensure the boron-10 content before acceptance for plant use.
Also, enriched sodlum pentaborate powder will be bought instead of enriched boric acid and corax to diminish the possibility of mixing errors which might occur when preparing the sodium pentaborate solution.
3.
Both the Final Safety Analysis Report and the Technical Specifications will be revised as a result of this modification.
The margin of safety as defined in the current Technical Specifications will be maintained.
Also, the new temperature and level alarm setpoints will be within the margin of safety as set by the current Technical Specifications.
24
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86-319 Reroute the specified cebles to satisfy'the electrical U2 separation criteria in the Final Safety Analysis Report as it relates to the voltage level in the secondary penetrations.
1.
This modification is being made to relabel certain secondary conte.inment penetrations to meet the voltage i
level requirements as stated in the Final Safety Analysis Report.
No system function is affected.
2.
This modification.does not affect the logic of any system or change any wiring terminations.
Therefore, no system function is affected.
3.
Only the secondary containment nomenclature will be changed.
This modification will not impact the ability of i
existing plant equipment to function in a design basis event.86-341 Replace the existing Reactor Pressure Vessel Shroud Head U2 Bolt Assemblies with new bolt assemblies to mitigate the effects of Intergranular Stress Corrosion Cracking.
1.
This modification results in a component upgrade incorporating such new design features as the elimination of crevices, improved water circulation and material upgrade.
2.
The function of the replaced components has not changed.
The new design features represents component improvements and the failure modes are the same as previously analyzed.
3.
The replaced components will not reduce the Integrity of the reactor vessel'to maintain its safety function.
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1 86-430 Replace the current Containment Radiation Monitors with m
U1 the General Electric NUMAC Radiation Monitors in order to l
provide on scale. radiation detection and eliminate voltage spikes that are inherent to the current radiation monitoring system.
e 1.
The.NUMAC' logarithmic radiation monitors are designed by
~
General Electric as a replacement for the existing logarithmic radiation monitors.
However, operation of the replacement logarithmic radiation monitors is different due to the microprocessor based design which enhances the system operation.
Interfaces of the replacement L
logarithmic radiation monitors with the related equipment l
will be the same as the existing logarithmic radiation monitors.
The NUMAC equipment is seismically quallfled and analysis has shown the Integrity of the panel in which-i the.new monitors will be installed is not adversely 1
affected.
2.
' Replacing the existing equipment with the General Electric designed replacement does not introduce any new failure mechanism or modes.
The Primary Containment Radiation Monitors will continue to function as before.
The system design bases remain unchanged.
Since the range is unchanged and the equipment is seismically quallfled, operation during post accident conditions is unaffected by this modification.
3.
These monitors meet all current Technical Specification requirements and are safety related and seismically qualified.
i 28
L 86-432 Replace the existing Standby Liquid Control System Pump's L
U1.
local pushbutton test switches with maintain contact l'
switches.
.1.
The control functions of the Standby Liquid Control. Pump T6st-Switches are not modified.
The replacement of these-switches with maintain contact switches will enhance-the surveillance' testing and emergency operating-procedures, k
2.
The functional-operation of the test switches has not been
. changed.
Also, having a maintainicontact switch enhances the pump's availability during an emergency condition requiring operation.of the Standby Liquid Control System Pumps.
3.
The operation or availability of the pumps are not modified by this change.
The replacement switch wlli-enhance' surveillance testing and emergency operation of the system.87-007 Replace the High Pressure Coolant injection System's U1-Turbine Steam inlet and the Exhaust Drain Pot Level Switches with a thermal activated level switch.
1.
The existing level switch is to be replaced by a current
' design level switch level switch with.the same operational capability.
New condults being added to effect this change will be supported by new and existing supports.
A design review of the new and existing supports has been canpleted and shown them to be adequate.
The supports were design as Selsmic Class I structures and rcet all the requirements of the Final Safety Analysis Report.
2.
The new switch will perform the same function as the existing switch which is to provide annunciation and open the bypass valve.
Loss of power to the electronics will drive the switch contacts to the alarm position.
3.
The operation of these level switches is not addressed in the Technical Specifications and is not critical to the proper operation of the High Pressure Coolant injection System.
I 27 m
e,87-008 Replace the High Pressure Coolant injection System's U2 Turbine Steam inlet and the Exhaust Drain Pot Level Switches with a thermal activated level switch.
1.
The existing level switch is to be replaced by a current design level switch level switch with the same operational capability.
New condults being added to effect this change will be supported by new and existing supports.
A design review of the new and existing supports has been completed and shown them to be adequate.
The supports were design as Seismic Class I structures and meet all the requirements of the Final Safety Analysis Report.
L 2.
The'new switch will perform the same function as the h
existing switch which is to provide annunciation and open the bypass valve.
Loss of power to the electronics will drive the switch contacts to the alarm position.
3.
The operation of these level switches is not addressed in the Technical Specifications and is not critical to the proper operation of the High Pressure Coolant injection System.87-009 Replace the Reactor Core Isolation Cooling System's U1 Turbine Steam inlet Exhaust Drain Pot Level Switch with a thermal activated level switch.
1.
The existing level switch is to be replaced by a current design level switch level switch with the same operational capability.
2.
The new switch will perform the same function as the existing switch.
Loss of power to the electronics will drive the switch contacts to the alarm position.
3.
The operation of this level switch is not addressed in the Technical Specifications l
28
_----_____-..,,.--.a
____.___--,----------_m__.1
87-010 Replace the Reactor Core isolation Cooling System's U2 Turbine Steam Inlet Exhaust Drain Pot Level Switch with a thermal activated 'evel switch.
1.
The existing level switch is to be replaced by a current design level switch level switch with the same operational capability.
2.
The new switch will perform the same function as the existing switch.
Loss of power to the electronics will drive the switch contacts to the alarm position.
3.
The operation of this level switch is not addressed in the Technical Specifications87-025 Install Belleville Springs on packed valves in order to U2 reduce valve stem leakage, extend packing life, increase stroke endurance and reduce maintenance costs.
1.
.Through the use of Belleville Springs proper gland bolt loads can be maintained on all valves that are reworked.
This will result in reduced valve stem leakage, extended packing life and increased stroke endurance thereby making any reworked valve more reliable.
System function and operation will not be affected.
2.
The only consequence of a failure of the Belleville Springs would be valve leakage.
This is a situation which has been evaluated previously.
3.
The Technical Specifications are unaffected by this modification.
87-061-Provide an opening on the Diesel Generator's Generator U1 Bearing Housing to allow for checking of the Generator Bearing insulation as recommended by the vendor.
1.
The addition of an inspection plate will not degrade or adversely affect the ability of the Diesel Generators to perform their intended function.
2.
The modification will not degrade any equipment or prevent l
any system from functioning as stated in the Final Safety I
Analysis Report.
No new modes of failure are introduced.
3.
This modification does not affect the Technical Specifications.
I l
29
Provide an opening on the Diesel Generator's Generator U2 Bearing Housing to allow for checking of the Generator Bearing insulation as~ recommended by the vendor.
1.
The addition of an inspection plate will not degrade or adversely affect the ability of the Diesel Generators to perform their intended function.
2.
The modification will not degrade any equipment or prevent any system from functioning as stated in the Final Safety Analysis Report.
No new modes of failure are introduced.
3.
This modification does not affect the Technical Specifications.
i 87-077 Provide for safe operation of the main hoist on the U1 Reactor Building overhead crane by reinforcing'the bearing support pedestals as recommended by the vendor, i
1.
This modification is being implemented using approved i
plant procedures.
This design change will improve the operation and reliability of the overhead crane by providing stiffeners on the bearing support pedestals as well as the hoist bearings.
This alteration has no impact on the.the probability of occurrence and consequences of an accident or malfunction as the predicted bearing wear during normal use will be reduced.
The crane-is not required to be operable for any accident.
Additionally the trolley will be located over a safe heavy load path or laydown area during modification.
Crane operability will not be degraded and reliability will be increased due to reduced bearing wear.
Crane operation will remain unchanged.
The affected equipment will not be released for use until the final drawings and calculations have been approved and documented.
2.
No new modes of failure are being introduced as this design strengthens the bearing support pedestals of the bearing hoist trolley.
Crane operability remains l
unchanged while reliability is increased.
3.
No margin of safety is affected.
This modification does not affect the Technical Specifications as the changes I
will be made with the trolley over a safe heavy load path or laydown area and not over the Spent Fuel Pool.
4
)
30
~ 87i080 Restore the pressure retalning integrity of the Residual U2 Heat Removal System. Service Water piping adjacent;to the identlfled' support by welding a carbon steel plate over the crack Indications.
1.-
The repair'of Residual Heat RemovalfSystem Service Water.
pipe.willfrestore the piping to the design requirements of
~
the system.
2.
The design. requirements described'In Table 3.2-1 of the Final Safety Analysis Report will be maintained.
No new.
modes of failure are introduced.
3.
The repair will ensure the margin of safety as defined in the Technical Specifications.87-086 Allow the use of. ASEA-ATOM type CR-82 Control Rod Blade so U2 as to obtain.the best. control rod blade available for long term cost effectiveness.
1.
The blade geometry, weight.and worth are virtually the same as a General Electric blade.
Therefore, the scram speed and worth are the same and the rod' drop velocity will not be increased.
Because the geometry Is the same, there will be no change in rod Insertion capability, withdrawal capability, or the effectiveness of the blade cooling and bypass flow.
2.
The blade geometry, including velocity limiter and coupling, is similar'enough to a General Electric blade that the ASEA-ATOM blade will behave the same as the ones
' described in.the Final Safety Analysis Report.
The use of different materials will lead to no additional blade cracking or absorber material loss.
3.
The blades worth, the scram. speeds and the drop out velocities for the ASEA-ATOM and General Electric blades are essentially identical.
Therefore, all safety related aspects of blade performance that could influence the margin of safety as defined in the Technical Specifications are met by the ASEA-ATOM CR-82 Control Rod Blades.
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31
o 87-087
. install two keylock manual inhibit switones, Indicating U2 lights and annunciator on the_ Automatic Depressurization System control panel to complete the Automatic Depressurization. System modifications required by NUREG-0737.
1.
The Automatic Depressurization System's function is.not
. changed or.affected by this modification.
The two-Independent manual keylock switches will only be used to inhibit automatic depressurization during an Anticipated Transient without a Scram event.
The keylocks, alarm and Indicating lights' ensure that automatic depressurization wlll not be inhibited unless the operator deliberately elects to do so.
When deemed necessary, prevention of'the Automatic Depressurization System's initiation will. allow the-operator to take actions to maintain proper reactor water level and will enhance the effectiveness of-the-Standby. Liquid Control System in shutting down'the reactor.
2.
No.new modes of. failure are introduced by this modification.
The addition of the inhibit switches does not affect:the ability of the Automatic Depressurization System to automatically initiate with a single failure.
?The switches are independent.-Class 1E and Seismic Category 1.
If'a single inhibit switch is placed in the inhibit position a light at-the switch and an alarm common to both switches will alert the operator of this condition. _Both inhibit switches must be in the Inhibit position to operationally prevent the initiation of the Automatic Depressurization System.
3.
This modification does not affect the Automatic Depressurization System's logic initiation or input parameters _when the system is required, nor does it affect the operation of the Automatic Depressurization System
-once it has Initiated.
The inhibit switches enhance the effectiveness of the Standby Liquid Control System, when the operator elects to use it, in shutting down the reactor in an Anticipated Transient without a Scram event.
1 32
1 l
87-099 Upgrade the Transfer Canal Leak Detection System by
[.
U1 providing flow routes from both Unit 1 and Unit 2 Transfer L-Canal Seal Assemblles.
1.
The addition of two one half (1/2) inch holes in the one (1) inch thick plate will allow for leakage water to reach the leak detection drain.
The sheet metal channel added.
at the gap between the two units will_ hold water until it reaches the leak detection drain.
These modifications wllI improve the rellabllIty of the leak detection system.
2.
The addition of two one half (1/2) Inch holes in the one (1) inch thick plate will have no effect in the plate's function.
3.
This modification will increase the reliability of the leak detectlon system.
l 33
--____-_----_______-____-________-_________-_-__________--.__-_-_-_-_-_____-_-_--__--_-_-___-_______-__a
87-161' Remove the space heaters and.the associated wire from the Rev. 1 limit switch compartment on specified Residual Heat-U2 Removal.Systemt and Main Steam isolation Leakage Control System valves.
1.
The removal'of the motor operated valves limit switch compartments: space heaters will lower the compartments internal temperature.
High ambient temperature eliminates-the need for internal space heaters.
The implementation i
of'this modification will not adversely affect the environmental of seismic qualification of the subject motor operated valves.
2.
Removal of the space heaters will not effect the valves' l
ability to carry out their function in system operation and performance.
3.
Renoving.the valve limit switch compartment space heaters will lower Internal. temperature of the limit switch compartment and will help the valves maintain their design temperature in a high temperature environment.
Therefore,
.the margin of safety will be Increased when compared to
.present conditions.87-167 Replace ~the Reactor Water Clean Up Piping from the
'U1 Residual Heat Removal System connection to the Reactor Water Clean Up System Outboard Isolation Valve and the system's inlet Isolation valve with nuclear grade piping.
1.
The new equipment will perform the same function as the old. equipment in all modes of operation.
The sole effect of this change will be reliability improvement.
2.
No new modes of failure are introduced.
Any failure of the new equipment would cause identical results as a failure of the existing equipment.
3.
The fallure of the new equipment will neither prevent operation of any safety related system previously evaluated, nor will It create the possibility of an unanalyzed failure occurring.
1 1
i j
34
87-169 Change the Reactor Water' Clean Up Area's high temperature U2
' annunciation and isolation setpoints in. order to provide an adequate margin between the normal ambient temperature and high ambient temperature setpoints-so.that unnecessary isolation of the system may be prevented during the summer months when the ambient temperature approaches.120.F.
1.
The new setpoints for the Reactor Water Clean Up Area's high ambient temperature annunciation and Isolation are conservative with regards-to the plant' design basis.
2.
This modification will maintain the' function of the system's annunciation and isolation for the high ambient temperature and does not result in new modes of plant operation.
3.
.The new Isolation setpoint of 140.F provides a greater marginlof safety based on the new maximum allowable value of 150.F than was provided by an isolation setpoint of 120.F with a maximum allowable value of 124.F.87-170 Replace the ASME Section 3 Class 2 sampling and instrument U2 root valves on the Main Steam Lines in the. Condenser Bay associated with' Main Steam and Electro Hydraulic Control Systems with bellows seal valves to reduce steam leakage.
-1.
The change does.not alter the original design intent.
2.
This modification will not affect the operability of the Instrumentation and the new valves will allow-reduction of steam leaks in the Condenser Bay.
3.
This modification does not alter the original function of the component.
l 1
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l 35 L
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'87-174=
Provide an Interlock circuit to prevent opening the U1.
Residual Heat Removal System's Torus Spray Outboard or
-Containment Spray Outboard Valves when a Shutdown Cooling Suction Valve in the same loop is open.
This. Interlock should also prevent opening a Shutdown Cooling Suction Valve-when the Torus Spray Outboard or Containment Spray Outboard Valves are open.
This modification will prevent the possibility of draining the Reactor Pressure. Vessel when the system is aligned to the Shutdown Cooling Mode.
1.
This modification is in response to IE Notices 84-81 and
'86-74 which recommend providing a control switch interlock-to prevent opening the Residual Heat Removal System's Torus Spray Outboard and Containment Spray Outboard Valves when a Shutdown Cooling Suction Valve in the same loop is not in the closed position.
This will preclude draining the Reactor. Pressure Vessel when the system is aligned.to.
the Shutdown Cooling Mode and not increase the probability or consequence of an accident.
This modification does not-affect any mode of the Residual Heat Removal System so there.ls no impact.on any of the events described in the Final Safety Analysis Report.
2.
This modification does not alter the original design intent of the operational modes of the Residual Heat Removal System.
The interlock will limit the flow path to a single design mode and reduce the possibility of loss of coolant. inventory through a secondcry path while the reactor is shut down and during start-up.
3.
This modification does not affect the design function of any of'the modes of the Residual Heat Removal System.
The.
modification provides a hardware solution to valve misalignment problems as addressed in General Electric Service information Letter 388 and IE Notices 84-81 and 86-74.
This modification may increase the margin of safety by reducing the possibility of operator error.
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36
87-175 Provide an interlock circuit to prevent opening the U2 Residual Heat Removal System's Torus Spray Outboard or Containment; Spray Outboard Valves when a Shutdown Cooling Suction Valve in the same loop is open.
This interlock should also prevent opening a Shutdown Cooling Suction Valve when the Torus Spray Outboard or Containment Spray Outboard Valves are open.
This modification will prevent the possibility of draining the Reactor Pressure Vessel when the system is aligned to the Shutdown Cooling Mode.
1.
This modification is in response to IE Notices 84-81 and 86-74 which recommend providing a. control switch interlock to prevent opening the Residual Heat Removal System's Torus Spray Outboard and Containment Spray Outboard Valves when a Shutdown Cooling Suction Valve in the same loop is not in the closed position.
This will preclude draining i
the Reactor Pressure Vessel when the Residual Heat Removal System is aligned to the Shutdown Cooling Mode and not increase the probability or consequence of an accident.
This modification does not affect any mode of the Residual Heat Removal System so there is no impact on any of the events described in the Final Safety Analysis Report.
2.
This modification does not alter the original design Intent of the operational modes of the Residual Heat Removal System.
The interlock wil-1 limit the_ flow path to a single design mode and reduce the possibility of loss of coolant inventory through a secondary path while the reactor is shut down and during start-up.
3.
This modification does not affect the design function of any of the modes of the Residual Heat Removal System.
The l
modification provides a hardware solution to valve misalignment problems as addressed in General Electric Service Information Letter 388 and IE Notices 84-81 and 86-74.
This modification may Increase the urgin of safety by reducing the possibility of operator error.
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4.
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l 87-182 Modify the Reactor Core isolation Cooling System wiring in Rev. 1 the control room panel to comply with Appendix "R"
and U2 preclude the necessity of manual operator action in a postulated Appendix "R"
event.
l l
1.
This modification does not affect Reactor Core isolation j
Cooling System's operation during normal or accident-j conditions.
The present operating configuration of the j
system is such that a fire could potentially force the i
operator to operate the entire system manually.
The I
reliability of the Reactor Core isolation Cooling System will be greatly increased upon implementation of this modification as the main autostart logic will be available in the event of a fire in the subject areas.
This modification has no effect on the probability or consequences of an accident or malfunction as discussed in the Final Safety Analysis Report as a design basis fire is not defined.
The Fire Hazard Analysis discusses the potential fires in each area and this modification reduces the consequences of the postulated fire.
2.
No new modes of failure are introduced.
Therefore, the possibility of an accident or malfunction of a different type from any evaluated previously in the Final Safety Analysis Report will not be introduced.
3.
This modification will affect the system operation, setpoints or response times during normal operation.
During an Appendix "R"
event the system operability is improved by preventing a short to ground fault from completely de-energizing the logic.
No margin of safety is affected.
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1.
a l
87-183 Provide the required fire barriers which are acceptable U2 for protecting the specified essential shutdown circuits as per Appendix "R"
requirements.
1.
This modification Will provide the specified condults with a three (3) hour fire barrier system with a combination one (1) and three (3) hour fire barrier system for specified circuits.
No physical changes to the subject circuits will be performed by this modification; therefore, the affected components will continue.to perform the. intended functions.
The. fire barrier material added increases equipment reliability by protecting the subject circuits against the effects of a fire.
The fire barrier material has been seismically quallfled and the raceway supports analyzed and modified as necessary for the addition of the fire barrier material to ensure the seismic qualification of the supports are maintained.
2.
This modification provides for protection against a fire of the subject circuits without physically changing these circuits.
This in turn will not affect the operation of the affected components.
Because the operation or function of the protected circuits have not been modified no new modes of failure are introduced.
3.
The margin of safety as defined in the Technical Specifications is unaffected by this modification as the system configuration, operation or operation of the affected components are not changed.
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e 4
87-184-Provide separate fusing for specified battery charger U1.
cablessIn. order to' prevent a short't.o ground fault-for n
battery chargers to station batteries.
1.
This modification will not-affect battery charger j
availability or ammeter readings under normal operating conditions.
The non-essential ammeter circuits are.
separated from the Class 1E switchgear frames with fuses, thereby preventing adverse interactions with the safety related portion of.the battery charger circuits.
The new L
circuitry will be seismically quallfled and seismically installed so.that the-qualifications of-the existing
[~
equipment-will not be violated.
The addition of fuses in-L the ammeter circuits does not alter the existing system.
logic, system operation, setpoints or response times.
This modification'wlll improve the reliability of the subject' battery chargers in the event of a fire induced fault in'an, ammeter circuit because the separation-provided by the fuses will prevent the battery chargers and switchgear supply breaker from tripping.
2.
Fallure-of the new fuses would result in a loss of battery charger current Indication but would not interrupt power distribution to the load.
Therefore, this modification will not degrade the battery chargers-or prevent them from functioning as assumed in the Final Safety Analysis Report.
3.
The Technical Specifications bases require-the operability of the DC power distribution systems.
The implied margin of safety as defined in the Technical Specifications is to maintain _the operability of the DC power distribution system as stipulated in the Limiting Conditions of Operation.
Since a failure of the.non-essential ammeter circuit will not affect the safety related battery charger circuits this modification will not affect system logic, operation, setpoints or response times and operability of the.DC power distribution system will be maintained.
Therefore, the margin of safety as defined in the Technical Specifications is not reduced.
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i e
87-185 Provide separate fusing for specified battery charger U2 cables in order to prevent a short to ground fault for battery chargers to station batterles, i
1.
This modification will not affect battery charger availability or ammeter readings under normal operating conditions.
The non-essential ammeter circuits are separated from the Class 1E switchgear frames with fuses, i
thereby preventing adverse interactions with the' safety-related portion of the battery charger circuits.
The new circuitry will be seismically quallfled and seismically f
Installed so.that the qualifications of the existing equipment will not be violated.
The addition of fuses in the ammeter circuits does not alter the existing system logic, system operation, setpoints or response times.
i This modification will improve the reliability of the subject battery chargers in the event of a fire. induced fault in an ammeter circuit because the separation provided by the fuses will prevent the battery chargers and switchgear supply breaker from tripping.
2.
Failure of the new fuses would result in a loss of battery charger current indication but would not Interrupt power distribution to the load.
Therefore, this modification will not degrade the battery chargers or prevent them from functioning as assumed in the Final Safety Analysis Report.
3.
The Technical Specifications bases require the operability of the DC power distribution systems.
The impiled margin of safety as defined in the Technical Specifications is to maintain the operability of the DC power distribution system as stipulated in the Limiting Conditions of Operation.
Since a failure of the non-essential ammeter circuit will not affect the safety related battery charger circuits this modification will not affect system logic, operation, setpoints or response times and operability of the DC power distribution system will be maintained.
Therefore, the margin of safety as defined in the Technical Specifications is not reduced.
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41
s:
s87-194 Modify specified pressure transmitter circuits that could Rev. 1 experience an error of up 1.25% of their calibrated ranges U1 due to leakage current associated with the States terminal blocks.
1.
-This change will reduce.the possibility of current leakage.
Splicing the cable will'not change the design intent, logic or operation of the subject transmitters.
1The cable splicing Will not constitute an unreviewed safety question because the safety related function of the Instruments is not affected.
2.
The cable splicing does not alter the original design intent, logic or operation of the subject transmitters.
3.
.The cable splicing does not affect the design function of The operational modes of the subject transmitters.
This modification will eliminate the possibility of any phase-to-phase or phase-to-ground current leakage.
The margin of safety is not reduced.
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42
4' t-I l
87-195 Modify specified pressure transmitter circuits that could Rev. 1 experience an error of up 1.25% of their callbrated ranges U2 due to leakage current associated with the states terminal blocks.
1' This modification-changes only the type of field connection; the safety related function of the instruments remains unaffected.
The subject transmitters will provide the same design function and interfaces with'other systems as before.
The terminal blocks, which may contribute to
' leakage currents in harsh environments, will be replaced l:
with' nuclear quallfled splices.
This modification will reduce the possibility of leakage. currents thereby maintaining the accuracy of the instrument loops.
The system function is unchanged from the previous design and system performance will be improved.
2.
This modification changes only the type of field connection; the safety related instrument functions or their safety related interfaces with other Instruments I
remains unaffected.
Consequently, this modification will not introduce any new modes of failure but will improve the accuracy of the instrument loops by reducing potential leakage currents.
3.
The modified circuits will maintain the required accuracy of the-Instrument loops by reducing possible leakage currents in the circuits.
Since this modification will allow each system-to function as previously designed the margin of safety as defined in the Technical Specifications is not reduced,87-210 Remove the drywell fire detection system.
U2 1.
Implementation of this modification does not impact the operation of the fire detection system.
The equipment being removed from the drywell has been rendered obsolete by the Installation of a nitrogen system.
The removal will be in compliance with the Fire Hazard Analysis.
2.
The deletion of this equipment does not affect primary containment fire protection as defined in the Fire Hazard Analysis.
3.
The plant fire. protection and detection will remain in l
i compliance with the current Technical Specifications.
43
,AV
\\,88-002 Spilce the motor leads inside the limit switch compartment U1 of the Reactor Recirculation System Discharge Valves, 1.
This modification will eliminate the possibility of any l
current leakage.
Splicing the cable will not change the design intent, logic or operation of the valves.
There is no impact on the accident analysis identified in the Final l
Safety Analysis Report.
The cable splicing will not constitute an unreviewed safety question.
2.
Splicing the cable does not alter the original design intent, logic or operation of the involved valves.
3.
Splicing the cable does not affect the design function of the operational modes of the involved valves.
This modification will eliminate the possibility of any phase-to-phase or phase-to-ground current. leakage.
The margin of safety as defined in the Technical Specifications will not be reduced.88-007 Provide adequate separation for specified valve pairs in U1 order to assure compliance with Appendix "R"
requirements for high/ low pressure interface components.
1.
The design requirements for internal panel wiring separation will be met.
Separating the panel's internal
, wiring will not impact any safety evaluation in the Final Safety Analysis Report.
2.
The rerouting of the panel's internal wiring does not alter the original design intent of the operational modes for the valves involved.
3.
The rerouting of the panel's internal wiring does not affect the design function of the operational modes for the valves involved.
The separation of the panel's internal wiring will increase the present margin of a
safety.
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.o 88-015-Provide a replacement transmitter'for the High Pressure U1 Coolant injection System's Pump Discharge Pressure Transmitter.
The current transmitter cannot be repaired.
1.
The new pressure transmitter will provide the same function as.the existing pressure transmitter and has the same range.and accuracy as the existing transmitter.
i 2.
The replacement transmitter will provide.the same signal to the associated indicator as the existing transmitter..
3.
.The High Pressure Coolant injection System's Pump Discharge Pressure Transmitter is not addressed in the Technical Specifications and the replacement of this I
transmitter will have no effect,on any components addressed in-the Technical Specifications.88-017 Remove the fuses in the circuit supplying power.to the U1 specified 1 test valves in the Core Spray System to assure compilance with Appendix "R"
requirements for.high/ low interfaces.for valve pairs.
1.
There-is no change in the operation or function of the 1 Core Spray System valves'affected by this modification.
2.
=The control panel's internal wiring must be rerouted to gain the required separation.
There will be no change.in the control logic, operation or safety function of the involved valves.
3.
The valves affected will operate and function the same after this modification as the did before as this change is limited to rerouting the control wire and. cable within the control panel.
45
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88-020
. Provide a replacement chlorination system for the plant
'U1 raw water systems.
This includes the removal and/or installation of ancillary instrumentation.
1.
The removal of the chlorine eliminates the probability of a hazardous chemical release accident, The storage of liquid sodlum hypochlorite on site cannot. produce.
sufficient amounts of lethal chlorine gas.in any credible scenarlo to. invalidate the accident. analysis in theLFinal Safety Analysis Report.
Additionally, the consequences to personnel,outside of the Main Control. Room are eliminated by the removal of the chlorine cylinders.. The consequences of a malfunction of equipment, specifically-the chlorine detectors and the Main Control Room Environmental. Control System isolation 1 dampers, is reduced by the elimination of the reliance on this equipment to
. mitigate the accident.
2.-
This modification will eliminate the chlorine on site and thereby eliminate the need for reliance on automatic main control room isolation.
The liquid. sodium,hypochlorite storage tanks will be. located in the chlorination' building within'a safety; dike designed to contain the entire inventory with a 10% margin.
In case of a tank rupture, release of chlorine vapor or toxicLvapors is negligible; therefore, no danger to-the operators is present.
The potential for release of chlorine from chemical reactions of.the stored sodlum hypochlorite with acid is eliminated by the use of.noncompatible fittings in the tank fill
. connections to the acid storage tank and the sodium hypochlorite tanks.
Administrative controls will be-established to ensure chemical tank trucks are sampled prior to filling a tank to ensure the right chemical is being put into the tank.
3.
A margin of safety equal to or greater than that provided by the automatic isolation of the Main Control Room is provided by the removal of chlorine from site.
The bases in the Technical Specifications regarding the automatic i
isolation of the Main Control Room Environmental Control System will be superseded by the commitment that no J
gaseous chlorine will be stored on site.
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l 46
88 Provide a design which will al' low secondary containment to U1 be-maintained on the Refueling Floor when both rallroad airlock. doors are.open and the equipment hatch la closed by isolating the Refueling Floor from the Unit 1 Reactor Building.
1.
The basic' function of the Unit I secondary. Containment System is to maintain a.one quarter (1/4)-Inch negative pressure differential when compared to atmospheric L'/
conditions remains unchanged-in the modified L
configuration; this includes the common Unit 1 and Unit 2 refueling floor area, the Unit 1 and Unit 2 Standby Gas
-Treatment. Systems and the Main Stack.
Therefore, the refueling activities and the Unit 2 reactor operation are permitted to continue.
2.
_For this modification proper precautions.will be taken to exclude the Unit'1. reactor building. area below the Refueling Floor and.the Standby gas Treatment System':s suction to that area so that secondary containment i
integrity.in the common Unit 1 and Unit 2 area above the refueling floor can be maintained.
This condition is permissible sir.ce plant activities that may lead to a postulated release of radioactivity during a Unit 1 refueling outage will be confined tc the common Unit 1 and
. Unit 2 refueling-floor area.
Therefore, the service of the Standby Gas Treatment System is not required in the Unit.1 reactor building are below the Refueling Floor.
3.
Secondary containment integrity for the modified configuration will be maintained and the Standby Gas Treatment System will continue to maintain the capability to draw a one quarter (1/4) Inch negative pressure differential when compared to' atmospheric conditions.
Therefore. the radiological consequences to the environment following a postulated accident are not increased.
1 I
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l 88-027 Modify the Standby Gas Treatment Outlet to Stack flow U1 recorder to indicate a full scale value of O to 6250 Standard Cubic Feet per Minute.
1.
The function of the affected instruments will remain the same and the accuracy of the recorder will be improved.
2.
This modification will not degrade any equipment or prevent any system from functioning as stated in the Final Safety Analysis Report. No new modes of failure are introduced.
3.
This modification does not change any design parameter in the Technical Specifications.88-030 Provide the necessary design which will allow for an U1 isolation boundary for primary containment to be created when testing the Torus-to-Drywell Vacuum Breakers.
1.
Certain accident modes, as stated in the Final Safety Analysis Report, require primary containment isolation.
The capability of the Torus-to-Drywell Vacuum Breakers' solenoid valves to remain closed against primary containment pressure during accident conditions will be improved by implementation of this modification.
This change is internal to the valves and will not impact any other system or equipment important to safety..The seismic qualification of the valves will not be affected.
The system / component function remains unchanged.
2.
Replacing the springs in the subject solenold valves will improve the containment isolation capability of the plant.
The operability of the Torus-to-Drywell Vacuum Breakers will not be affected by this modification.
No new modes of failure will be introduced since the solenold valves and the test system for these vacuum breakers will function exactly as they did before.
l 3.
The Technical Specifications require that primary containment be maintained.
The Torus-to-Drywell Vacuum Breakers are considered primary containment isolation valves.
This modification supports this commitment by ensuring these valves can remain isolated ag'ainst the peak primary containment accident pressures anticipated should the inboard containment isolation boundary fall.
The seismic qualification of these valves will not be affected.
The vacuum breaker's operability requirements will not be affected by this modification.
The subject I
solenoid valves serve to facilitate remote testing of the Torus-to-Drywell Vacuum Breakers.
The vacuum breaker testing function will not be affected by this modification.
I 48
t. 031 Provide the necessary design which will allow for an U2 Isolation boundary for primary. containment to.be created when testing the Torus-to-Drywell Vacuum Breakers.
1.
Certain accident modes, as stated in the Final Safety Analysis Report, require primary containment isolation.
The capability of the Torus-to-Drywell Vacuum Breakers *
. solenoid valves to remain closed against primary containment pressure during accident conditions willibe Improved by Implementation of:this. modification.
This change is internal to the valves and will not impact any other system or equipment important to safety.
The seismic' qualification of the valves will not be affected.
The system / component function remains unchanged.
2.
Replacing the springs in the subject solenold valves will
- Improve the containment isolation capability of the plant.. The-operability of.the Torus-to-Drywell Vacuum Breakers will not,be affected by this modification.
No-new modes of failure will be introduced since the solenold valves and the test system for these vacuum breakers will function exactly as they did'before.
3.
The Technical 1 Specifications require that primary containment be maintained.
The Torus-to-Drywell Vacuum Breakers are considered primary containment isolation
. valves.
This modification supports this commitment by ensuring these valves can remain isolated'aga' inst-the peak primary containment accident pressures anticipated should the inboard containment isolation boundary fall.
The seismic qualification of these valves will not be affected.
The vacuum breaker's operability requirements will not be affected W this modification.
The subject solenoid valves serve a facilitate remote testing of the Torus-to-Drywell Vacuum Breakers.
The vacuum breaker testing function will not be affected by this modification.
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88-033 Add a spacer to the Reactor Water Clean Up System's Return U2 Check Valve's shaft to allow for proper seating of the l
disk per the manufacturer's original intent.
L 1.
This modification only adds a spacer to the valve shaft to 1
l ensure the disk seats properly.
The system function is unchanged and the system performance will be per the original design.
2.
The function of the check valve and system interfaces remains unchanged.
Consequently this modification does not introduce any new modes of failure and will ensure the seating function of the valve is consistent with the original design intent.
l 3.
This modification will ensure proper seating of the check valve and prevent back flow to the Reactor Water Cleanup System during operation of the High Pressure Coolant injection System.
As this modification maintains the original design intent and function of the system the margin of safety as defined in the Technical Specification is unaltered.88-038 Remove the second inner carbon ring of the High Pressure U2 Coolant injection System's Turbine Casing Gland Seal to prevent further loss of chrome plating from the turbine shaft until such time as the shaft can be repaired.
1.
There will be no affect on the operation or response of the High Pressure Coolant injection System due to this change.
No other safety systems are affected by this modification.
2.
No new modes of failure are introduced as there is no affect in the operation and response of the High Pressure Coolant Injection System.
3.
The margin of safety as defined in the Technical Specifications is unaffected.
50
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o 039 Replace:the loose anchor bolts on the High Pressure
- U2 Cootant injection Pump Discharge Piping Hanger with larger anchor. bolts.where necessary to restore the hanger to an
~
operable condition.
1.
This modification restores'the' system to the original design condition and meets or-exceeds the original design requirements.
2.
The High Pressure Coolant Injection System will be restored to its original design condition and there will be no change in the system operation or, response.
3.
The High Pressure Coolant injection System will be restored to'Its' original design condition and there w!Il be no change in the system operation or response as defined in the Technical Specifications.
! 040 Improve the cooling capability of the Steam Chase heating, l
U2.
Ventilation and Air Conditioning System by installing a fan.coli unit in the Steam Chase.
1.
This modification will affect only systems which are non-safety related and are not required to function in an accident.
The Engineered Safety Feature Coolers for the High-Pressure Coolant injection area, the Residual. Heat 1
Removal area, etc.1 vill not be affected, s
2.
Failure of this cooling unit wrcid produce <the same consequences to the chilled water system as any of the' other. unite; therefore, no new modes of failure are introduced, n
3.
The Intent of this modification is to achieve design ambient temperature in the Steam Chase.
This will increase the operating differential temperature between-ambient temperature and the trip setting definco in the Technical Specifications thereby maintaining the margin of safety.
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I 51 j
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,i 88-043 Modify the supports as necessary to seismically support IJ2 the air supply test lines to the suppression chamber for i
the Torus-to-Drywell Vacuum Breakers.
1.
This modification will ensure the ability of the subject piping system to withstand a design basis accident and.
maintain its structural integrity.
No other system or equipment important to safety will be affected by this change.
As the instrument lines are not regulred to be functional during periods of cold shutdown this modification will be performed then.
During normal operation the system / component function will remain unchanged.
This modification does not affect the stress analysis on the Torus as there is an anchor and flex hose between the valve and the Torus.
2.
Currently there le no documentation available which supports the seismic qualif.lcation of the piping associated with the Torus-to-Drywell Vacuum Breakers upstream.of the existing TAP anchor.
Therefore, the system's integrity must be assured.
This modification involves documenting the seismic qualification and subsequent seltmic upgrade of the twelve (12) Instrument test lines from the Install TAP anchor to the new anchor being Installed.
This includes the outbcard side of the Torus-to-Drywell Vacuum Breakers.
This modification will ensure the Integrity of the piping which forms an outboard containment isolation boundary during a design basis accident.
The operability of any associated components will not be affected by this modification.
3.
The Technical Specifications require that primary containment integrity be maintained.
The subject valves with the associated piping form the outboard primary containment isolation barrier.
This modification supports this commitment by ensuring that the piping and supports associated with the subject valves will maintain their structural Integrity during a design bacis accident.
Therefore, the margin of safety as defined in the Technical Specifications is not reduced, f
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- P8-046
' Change the ReactorLWater Clean Up Area's high temperature
'U1 annunciation and isolation setpoints in order to. provide an adequate margin between the normal ambient temperature and.hlgh ambient temperature setpoints.so that. unnecessary isolation of the' system.may be prevented.during the summer.
months.when the ambient temperature approaches 120.F.
1.
This. temperature instrumentation-la part of the area leak p
detection system as described in the Final Safety Analysis-Report.
Evaluations have been performed to' demonstrate the acceptability of the new allowable value of 150.F In regard'to the design.crlterla of isolation on a 25 gallon per minute break, to the potential additione.1 inventory-losses duefto raising the limit and to the environmental qualification requirements for the equipment in the areas.
The new setpoints for the area high ambient temperature annunciation and isolation are conservative with.regards to these evaluations and the plant design basis.
2.
This modification will maintain the function of the.
Reactor Water Clean Up System's annunciation and Isolation for the high ambient temperatures described in the Final Safety Analysis Report and does not result In new modes of plant operation.
3..
Theinew Reactor Water Clean Up System's isolation setpoint-of 140.F is based on the prevention of excessive. inventory loss-in the event of a p.ipe break resulting in a 25 gallon per minute or greater leakage-Into the area.
For a postulated pipe break resulting in leakage less than 25 gallons per minute an alarm function set at 130.F la provided in the control room.
The GE system design specification for'the leak detection system requires annunciation of a 5 gallon per minute-leak.
A leak of this magnitude will'be detected when the flow rate into the Reactor Building Floor Drain Sump exceeds its alarm setpoint.
This alarm is ba'sd on the elapsed time since the last sump pump start.
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88-055-Modify the dry'well chiller control logic'such that'the~
U2 chillers, if operated in a post Loss of Coolant Accident situationk will: automatically trip should'a' Loss of'Off E
Site Power event occur.
1.
The operability of the Drywell Chilled Water System is not
. safety related, its use'is not required after a' Loss of-Coolant Accident and no credit is taken for its operation in the-Final-Safety Analysis Report.
The operation of the Drywell Chilled Water System does not actively interact with~any other' equipment.or system important'to safety i
except1for.the Plant Service Water System and the Diesel Gener7 tors.
The Plant Service Water System has adequate capatity to accommodate-the chillers under Loss of Coolant Accident or Loss of Off Site Power conditions.
The Drywell Chilled Water System is independent of any system that forms a part of the reactor coolant pressure.
boundary.
However, availability of-the drywell chillers is desired when a high drywell pressure condition' exist.
.Should a Loss of Off-Site Power event occur after too operator manual action to bypass the Loss of-Coolant Accident Signal and restart the chillers, the chillers would automatically trip and remain off.
Thus, bypassing-the trip signal for the drywell chliiers during a' Loss of Coolant Accident followed by a Loss of Off Site Power event would not cause the chillers to load onto the Diesel Generators.
2.
The operation of the Drywell Chilled Water System does not actively Interact with any other equipment or system-Important to safety except for the Plant Service Water
- System and the Diesel. Generators.
The Plant Service Water System has adequate capacity to accommodate the chillers under Loss of Off Site Power conditions.
In the event of a Loss of Coolant Accident followed by a Loss of.Off Site Power event all-operating equipment will stop and emergency loads would automatically sequence onto the Diesel Generators except for the drywell chillers.
If one of the wires in the affected 4160V switchgear should fall the continuous rating of the associated diesel generator could be exceeded if the corresponding dryweil chiller should start.
The Loss of Coolant Accident Trip and Trip Bypass Signals for the chillers are electrically and physically separate and of opper 'ia electrical divisions.
This results in the potential
%S of only a single diesel generator, which is a situation that has already been evaluated.
if the single Loss of Coolant Accident Bypass Switch should fall, either open or closed, with offsite power available the drywell chiller may inadvertently start.
If the switch should fall with a Loss of Coolant Accident and a Loss of Off Site Power present a contact in the switchgear would prevent the drywell chillers from starting and potentially adding drywell chiller loads to the associated Diesel Generators.
54 8.
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3 '. -
This modl'fication prevents the operation of the drywell m
chillers during any combination of a Loss of Coolant Accident and Loss of Off Site Power event and ensures the diesel' generator operability is not affected by.the chillers.
Therefore,'the margin of safety as defined in the Technical Specifications is not reduced.88-059 Provide a replacement for the for the Outboard Main Steam U1 Drain Valve as the current valve is no longer manufactured as a nuclear certified valve.
1 Psplacement of the existing steam drain valve containment isolation valve with a new KSB gate valve of the same class, rating and function will not change the system I
operation.
L 2.
.The change reduces the possibility of seat leakage.
No new modes of failure are introduced.
3.
The margin of safety is not decreased because the function of the valve is not being changed and all design requirements are still met.88-074 Add seal-in circuits to the Source Range Monitor and U1 Intermediate Range Monitor controls to eliminate the
. necessity of operators having to continuously hold the control pushbuttons in the contact position sfter a scram in order to properly position them.
1.
Changing out a pushbutton from momentary operation to momentary with a holding coll and adding a new pushbutton will not affect the capability of the logic or operation of any. safety equipment to perform its safety related function.
2.
The implementation of a holding coil pushbutton and addition of a pushbutton does.not change any aspect of the system logic or operation that could initiate a new kind of accident or malfunction.
3.
There are no modifications which will adversely affect any existing safety system.
The margin of safety is not
- reduced, i
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- =--__________-__-___. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _
88-093 install a compressor system and instrumentation necessary U1 for the performance of an integrated leak rate test.
1.
This modification is temporary and will only be in place while the reactor is in cold shutdown and the Integrity of the secondary containment is maintained.
2.
This is a temporary modification being performed in support of a test described in-the Final Safety Analysis Report and required by the Technical Specifications.
3.
This modification is in support of a test required by the Technical Specifications and will be removed prior to the unit leaving the cold shutdown condition.
The high drywell pressure signals will be made functional prior to leaving cold shutdown and are not required by the Technical Specifications in this mode of operation.
The Integrity of secondary containment will be maintained.88-095 Change the setpoints for specified high drywell U1 temperature alarms to eliminate nuisance annunciators in the control room caused by plant temperatures above the alarm setpoints.
1.
This modification only affects the annunciation of drywell high temperature and does not prevent the drywell chillers from performing their intended function.
2.
This modification does not degrade sny equipment or prevent any system from performing the functions as stated in the Final Safety Ana'iysis Report.
3.
The alarm function of the recorders is not addressed in the Technical Specif! cations.
All temperatures required for accident mitigation will continue to be recorded; only the annunciation function will be affected.
56
88-101' Change the setpoints for specified high drywell U2 -
temperature alarms to eliminate nuisance annunciators in the control. room caused by plant temperatures above the alarm setpoints.
1.
This modification only affects the annunciation of drywell high temperature and does not prevent the system'from performing its intended function.
2.
This modification does not degrade any equipment or prevent any system from performing the functions as stated I
in the Final Safety Analysis ~ Report.
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3.
All temperatures required for accident mitigation will continue to be recorded; only the annunciation function I
wlll be affected.
i 88-112 Isolate the Reactor Building from the Railroad Airlock and Rev.
1' the Hot Machine Shop by removing the power to the U1 ventilation fans and installing blank flanges on the ventilation ducts.
1.
The fans and ductwork are not safety related nor are they required to function after an accident.
The equipment is provided for-personnel comfort'and area ventilation.
Removal of this equipment will enhance the ability of the existing Heating, Ventilation and Air Conditioning Systems and the Standby Gas Treatment System to maintain secondary containment.
2.
No new'fallure modes are introduced.
The fans will have their source of power removed and the ductwork will be isolated by installing blank flanges.
3.
This modification will enhance the secondary containment requirements defined in the Technical Specifications by educing air flow from the Reactor Building into the Railroad Airlock.
57
88-124 Add a time delay to the down scale alarm of the Area U1 Radiation Monitors and delete the seal-in feature of the down scale alarm.
1.
The basic function of the Area Radiation Monitors remains unchanged.
The five (5) second time delay that will be i
added to the down scale alarm will eliminate the nuisance j
alarms in the control room, l
2.
The upscale trip provided by the Refueling Floor Area Radiation Monitors which actuate the Main Control Room Environmental Control System is unaffected by this change.
3.
The down scale alarm function will remain intact, only the speed of the actuation is reduced.
The down scale alarm is an indication of equipment failure and has no safety function.
The upscale function which actuates the Main Control Room Environmental Control System remains unchanged.88-125 Add a time delay to the down scale alarm of the Area U2 Radiation Monitors and delete the seal-in feature of the down scale alarm.
1.
The basic function of the Area Radiation Monitors remains unchanged.
The five (5) second time delay that will be added to the down scale alarm will eliminate the nuisance alarms in the control ro ors.
2.
The upscale trip provided by the Refueling Floor Area Radiation Monitors which actuate the Main Control Room Environmental Control System is unaffected by this change.
3.
The down scale alarm function will remain intact, only the speed of the actuation is reduced.
The down scale alarm is an indication of equipment failure and has no safety function. 'The upscale function which actuates the Main Control Room Environmental Control System remains unchanged.
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88-131 Remove.the Manual Reactor Vent Valves as they are U1 redundant to the Air Operated Vent Valves and are causing unnecessary maintenance and surveillance problems.
1.
The' Manual Reactor Vent Valves are primary containment j
isolation valves but otherwise serve noJsafety function.
l Their isolation function will.be maintained by replacing the valves with blind couplings and spool pieces.
This
.will ensure the seismic integrity of the Reactor Vent Piping System.
j 2.
The removal of the Reactor Manual 1 Vent Valves does not change the operation of the system.
The safe \\Y function-of the valves is maintained.
3.
The safety function of these valves will be sustained by.
replacing the valves with blind couplings.and spool pieces.
This will eliminate a primary system leakage path.88-144 Ralse the setpoint of the Steam Leak Detection-Temperature U2 Switch in order to eliminate unnecessary annunciation as
- the current setpoint is too low for normal plant operating conditions.
1.
.The setpoint increase will still preserve the ability of the switch to detect and annunciate on a steam leak equivalent to five (5) gallons per minute.
The function of annunciating on a five (5) gallon per minute leak Is-1 consistent with the design basis as described by General Electric.
A calculation study has demonstrated that the new setpoint of 177.F for the non-safety related high l
temperature alarm with a normal ambient temperature on the l
order of 140.F will preserve the design basis of'the plant and eliminate an existing nuisance alarm in the control room.
2.
The setpoint change will not produce any new modes of failure.
The function of the subject switch is to detect and annunciate in the control room upon a steam leak l
equivalent to five (5) gallons per minute.
This non-safety alarm does not contribute to nor introduce any accident or malfunction of a different type from any previously evaluated in the Final Safety Analysis Report.
3.
No margin of safety is specified or Implied in the Technical Specifications for the high temperature indicating / alarm function of the subject switch to detect a steam leak equivalent to five (5) gallons per minute.
1 59 L.
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a-88-170.
Replace the existing. local Junction boxes for.the Drywell Rev. 1 Wide Range Monitors High Range Detector with assemblies U1 which do not require additional environmental sealing.
1.
The ability to calibrate the wide range monitor will not
'be affected and the monitor's connector cable life will be improved by Installing a new pull box assembly.
This will enhance the environmental qualification configuration.
2.
No new failure modes'are introduced.
3.-
This change will improve cable connector life and will.not affect calibration of the wide range monitor.
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88-177 Provide ~a method to isolate the non-safety related signal U1, U2 of the Refueling Floor Area Radiation Monitoring System from the safety related portion associated with the Main
. Control Room Environmental Control System.
1.
'The control logic of Main Control Room Environmental Control. System remains unchanged with-respect to statements made in the Final Safety Analysis Report.
This modification eliminates the existing condition where a ground fault of the Area Radiation Monitors' Inputs to the system could cause a loss of the automatic portion of the control room isolation and pressurization modes due to the loss of control power.
2.
The addition of fuse to isolate the control power of.the Main-Control Room Environmental Control System will increase:the reliability of the system.
This modification will prevent a malfunction of the Area Radiation Monitoring System from causing a loss of control power for the control room teolation and pressurization modes.
'3.
' Modifying the control circuitry inputs provided by the Area Radiation Monitors will isolate the possibility of a ground fault from causing a loss of control power to the Main Control Room Environmental Control System.
Therefore, safety margins are improved and will not be adversely affected by this change.
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- 88-182 Provide a three (3) hour fire barrier for the specified U1 raceway.In the Diesel Generator Building.
.1.
'The affected raceway was. evaluated and found to be L
acceptable regarding cable derating due to the use of fire barrier. material.
AlI structural supports have been seismically analyzed and found to be adequate for accepting the additional weight added by the fire barrier material without any modification.
2.
The method of system operation Is.not affecteo.
This modification involves protecting the specified raceway in order to meet Appendix."R" requirements and assure adequate protection of the two safe shutdown paths.
Therefore, no new modos of.fallure are introduced.
3.
' Safety limits are not affected by this modification and no changes to the Technical Specifications are required.
Installation of the fire barrier material will ensure operability of the plant equipment located in the Diesel Generator Building.88-183 Provide a one (1) hour fire barrier for the specified pull U1 -
box in the Control Building in order to meet Appendix "R"
Requirements.
1.
The affected pull box was evaluated and found to.be acceptable regarding cable derating due to the use of fire barrier material.
All structural supports have been seismically analyzed and modified as necessary to satisfy requirements for-the additional weight added by the fire barrier material.
2.
The method of system operation is not affected.
This modification involves protecting the pull. box in order to.
meet Appendix "R"
requirements and assure adequate protection of the two safe shutdown paths.
Therefore, no new modes of failure are introduced.
~
3.
Safety limits are not affected by this modification and no changes to the Technical Specifications are required.
i installation of the fire barrier material will ensure operability of the plant equipment located in the Control Building area on the 130' elevation.
61
88-184 Provide a fire barrier'for the specified raceway in order Rev. 1 to meet Appendix "R"
requirements.
U1 1.
The affected raceway-was evaluated and found to be acceptable regarding cable derating due to the use of fire barrier material.
All structural supports have been seismically analyzed and found to be adequate for accepting the additional welght added by the fire-barrier material without any modification.
2.
The method of system operation is not affected.
This modification involves protecting the specified raceway in order to meet Appendix "R"
requirements and assure adequate protection of the two. safe shutdown paths.
Therefore, no new modes of failure are introduced.
3.
Safety limits are not affected by this modification and no changes to the Technical Specifications are required.
Installation of the fire barrier material will ensure operability of the plant equipment located in the area.88-190 Replace the Residual Heat Removal System Pump Motor Surge U1 Ring Brackets as recommended by General Electric in order to improve the motor's reliability.
1.
The existing brackets will be replaced with the new improved General Electric design.
This change will not affect the operation of the Residual Heat Removal System or its original design function.
'l 2.
'The existing brackets have been redesigned by General Electric to eliminate the possible cracking defect as noted in Information Notice 87-30.
3.
The replacement of existing surge ring brackets with an improved design will not affect the system's design intent or any existing margin of safety as defined in the
)
Technical Specifications.
J l
L I
i 62
__________ ___ _ - - a
x
! 191 Replace the Core Spray System Pump Motor Surge Ring U1 Brackets as: recommended by General Electric in order,to improve'the motor's reliability.
1.
The existing brackets will be replaced with the new improved General Electric design.
.This' change wllI not affect the operation of the Core Spray System or its original-design function.
2.
The existing brackets have been redesigned by General Electric to eliminate the possible cracking defect as noted in Information Notice 87-30.
3.
The replacement of existing surge ring brackets with an improved design will not affect,the system's design intent or any existing margin of safety as defined'in the Technical Specifications.88-197 Provide separate indications for the Standby Gas Treatment U1 System isolation Dampers so the possibility of misleading
-damper indication due to shared indicators can be reduced.
1.
.The lights used were evaluated and found to be acceptable regarding~past performance and reliability.
The control room panels which houses the Standby Gas Treatment System indication lights have been seismically analyzed and no modifications are necessary to support the additional weight added by the Indicator lights and wiring Jumpers.
2.
The method of system. operation is enhanced.
This modification involves the addition of Indicator lights to improve.the reliability of the Standby Gas Treatment System damper Indication in the control room.
Operation l
of the system is unaffected by this modification.
No new l
modes of failure are introduce.
1 3.
The Standby Gas Treatment System's safety limits are not affected by this modification and no changes to the Technical Specifications is required.
Installation of the indicator lights will improve the position Indication of the dampers involved.
l i
I l
63 1
c.
o 88-211 Provide a one (1) hour fire barrier for the specified U1 raceways.
1.
The affected raceways have been evaluated and found to be acceptable regarding cable derating due to the use of the fire barrier material.
All structural supports have been seismically analyzed and will be modified as necessary to ensure their seismic qualification is maintained for the additional weight added by the fire barrier material.
2.
The method of system operation Is not affected.
This modiflection involves only the protection of the specified racewass in order to meet Appendix "R"
requirements and ensur6 Adiquate protection of the two safe shutdown paths.
Theret 7 0, no new modes of failure are introduced.
3.
Safety limits are not affected by this modification and no changes to the Technical Specifications are required.
Installation of the fire barrier material, structural support modifications and the addition of a structural support will ensure the operability of plant equipment located in the affected area.88-233 Remove and replace any equipment and structures necessary U1 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 may include but is not limited to the addition of eyes for rigging, 1
l 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.
Eculpment and structures that interfere with the required motor maintenance will be removed.
Restoration will be completed to duplicate, to the extent possible, the original design of the affected systems.
2.
Interfaces that must be removed will be functionally l
tested and nondestructively examined, as required, l
following restoration.
3.
The Technical Specifications are not affected.
64 l
.s88-234 Repair the1 valve seat ring and' improve the ring to valve U1 body ~ Integrity for the Residual Heat Removal System's Test.
Line' Torus isolation Valve.
1.
The changing of material for the seat ring is a vendor recommended replacement and has the same fit, form'and function as the original seat. ring.
The vendor recommendation that the new seat ring be seal welded into the' valve body will'be incorporated.
This should prevent the screwed-in' seat ring from backing out and resulting in i
further damage to the valve.
2.
The vendor recommended change to the seat. ring material and the-use.of~ seal welding does not change the performance or function of the valve.
3.
This modification does not affect the Technical Specifications.
The new seat ring material and seal weld does not change the performance or function of the valve.88-251.
Provide a design for the replacement of two (2). Inch and U1 smaller gate and globe valves manufactured by Velan.
These particular-valves are no longer manufactured by Velan~and replacements are necessary for the ones I
presently installed.
1.
The replacement of the affected valves with equivalent valves of the same specification (s) will not have any impact on the system design or operation.
2.
The replacement of the affected valves will not change the function of the affected systems.
No new modes of failure ra re introduced.
3.
The safety of the affected systems is maintained due to the use of replacements of the same specification (s).88-265 Repair the deformation Indications at the bottom of the U1 Spent Fuel Pool in an effort to stop leakage from the Spent Fuel Pool.
1.
There is no change to any plant equipment by this modification.
The intent is to return the liner to its' original design.
2.
This modification adds steel plate to stop leakage from the Spent Fuel Pool.
There is no change to the intent of the original design.
3.
There is no actual change to the Intent of the original design.
The added steel plate will not affect safety.
l I
I 65
)
o 88-267 Replace the Impellers of the Core Spray Pumps with new, U1 Improved impellers in order to reduce vibration.
1 1.
This modification does not significantly alter the operation of.the Core Spray System.
The system will still be able to meet the design requirements in case of an accident.
This modification will increase the pump reliability and therefore the system reliability.
The performance of the Core Spray System is enhanced and is in no way degraded by this modification.
2.
The logic and operation of the system are not significantly altered.
The performance and reliability of the pump and the system is enhanced.
No new modes of failure are introduced.
3.
No safety limit or setpoint in the Technical Specifications is altered.
Since this modification increases the reliability of the pump, the availability and ability of the Core Spray System to meet the requirements in the Technical Specifications is enhanced and not reduced.
66 j
j
n.88-289 Modify the: supports located.in the southwest corner of the
-U1 Spent Fuel Pool so as to allow access to repsir the liner plate and to reduce' interferences and colliolons between the Transfer Canal Gate and the Spent Fuel Pool Cooling.
Line Supports.
.1.
This' modification'does not involve any equipment Important to safety.
The Spent Fuel Pool Cooling System is.
non-safety'related and.the non-seismic design remains unchanged by this modification. -The Final Safety Analysis Report does not specifically address the unlikely. event of the southwest pipe run in the Spent Fuel Pool impacting the Spent Fuel Racks.
However, the pipe weight and geometry have always been enveloped by the Technical Specification limit ~of 1600 pounds.
2.
This modification does not introduce any new modes of failure.
.The non-safety, non-selsmic design of the southwest cooling leg of the system wil.1 not be altered.
The. probability of the holsted gate colliding with line is decreased as this modification creates a greater clearance for movement.
The seismic integrity of the Spent Fuel Pool Liner will not be adversely affected by this modification.
3.
No margin of safety is specified or implied in the
. Refueling Section of the Technical Specifications relative to' cooling of the Spent Fuel Pool.
The Technical Specification limit of 1600 pounds over the Spent Fuel Racks will not be violated either directly or indirectly due to the total collapse of the piping run.88-311
-Repair the cracks in the specified Reactor Pressure Vessel U1 Nozzle to Safe End Welds on the nozzles.
Repair any other l
. defects found in other nozzle to safe end welds found by' the expanded scope of in Service Inspection.
1.
A weld overlay on the Recirculation inlet Nozzle Safe End to the Recirculation inlet Nozzle will not have any impact on system design or operation.
2.
A weld overlay on the Recirculation inlet Nozzle Safe End to the Recirculation Inlet Nozzle will not change the function of the system.
No new modes of failure are introduced.
3.
The nozzle condition is restored to equal or better than the original design requirements by this repair.
Therefore, the ability of this safety related component to j
perform its function is not degraded and no safety or design limits are affected.
I 6'7
88-312 Provide supplemental interlocks on the Residual Heat U1 Removal System's Shutdown Cooling, Torus Spray and Containment Spray Valves to prevent the outboard Torus Spray Valves from opening when a Shutdown Cooling Valve in the same loop is not in the closed position.
1.
This design provides an instantaneous interlock in the affected valves open circuit in addition to the interlock contacts provided by Design Change Request 87-174.
This additional contact is required as a result of the approximate 120 second closure time of the Shutdown Cooling Valves.
The additional interlock will prevent opening the outboard Residual Heat Removal System's Torus-Spray and Containment Spray Valves when a control switch for a Shutdown Cooling Valve in the same loop is not in the closed position.
2.
This modification does not alter the original design intent of the operational modes of the Residual Heat Removal System.
The interlock will limit the flow path to a single design mode and reduce the possibility of loss of coolant inventory through a secondary path while the reactor is shut down and during start-up.
3.
This modification does not affect the design function of any of the modes of the Residual Heat Removal System.
This modification provides a hardware solution to valve misalignment problems as addressed in the General Electric Service Information Letter 388, IE Notice 84-81 and IE Notice 86-74.
This modification may increase the margin of safety by effectively reducing the possibility of operator error.
l 68
o 1
l 88-313 Provide supplemental Interlocks on the Residual Heat U2 Removal System's Shutdown Cooling, Torus Spray and Containment Spray Valves to prevent the outboard Torus Spray Valves from opening when a Shutdown Cooling Valve in the same loop is not in the closed position.
1.
This design provides an Instantaneous interlock in the affected valves open circuit in addition to the interlock contacts,provided by Design Change Request 87-174.
This additional contact is required as a result of the approximate 120 second closure time of the Shutdown i
Cooling Valves.
The additional Interlock will prevent opening the outboard Residual Heat Removal-System's Torus Spray and Containment Spray Valves when a control switch for a Shutdown Cooling Valve in the same loop is not in the closed position.
2.
This modification does not alter the original design intent of the operational modes of the Residual Heat Removal System.
The interlock will limit the flow path to a single design mode and reduce the possibility of loss of coolant inventory through a secondary path while the reactor is shut down and during start-up.
3.
This modification does not affect the design function of any of the modes of the Residual Heat Removal System.
This modification provides a hardware solution to valve misalignment problems as addressed in the General Electric Service Information Letter 388, lE Notice 84-81 and IE Notice 86-74.
This modification may increase the margin of safety by effectively reducing the possibility of operator error.88-314 Due to the degradation of the carbon steel, inbody seat U1 ring threads of the Residual Heat Removal System's Service Water Heat Exchanger Outlet Flow Control Valve an alternate method is required for installation of stainless steel seat rings.
1.
The change in the installation method of the seat rings is eQulvalent to the original installation and will not change or degrade the operation and function of the valve.
2.
The performance or function of the affected valve is not changed as a result of this modification.
I 3.
The change in installation of the seat rings does not affect the performance or function of the valve.
69
= _ _ _ _ _ - - - _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _. _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
r l
l 1
l 88-318 Seal weld around the seat ring of the Residual Heat U1 Removal System's Torus Spray Inboard Isolation Valve in l
order to prevent leakage from a small inbody crack through the area between the seat ring and valve body threads for the seat ring.
I 1.
The operability of the valve is completely restored and the function of the system is not impaired.
No credible failure mode of the existing crack will impact either the valve or system function or impact the pressure boundary.
l 2.
Seal welding the seat ring in place does not change the performance or function of the valve.
Leaving the existing crack will not introduce any new credible mode of failure to the system.
3.
This modification does not affect the Technical Specifications.
Seal welding the seat ring in place does not change the performance of the valve.
Leaving the crack in place will not affect the performance of the valve.
The ability of the system to perform its intended function is not reduced.88-320 Relocate the Olvision il back-up power supply for the U2 Automatic Depressurization System's logic from the "A"
to "B"
train in order to make the logic single failure proof with regard to the power supplies.
1.
Due to this modification if a failure of the back-up power from the "A"
Station Service Lattery for~the "A"
logic occurred the "B"
logic train would be available to automatically initiate the Automatic Depressurization System as it will now receive back-up power from the "B"
Station Service Battery.
If a failure of the back-up power from the "B"
Station Service Battery for the "B"
logic occurred the "A"
logic train would be available to automatically initiate the Automatic Depressurization System as it will still receive back-up power from the "A"
Station Service Battery.
This modification will require changes to the Final Safety Analysis Report but does meet divisional separation criteria and does not impact the Appendix "R"
Analysis.
2.
This modification will make the Automatic Depressurization System meet the single failure criteria with respect to the back-up power supplies.
This will improve the availability of the system for operation during an accident condition.
3.
The function of the system logic is not changed.
Only the source of the back-up power supply for the "B"
logic train is being relocated.
70
[..
s.
l 88-321 Provide mitigation for a weld on the Residual Heat Removal U1 and Reactor Water-Clean Up System's Interface which is l'
susceptible to Intergranular Stress Corrosion Cracking.
1 ~.
The. weld overlay of the affected Interface will not have an impact on the original system design or operation.
2.
System design and operatl0n will remain unchanged.
No new i
modes of failure will be introduced.
f 3.
The Interface weld will be restored to a condition which is equal to or better than the original design.
Therefore, the ability of this safety related component to perform its function is not degraded.88-322 Relocate the Division il back-up power supply for the U1 Automatic Depressurization System's logic from the "A"
to "B"
train in order to make the logic single failure proof with regard to the power supplies.
1.
Due to this modification if a failure of the back-up power from the "A"
Station Service Battery for the "A"
logic occurred the "B"
logic train would be available to automatically initiate the Automatic Depressurization System as it will now receive back-up power from the "B"
Station Service Battery, if a failure of the back-up power from the "B"
Station Service Battery for the "B"
logic occurred the "A"
logic train would be available to automatically initiate the Automatic Depressurization System as it will still receive back-up power from the "A"
Station Service Battery.
This modification will require changes to the Final Safety' Analysis Report but does meet divisional separation criteria and does not impact the Appendix "R"
Analysis.
2.
This modification will make the Automatic Depressurization System meet the single failure criteria with respect to the back-up power supplies.
This will improve the availability of the system for operation during an accident condition.
3.
The function of the system logic is not changed.
Only the source of the back-up power supply for the "B"
logic train is being relocated.
71
i 88-323' Replace the-failed Reactor Recirculation Seal Water U1
. System's Primary Containment isolation Check Valve I
manufactured by Velan with a Class 1 valve of more current manufacture as Velan no longer produces nuclear grade valves of this type.
1.
The design requirements of the replacement valve meet or exceed those of the existing valves.
The design of the j
system is not degraded by this modification.
2.
The pipe stress analysis and the replacement valve's selsmic qualifications have been reviewed and found to meet the original design requirements.
Therefore, no new modes of failure are introduced.
3.
The leakage rate requirements of the valve are in accordance with the requirements set forth in the Technical Specifications.
Therefore, the margin of safety are not reduced by this modification.88-324 Modify the Residual Heat Removal System's Shutdown Cooling U1 Outboard Isolation Valve so as to securely attach the valve yoke to the valve body and restore operability.
1.
The increase in the diameter of the bolts and associated bolt holes is minute and far removed from the pressure body.
The step bolt used is of the same material as the existing bolt and the manufacturer recommended torques will be used.
2.
There is no change to the bolt diameter or material for support of the yoke from the existing configuration.
The seismic qualification requirements for the valve are unaffected.
Therefore, no new modes of failure are introduced.
3.
There is no change to system or valve operation by this modification.
The system's Shutdown Cooling Mode of operation is unaffected by this modification as is the ability of the valve to perform its primary containment isolation function upon receiving an isolation signal.
72
c 88-339 Perform full structural overlay (s) on weld (s) made on the U1 Reactor Water Clean Up System Piping installed by Design Change Request 87-167.
1.
The weld overlay will not have any impact on the original system design or system operation.
2.
System design and operation remain unchanged.
No new modes of failure are introduced.
3.
The affected weld (s) will be restored to a condition which is equal to or better than the original design.
Therefore, the ability of this safety related component to perform its function is not degraded.
73
- a TEST or EXPERIMENT REQUESTS87-003 This test provides verification of rated flow through U1 minimum flow check valves on the Residual Heat Removal, Core Spray, and High Pressure Coolant injection Systems for the.in Service inspection commitment to satisfy ASME Section XI Code requirements.
1.
The affected equipment will be operated only in modes and within parameters for which it was designed.
2.
The affected components will not be subjected to any operating conditions more severe than they would see during normal system operation.
3.
This request will be implemented in conjunction with existing surveillance procedures.
No new failure mechanisms are introduced.88-001 This test shuts off power to the Emergency Response Data U1, U2 System Computer and the associated disk drives and records the data ^ history on tapes after the system is powered up again.
This action is being performed in order to investigate possible fallure modes of the system so that the software may be rewritten to preserve the data history after a power failure.
1.
The affected equipment has no safety function.
2.
No new modes of failure are introduced.
3.
The margin of safety is unaffected.
74
.i 6
i 88-002 This request install biological test chambers and U1 corrosion test racks in the Plant Service Water. System.
This will allow studies and evaluations to be performed on i
the rates of corrosion and material deposition within the j
system piping so the proper chemical treatment (s) may 1
determined for pipe preservation.
1.
This request allows for the collection of data necessary to evaluate the rate of corrosion and material deposition which is present'ly occurring within the Plant Service Water System piping.
This will be accomplished by attaching a set of side stream test chambers and racks to the actual system piping.
This test does not regulre a change to the Final Safety Analysis Report text or figures as the modification is of a temporary nature.
The only procedural change required is that which specif,les the system lineup so as to allow for proper operation of the test equipment.
This required chance is minor and has no affect on the system's operational parameters or characteristics.
2.
The installation of the test equipment does not modify the i
operation of the system for any accident analysis.
3.
The modes of operation of the system are maintained.
l 75
0 d
l OCCUPATIONAL PERSONNEL RADIATION EXPOSURE FOR 1988 This section has been compiled to satisfy the requirement of Plant E.
l.
Hatch Unit 1 and 11 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 methods 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 1,
1988 through December 31, 1988.
All monitored personnel were included in summary as provided under 10CFR20.407.(a)(2).
Individual exposures as Indicated bc self-reading pocket lon chambers were recorded daily with the use of an ALARA Computer 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 lon chamber results as recorded on the disc dosimetry flies 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 identifled 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 individual exposure records, is added for individual exposure records, and is then added to the total representing the cumulative dose in the appropriate job category.
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 prevall in all significant cases.
76 1
4 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 j
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 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 provide this information in the format specified by these requirements and the suggestions of the Regulatory Guides.
l 77 i
9, SUP91ARY OF PERSONfEL MONI70RIN3 ENDI!G DECEMBER 31, 1988 l
l GE0fGIA POWER COMPANY - NUCLEAR GENERATION PLANT E.I. HA7CH P.O.'BCX 439
'BAXLEY, CA 31513 DPR -57.
[
'NPF - 5 Estimated whole body exposure range Number of
( remst individuals in each rance No measurable exposu re...................................... 14 81 Measurable exposure less than 0.1............................ 676 0.1 to 0.25.................................................. 446 0. 25 t o 0. 5..................,............................... 415 0. 5 t o 0. 75.................................................. 2 72 0.75 to 1.0.................................................. 188
- 1. 0 t o 2. 0................................................... 4 14 2.0 to 3.0.................................................... 83 3. 0 t o 4. 0.................................................... 15 4.0 to 5.0....................................................
0 5.0to6.0....................................................0 6.0 to 7.0....................................................
0 7.0 to 8.0....................................................
0 8.0 to 9.0....................................................
0 9.0 to 10.0...................................................
0 10.0 to 11.0..................................................
0 11.0 to 12.0..................................................
0 12+..........................................................O Total number of personel monitorino 3990 This report is submitted in accordance with paracraph (aM2) of 20CFR20.407 Total Nanken for 1988 is 1400.55 78
- =
a, GOERGIA FOUER C';MPANY.- r4UCLEAR GENERATION PLANT E.I. HATCH P.O. BOX 439, BAXLEY, GA 31513 LICENSE: CPR-57 LICENSE: NPF-5 REGULATORY GUIDE 1.16 INFORMATION END OF YEAR REPORT 1988
- PERSONNEL (>100 NREM)
TOTAL MAN-REM
-WORK 1 JOB FUNCTION STATION UTIL u_________________---__----,__-___.._,_____
ITY CONTRCT STATION UTILITY CONTRCT
____-_. __________ ______~____--___
Reacter coerations & Surveillance
- Maintenance 4-Construction 96 3
201 61.634 1.493 77.301 coorations 81 0
0~
38.191
.016
.107 Health Physics & Lab 44 0
60 31.071 0.000 29.765 Supervisory & Office Staff 33 6
8 14.793 1.410 2.468
__' Engineering Staff 17 3
24 7.111
.944.
8.577
' Routine Plant Maintenance
, Maint enance !- Const uct ion 155 9
395 65.219 2.665 166.880 Operations..
52 0
1 23.425 0.000
,195 Health Physics-& Lab 41 1
41 31.355
.278 20.560
' Supervisory & Office Staff 15 6
11 4.915 1.682 4.488-Engineering Staff 13 4
29 5.805 1.692 9.953 s..
'Incorvice Inspection Maintenance 4 Construction' 30 0
174 15.517
.064 65.507-Operations 6
0 0
1.978 0.000
.008 Hesith Physics & Lab.
5 1
11 5.842
.139 3.276 Supervisory & Office-Staff 1 6
1 5
3.432
.2?2 2.852 6
1 8
3.257
.449 3.665 Engineering Staff LSpecial' Plant M a i nt-e nance Maintenance & Construction 102 8
498 50.732 4.430 332.024 Operations 33 0
0 10.272
.098.
.008.
'Hoolth Physics & Lab 37 1
60 23.545
,132 32.550 Supervisory & Office Staff 16 7
10 5.685 2.193 5.653 Engineering Staff 15 2
38 5.617
.810 16.798 Waste Processing Maintenance & Construction 23 0
106 8.113
.064 36.620 Occrations 5
0 0
1.093 0.000
.008 Hoolth Physics & Lab 4
0 16 3.029 0,000 4.974 Supervisory & Office Staff 2
1 4
.764
.292 1.472 Engineering Staff 0
0 1
.053
.017
.743
________ __________.---___-~~
2R+ Fueling Maintenance & Construction 38 0
177 14.757
.064 88.315 i
Operattens 17 0
0 5.198 0.000
.036-Health Physics & Lab 3
0 17 1.401 0.000 6,.687 Supervisory 6 Office Staff 2
1 4
1.039
.292 1.408 Engineering itaff 7
0 12 4.275
.096 4.478
- Totals Maintenance 3 Construction 444 20 1551 215.970 8.780 766.648 Operations 194 0
1 85.157 114 421
-Health Physics 8 Lab 134 3
203 96.243 550 97.811
' Supervisory 4 office Staff 74 2r 42 30.628 6.?61 18.342 Encineerine Staff SS 14 112 26.113 4.009 44.215
_..__,..____.______._____c__..
Grand Totals 9 14 55 1911 454.116 19.614 927.437 L
l l.
79
i 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 11 values are also shown although it is not required that they be reported.
Isotopic values which are listed as "O"
are less than the low level I
density limit of the counting system.
4 4
80
Unit I 1988 SJAE Isotocics uCi/SEC DATE 1988 iMWT lXe-133 IXe-135 1Xe-138 iKr-85m iKr-87 i Kr-88 I
E6
_____________l_______l________l_______l________l________l________l
__l Jan. 14 l2436 l1.84E1 17.20E1 13.56E2 l1.49El I6.30E1 14.62E1 15.71E2
_____________l.______l________l________l________l________l
___l________l_
Feb. 15 12436 12.07E1 19.13r.
!4.95E2 l 1.76El 18.08E1 15.81El 17.63E2
_____________1_______l________l________t._______l_______l________l________l.
Mar. 17 I2436 13.01El l8.75El l3.2552 l2.03E1 l5.48E1 l5.52E1 15.72E2
___________l_
_l________l_______l_______l________l________l________l______
Apr. 14 l2300 l5.5EE0 12.98E1 10.00E0 l8.76E0 ~13.61El l 2.61El l1.06E2
_________l_______l_
__l________l______ l________l
___l________l May 16 10 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0
_____________l_______l________l________l_______;________l
____l l_
Jun. 16 l2435 i1.05E2 l3.02E2 11.77E3 15.89El l2.48E2 11.78E2 l2.66E3
____________l_______l________l_ _ ____l________l________!________I________l_
Jul. 14 l2431 17.45E1 l2.66E2 l2.02E3 16.15El l2.74E2 l1.92E2 l2.89E3
_____________l_ _ ___l________l_
_l________l____-
_l________l________l_________
Auo. 15 12080 19.41El l2.07E2 12.14E3 15.00E1 l2.76E2 l1.69E2 l2.94E3
_ _________l
___l_ _ ____,_ _ ____l________l________!_
___l_____ _ l Seo. 15 12436 17.67El l3.27E2 l3.93E3 17.78E1 14.22E2 l2.92E2 15.13E3 l
l____
l_
_l.
_l_ _ _ _ l l_
___l oct. 13 10 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 l_
l________l_
___l_ _ ____l________l
___l________j i
Nov. 14 l0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 t
l l___
l_
l_
___l l________l________l Dec. 15 l2436 13.07E0 l5.20E1 l1.65E3
'1.57El l1.05E2 I5.34E1 11.88E3
_____ _ _l_
__l________l_
__I________
_l
_ _ l ___ __ __ l -
REAC'IOR CHEMISTRY IODINES UCi/ml DATE 1988 iMwr II-131 iI-132 iI-133 iI-134 iI-135 l DEI-131 l
l___
l
____l,_
l l__ _ ___I__ _ __
Jan. 14 l2436 i 6.04E-5 l 1.50E_3 15.11E-4 13.02E-3 11.04E-3 13.91E-4 l
l___
l_
_l_
l_
l _ _ _ _l________ _ _
Feb. 15 l2436 l0.00E0 19.97E-4 l3.87E-4 13.25E-3 l1.03E-3 l2.82E-4 l__
l___
l_
___l
___l
____l
__l___________
Mar. 17 l2436 l3.94E-5 l8.09E-4 12.75E-4 l1.86E-3 I5.63E-4 l2.22E-4 l
__ l _. _. ___ j _
___l__
l_____ __l
_l_______
Apr. 14 12300 12.45E-4 I6.88E-4 l4.81E-4 l2,54E-3 I8.685-4 l5.15E-4
(
________,____l_______!_______l____ _ _j.__.__l_______t_ _ _ _ l__________-___ _ _
14ay 16 10 l0.00E0 10.00E0 10.00E0 {0.00E0 10.00E0 10.00E0 l
q-__ _ __l_
l_
l.._
l______
=
Jun. 16 l2435 l9.21E-4 l1.08E-2 {3.76E-3 11.09E_2 l5.62E-3 11.61E_3
_____________l_____l____..l_
1_____ _ l________I____ _ l l-Jul. 14 12431 17.13E.4 19.73E_3 12.43E-3 17.69E-3 13.39E_3 12.17E-3
.__________l______l________l__
_l
___ l _ __ ___ l ___ __ _ l _ _ ___ -
Aug. 15 12000 11.18E-3 18.73E_3 i3.14E-3 17.76E-3 14.56E-3 12.86E-3
-l_______l________!__
g_______l________l________l_______ _ ____ _
I Sep. 15 12436 11.19E-3 l1.72E-2 16.* 0E-3 11.82E-2 19.67E-3 l 4.58E-3 1
__ _______ _ l__
l _ __ --
_l____.____l
____l________l_______l______-.
oct. 13 IO lo.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0
_____________l_______l_______.a________l________l__
__l__
_l___.1_________-___-
Nov. 14 10 l0.00E0 10.00E0 l0.00E0 10.00E0 l0.00E0 10.00E0 l
l_______l_
_l___ _ _l________l l_______
Dec. 15 12436 I7.15E-5 l9.07E-4 l7.42E-4 l6.48E-3 l2.16E-3 15.94E-4
_____________l_______l________l__
l________l________l_____ _ l___________--_---
81
l Unit II 1988 SJAE Isotocics uCi/SEC l
~
DATE 1988 lMWP lXe_133 1Xe-135 iXe-138 iKr-85m iKr-87 iKr-88 i
&6
_________p______p______p______l-
___p ______p_______;_
___l Jan. 15 10 l0.00E0 10.00E0 l0.00E0 10.00E0 10.00E0 l0.00E0 l0.00E0
.__ __ _ __ _ __ _ p __ _._ _ p _ _ _ ___ _ p
__ __ p______ p ______ p__ __ __ _ p ______ i Feb. 16 10 10.00E0 10.00E0 l0.00E0 10.00E0 10.00E0 10.00E0 l0.00E0
_ _____ __ _ p __ __ _ _ p _____ __ p ___ _ ___ p ____ _ _ _ p _____ __ p___ __ _ p __
._ p.
Mar. 18 IO I0.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 l0.00E0
_____________l-p_____ p_____ p
__l
_p p_______l Apr. 15 l2435 10.00E0 l1.33E2 l3.74E3 13.07El l2.77E2 11.19E2 l4.30E3
___ __ _ _____ p ______ p ___ __ __ p____ ___ p ___ _ _ __ p
___ p_______ p_____ p May 17 I 560 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 10.00E0 l0.00E0
_ ____ _ __ _ ___ p ___ _ p _____ __ p_ _ __. p ___ __ __ l --
___l
__ p_______ p Jun. 17 12436 l0.00E0 19.48E1 13.30E3 I3.24E1 13.18E2 l8.36El l3.03E3
_____________p______j-
- ___ p p _ ____j-
___ p_______ p_______ p________
Jul. 15 12431 l8.20E0 l1.06E2 l3.06E3 l2.28E1 l 2.23E2 I8.57El l3.51E3
_______p__
l-
__ p
__ p_______ p_______l-
_p l
Aua. 16 l2436 I8.16E0 l1.13E2 l 3,.17E3 l 1.96El !2.10E2 17.57El l3.60E3
____l.
_ l --
_p
__l-
____p______ p_______l
_l Sep. 16 12436 l7.20E1 l1.08E2 l2.73E3 12.03E1 11.85E2 I7.39El 13.12E3
______l
__ p______ p_ ___ p
__ __ p ___
l
_ p___
l oct. 14 12436 I7.87E0 l9.83E1 12.77E.3 11.98E1 l1.83E2 I7.29E1 l3.15E3
________l p
l
__ p _ ____[
___p_______l-
_l Nov. 15 l2436 16.83E0 l9.89El 12.76E3 l1.93E1 l1.89E2 17.45El l3.15E3
_______p_______l
____ p
__j
___l
___ p
__l Dec. 16 0
10.00E0 10.00E0 l0.00E0 l0.00E0 10.00E0 10.00E0 l0.00E0
__ _ __ _ _ p
__ p ____ __ l ---
__l
___l-
__ p
_i_______l REACTOR CHEMISTRY IODINES uCi/ml
~
DATE 1988 iMWP iI-131 II-132 1I-133 jI-134 lI-135 lDEI-131
_l
___ l ______ p__
p____ p_____ p____ p Jan. 15 10 10.00E-0 10.00E-0 10.00E-0 10.00E-0 10.00E-0 10.00E-0
_____________p______p__
_ p_______l
_ _ p_______p_______l______
Feb. 16 10 l0.00E-0 10.00E-0 l0.00E-0 l0.00E-0 10.00E-0 10.00E-0
_____________p_____p__
l
___ p __ _
p_______p_______ p________
Mar. 18 10 lo.00E-0 10.00E-0 10.00E-0 10.00E-0 10.00E-0 10.00E-0 p_ _____ p _ __ ___ p _____ p t
l
__ p ___ _ _ __
Apr. 15 l2435 l1.08E-4
' 4.39E-3 11.89E-3 l 1.10E-2 14.58E-3 l 1.35E-3
. __ _ _ _ __ _ ___ _ p___ __ _ p ____ __4 __ ____ __ p____ _ p _ _ __ __ p_ ___ ___ p _ __ _ ____
May 17 I 56C l 2.16E-5 14.55E-6 l 1.09E-4 I 4.60E-5 l 9.83S-5 l 2.79E-4 p____ p___ __ p
_p I _ __ ____ p_ ____ l ___ _
Jun. 17 l2435 l 1.63E-4 14. 42E-3 12. 20E-3 11.32C_2 15.55E-3 11.61E-3
____________p
_ p __ - --
p___ ___ p p_-
p___
p____
Jul. 15 12431 16.58E_5 l3.95E-3 11.66E-3 l1.25E-2 i4.22E-3 11.22E-3
_______..___.._ p______l_____ __l___
p _ __ _ ___ p _ ____.. p _ _ _ _ ___ l _ _ _.. _ __ _ __ __ _s Aug. 16 12436 l1.09E-4 14.2SE-3 !1.90E-3 l1.39F_2 l 4.38E-3 11.38E-3 p_
.- p __
p
- p____ _ __ p______ p.
p_____
Seo. 16 12436 l 6.91E-5 i 4.11E-3 l 8.45E-4 11.09E-2 14.35E-3 l 9.96E-4 p____ _ p _ _ __ _ p _ _ _.__ _ p __ _. p __
_p_______l-Oct. 14 12436 l1.40E-4 13.71E-3 l1.71E-3 l1.30E-2 l4.12E-3 l1.20E-3
_ _ _ _ _ _ _ ______ p _ __ _ _ _ p _ _ __ _ _ p _
p __ _ _ __ p _ _
p_______ p_________
Nov. 15 12436 11.3 8E-4 13.68E-3 l 1.69E_3 l 1.15E-2 14.15E-3 l 1.27E-3
____________l-p- _
p p____ _ p_______ p______l______________
Dec. 16 10 10.00E-0 l 0.00E-0 10.00E_0 10.00E-0 10.00E-0 10.00E-0
_ p_ _____ p - -
-p_______p_______p_______p_______l__________________
82
'4:
e DAILY CHEMISTRY DATR - 1988 i
.DATE MWT(1) DOSE _E0_I131(1) MWT(2) DOSE _EQ_I131(2)
January 1, 1988
'2436
.000336 1818
.0129 January 2, 1988 2380
.00034 1837
.0145 J anuarv 3, 1988 2436
.000325 1830
.0151 Januarv 4, 1989 2436
.000333 1830
.0152 January 5, 1988 2435
.000298 1825
.0161 January 6, 1988 2435
.000286 1831
.0161 January 7, 1988 2433
.000301 1820
.0578 January 8, 1988 2434
.000291 1825
.0158 Januarv 9, 1988 2432
.000273 1825
.0152 January 10, 1988 2436
.000302 1833
.0177 January 11, 1988 2436
.000308
-1828
.015 January 12, 1988 2436
.00032
.725
.0318 January 13, 1988 2436
.000323 0
.253 January 14, 1988 2436
.000306 0
.13 January 15, 1988 2231
.000322 0
.0498 January 16, 1988 2436
.00029027 0
.027864 January 17, 1988 2434
.000284 0
.0128 January 18, 1988 2434
.00028432 0
.012822 January 19, 1988 2435
.000305 0
.00624 Januarv 20, 1988 2436
.000348 0
.00809 January 21, 1998 2436
.0003 0
.0034 January 22, 1988 2435
.000278 0
.00198 January 23, 1988 2436
.000293 0
.000801 January 24, 1988 2435
.0003029 0
.0007417 January 25, 1988 2436
.000282 0
.00209 January 26, 1988 2436
.000293 0
.000798 January'27, 1988 2429
.000296 0
.000386 January. 28, 1988 2436
.000324 0
.000335 January 29, 1988 2436
.000377 0
.0004877 January 30, 1988 2436
.000286 0
.000879 January 31, 1988 2436
.0003561 0
.00036752 February 1, 1988 2436
.00036189 0
.00024671 February 2, 1988-2436
.00066215 0
.00028705 February 3, 1988 2434
.000319 0
.000267 February 4, 1988 2436
.000313 0
.000384 February 5, 1988 2434
.000353 0
.000667 l
February 6, 1988-2434
.000283 0
.00034 February 7, 1988 2436
-.0003025 0
.00041867 February 8, 1988 2436
.0002768 0
.0065222 February 9, 1989 2432
.00028 0
.000363 February 10, 1988 2436 0003 0
.000231 February 11, 1988 2436
,000304 0
.000195 February 12, 1988
?434.
.000304 0
000144 February'13, 1988 2430
.000239 0
.08016 February 14, 1988 2435
.0003052 0
.00809189 February 15, 1988 2436
.00028199 0
.00011193 L
February 16, 1988 2436
.00033 0
.0000647 L
February 17. 1938 2436 300258 0
.0000731 p
Februarv 18, 1988 2436
.000256 0
.0000401 l
February 19, 1988 2431
.0007.54 0
.0090362 L
February 20, 1983 2436
.000264 0
.0000536 L
February 21, 1980 2436
.00025<
0
<0008208 n
Februarv 22. 1988 2434
.000274 0
,6009281 j
F4 bru ary 23, 1988 2433
.000227 0
<0000284 February 24, 1998 2436
.000268 0
.0000238 February 25, 1988 2433
.000287 0
.0000109 February 26, 1988 0
.00629 0
.00000699 February 27, 1988 0
.30138 0
.00000467 February 28, 1988 1186
.0006 0
.000019657 83
'I I
.w a
DRILY CHEMISTRY DATA - 1988 1
DATE MNT(1) DOSE _EQ I131(1) MWT(2) DOSE _EQ_I131(2)
February I?, 1988 2180
.00037 0
.0000134
- March 1, 1988 2436
.0000219 0
.0000105 March'2, 1988 2435
.000401 0
.0000114 March 3,- 1988 2435
.000408 0
.00000681
. M arch 4, 1988 2435-
.000465 0
.00000628 March 5, 1988-2435
.000409 0
.0008139 March 6, 1988 2436
.000406 0
.00000903 March 7,.1988 2435
.000582 0
.0000139 March 8.
1988 2436
.000394 0
.00000617 March 9,.1988 2436
.000413 0
.00000944
' March 10, 1988 2435
.000396 0
.000013 March 11, 1988 2433
.000381 0
0 March 12, 1988-2435
.000422 0
0 March 13, 1988 2435
.000378 0
0 March 14, 1988 2435
.000367 0
0 March'15, 1988 2436
.000337 0
.00000381 March 16,-1988 2436
.000388 0
0
-March 17, 1988 2436
.000224 240
.000000878 Marc h 18, 1988 2434
.000373 130 0
March 19, 1988 2349
.00034 100
.0000349 March 73, 1988-2436
.000416 245
.0000599 March 21, 1988 2435
.000403 0
.000039 March 22, 1988 2435
.000307 40
.000023 i
March 23, 1988 2435
.000405 915
.00045' March 24, 1988 2432
.000378 460
.00029 March 25, 1988
.2435.
. 000428-1300
.000593
' March 26, 1988 2182
.000408 1937
.00082 March ;27, 1988 2436
.000466 2272
.000615 r.
March 28, 1988 2435
.00037 2355
.00111 M arc h. 29, 1988 2436
.000449 2436
.00106 March 30, 1989 2435
.000634 2435
.0011 March 31, 1988 2436
.00033 2412
.000693 April 1,
1988 1317
.00048 2409
.00127 April 2, 1988 2436
.000914 2023
.00111 April 3, 1988 2435
.000641 2435
.001 April 4,
1988 1284
.000777 2436
.0013 April 5,
1988 0
.00234 2411
.00117 April 6, 1988 0
.000553 2436
.00163 April 7,
1988 0
.000304 2435
.00155 April 8,
1988 0
.000184 2435
.00126 April 9,
1988 0
.000396 2260
.00164 Apeft 10, 1988 0
.00032 2420
.00154 April 11, 1908 100
.00101 2429
.00136 April 12,-1988 583
.0001407 2436
.001087 April 13, 1988 1218
.0005407 2436
.001239
. April 14, 1988 242G
.000591 2436
.J8128
.Apeti 15, 1988 2423
.09121 243(
.00126 t
Apell 16, 1988 2435
.00178 2436
,000775 April 17, 1998 2435
.00104 0
.000128 April 18, 1988 2435
.00106 0
.000106
-April 19, 1988 0
.00225 0
.0000411 April 20, 1988 0
.000719 2
.0008302 k
April 21,-1988 0
.000171 0
.0008309 April 22, 1998 0
.000111 0
.000027 April 23, 1988 0
.0001 0
.0000344 April 24,'1988 0
.00012 0
.300011 Aeril 25. 1988 0
.0000969 0
.00000873 26, 1988 0
.000279 0
.00000699 April ~ 27, 1988 0
.000664 0
.00000622 1
April 1
84 l
\\
m DAILY CHEMISTRY LATA - ~988 1
DATE MWT(1) DOSE _EQ I131(1) MWT(2) DOSE _EQ_1131(2)
A p r i.1 28, 1988 0
.000114 0
.00000495 April 29, 1988 0
.0000564 0
.00000394 April 30, 1988 0
.00001611 0
.000003627 May 1, 1988 0
.0000105 0
00000261 May 2, 1988 0
.0000698 0
.00000266 May 3, 1988 0
.0000118 0
.00000189 May 4, 1988 0
.0000105 0
.00009112 May 5, 1988 0
.00000834 0
.00000124 May 6, 1988 0
.00000509 0
.000000373 May 7, 1988 0
.00000371 0
.000000695 May 8, 1988 0
.00000314 0
.000000783 May 9,-1988~.
0
.00000272 0
.000000453-May.10, 1988 0
.00000236 0
.000000374 May 11, 1988 0
0 0
.0000002 May'12, 1998-0 0
0
.09000037 May 13, 1988 0
0 0
.000000374 May 14, 1988 0
0 0
0 May 15, 1988 0
0 24
.000017739 May 16, 1988.
0 0
24
.000016519 May 17, 1988 0
0 560
.00027931 May-18,>1988 0
.000505 0
.0000726 May-19, 1988 307
.000143 0
.0000152 May 20, 1998 0
.000253 0
.0000141 May 21, 1998 0
.000183 0
.0000139 May-22, 1988-0
.00023836 384
.000096432' May 23, 1988 0
.00022386 820
.00038119 May 24,-1988 0
.000139 1205
.000577 May'25,'1988 25
.00009415 1309
.00062586 May 26, 1988 1272
.000245 2344
.00111 n
May 27, 1988 1482
.00023 0
.000109 May 28, 1988 2436
.000844 535
.000317 May 29, 1988 2431
.000777 0
.0000365 May 30, 1988 2436
.00105 24
.0000435 May.31, 1988 2436
.000948 200
.000154 June 1,.1988 2435
.00163 486
.000319 June 2, 1988 2435
.00176 2282
.000936
' June 3, 1988 1151
.00446 2430
.00106 June 4, 1988 1207
.000814 2436
.00137 June 5, 1988 2015
.00144 2436
.00136 June-6, 1988 2402
.00182 2436
.00138 June 7, 1988 2434
.00199 2434
.00129 June 8, 1988 2435
.00275.
2436
.00107 June-f, 1988 2435
.00287 2435
.00129 June 10, 1980 2436
.00254 2436
.00165 June 11, 1988 2432'
.00253 2333-
.6013?
June 12, 1988 2434
.00264 2435
.031k6 June 13, 1980 2435
.00288 2435
.00129 June 14, 1988 2436
.00285 2436
.00127 L
June 15, 1988 2436
.00285 2436
.00135 June 16, 1980 2436
.00259 2436
.00138 June 17, 1998 2435
.00307 2436
.00176 l
June 18, 1998 2436
.00269 2429
.00153 June 19, 1968 2435
.00316 2435
.80133 June 20, 1988 2435
.00321 e42*
.00173 l'
June 21, 4988 2497
.00354 2431
.00135 Jutur 22, 1988 2436
,0234 2436
.00133
':.ne 2 3, 1980-2436
.00757 2436
.netd6 June 24, 198'3 2436
.00343 2436
.00139 June 25, i988 1358
.00639 2434
.00136 i
85 L--__-_-_____-__-__________-___-___=--________________
DAILY CHEMISTRY DATA - 1988
-o DATE MWT(1) DOSE _E0_I131(1) MWT(2) DOSE _E0_I131(2)
June 26, 1988 2435
.00392 2435
.00138 June 27, 1988 2435
.00361 2436
.00119 June 28, 1988 2436
.0034 2435
.00159 June 29, 1988 2432
.00342 2432
.00122 June 30, 1988 2434
.00366 2431
.00168 Julv 1,
1988 2435
.00393 2270
.00119 July 2, 1988 2433
.00395 2285
.000842 July 3, 1988 2434
.00246 2436
.00075658 l
July 4, 1988 2432
.00251 2432
.000902 l
July 5, 1988 2436
.00256 2436
.00136 July 6, 1988 1119
.00304 2433
.000726 July 7, 1988 2431
.00286 2434
.00146 July 8, 1988 2432
.00294 2435
.00131 July 9, 1988 2436
.00277 2233
.00191 July 10, 1988 2436
.00228 2436
.00137 July 11, 1988 2436
.00237 2436
.0013 July 12, 1988 2436
.00216 2436
.00135 July 13, 1988 2434
.0023 2436
.00137 July 14, 1988 2432
.00221 2435
.00127 July 15, 1988 2434
.00253 2436
.00124 July 16, 1980 2433
.0022 1630
.00137 July 17, 1988 2436
.00215 2436
.00141 July 18, 1988 2436
.00238 2436
.00127 July 19, 1988 2435
.00189 2436
.00132 July 20, 1988 2434
.00186 2436
.00132 July 21, 1988 2433
.00179 2436
.00125 July 22, 1983 2432
.00192 2435
.00126 July 23, 1988 2436
.0020334 2436
.0012621 July 24, 1988 2435
.002104 2436
.0014858 July 25, 1988 2435
.0023403 2436
.001493 July 26, 1988 2436
.00262 2436
.00132 July 27, 1988 2431
.0027 2436
.001 July 28, 1988 2433
.00225 2436
.00093 July 29, 1988 2436
.002396 2430
.001562 July 30, 1988 1336
.00396 2435
.0019 July 31, 1988 2135
.0019065 2435
.0018488 August 1,
1988 2436
.00206 2435
.00141 August 2, 1988 2430
.002168 2435
.001213 August 3,
1988 2432
.00193 2436
.00114 August 4,
1988 1331
.00243 2436
.00156 August 5, 1988 2430
.0023 0
.000717 August 6, 1988 2436
.00209 0
.000124 August 7,
1988 2436
.00219 1475
.000546 Augus'.
8, 1988 2426
.00201 2436
.00115 August 9, 1989 2436
.00305 I416
.00125 August 10, 1988 2434
.0036 24?2
.001414 August 11, 1988 2436
.00268 2435
.00142 August 12, 1985 2436
.00398 1370
.00096226 August 13, 1988 1390
.00349 2436
.00142 August 14, 1988 1745
.00216 2436
.00131 August 15, 1998 2405
.00619 J435
.00137 August 16, 1980 2436
.00739 2436
.00143 August 17, 1988 2434
.0065965 2436
.0014501 August 18, 1980 2434
.0085617 2432
.0014657 August 19, 1980 2435
.00425 2435
.00142 August 20, 1908 2293
.00449 2436
.00123 August 21, 1988 2433
.00377 2435
.00115 August 22, 1988 2430
.00377 2435
.00126 August 23, 1998 2433
.00472 2432
.0012 86
4-DAILY CHEMISTRY DATA - 1988
.DATE MWT(1) DOSE _EQ I131<1) MWT(2) DOSE _E0_I131(2) p.
August.24, 1988 2436
.00368.
2436
.00136 l
August. 25,'1988 2436
.00353 2436
.00115 August 26, 1988 2436
.00362 2436
.00144 August 27, 1988 2385
.00503
- 2435
.00127 August 28, 1988 2430
.00338 2435
.00129 August 29, 1988 2430
.00339 2435
.00135 August.30, 1988 2435-
.00338 2435
.00137 August'31, 1988-2436
.00325 2431
.00116 Septemoer 1, 1988 2436
.00348 2436
.00339 September 2, 1988 2435-
.00309 2435
.00207
. September 3,.1998 2436
.00479 1250
.00182 September 4, 19881 0
.00128 2436
.00186 September 5, 1988 650
.0102 2436
.00103 September 6, 1988 1555
.0045645 2436
.0014655 September 7, 1988 2427
.00576 2423
.00133' September 8, 1988 2432
.00401 2436
.00134 September 9,1 1988 2200
.00343 2436
.00138 September'10, 1988 2436
.00365 2436
.000148 September 11, 1988 2436
.00322 2418
.00135 September 12, 1988 2435
.00348 2435
.00131 September 13, 1988 2436
.00381 2434
.00184 September 14, 1988 2435
.06489 2436
.00105 September 15,-1988 2435
.00334 2436
.00142 l
September 16,-1988
'2435
.00346 2436
.0011 September 17, 1988 1960
.002318 2436
.001419 September 118, 1988; 2436
.00299.
2436
.00126 September 19, 1988 2430
.003019 2436
.001645 September 20, 1988 2432
.0028809 2436
.P012475 September 21, 1988 2436
.00328 2436
.00141 September 22, 1988 2435
.00287 2433
.00108 September 23, 1988-2425
.00287 2436
.00147 September 24, 1988 2436
.0029 2436
.0011 September 25, 1988 2436
.00222-2433
.00138
. September 26, 1988 2436
.00382 2430
.00155 September 27, 1988-876
.002 2434
.00151 September 28, 1988 0
.00316 2436
.00128 September 29, 1988.
0
.000383 2436
.00135 Sep *, ember 30, 1988 0
.000169 2435
.00143 October 1, 1988 0
.000142 2436
.0014 October 2, 1988 0
.000037819 2435
.0010247 October 3, 1988 0
.0044 2436
.00117 October 4, 1988 0
.000047 2436
.00122 October 5, 1988 0
.0000677 2436
.000876 October C, 1988
'O
.0000555 2436
.08132 Octobtr 7,
1988 0
.00C0405 2436
.00155 1
Octeber 8, 1980 0
.000026522 2436-
.0813003
(
October 9,_1988 0
.000018739 2436
.0011J14 October 10, 1988 0
.00317268 2436
.0014246 Oc.tober 11, 1988 0
.000052264 2436
.0012348 W
October 12, 1988 0
.000000351 2436
.00122 0:tober 13, 1988 0
.00000213 2436
.00135 Octeber 14, 1988 0
0 2436
.00154 October 15, 1936 0
0 2300
.0016
)
0 tober 16, 1000 0
0 2436
.00109 i
October'17, 1988 0
0 2436
.001G9 October 18, 1988 0
0 2436
.00133 Octcber 19, 1988 0
0 2436
.00136 October 20, 1988 0
0 2436
.001428 i
October 21, 1988 0
0 2436
.001407
n DRILY CHEMISTRY DATA - 1988 DATE NWT(1) DOSE _E0_I131(1) MWT(2) DOSE _EQ I131(2)
October 22, 1988 0
0 1993
.001156 October 23, 1988 0
0 2436
.00145 October 24,'1988 0
0 2436
.00129 I
October 25, 1980 0
0 2436
.00141 October 26, 1988 0
0 2435
.0012792 October 27, 1988 0
0 2435
.0013839 i
October 28, 1988 0
0 1990
.000907 October 29, 1988 0
0 2436
.00134 October 30, 1988 0
0 2436
.00132 October 31, 1988 0
0 2432
.000974 November 1,
1988 0
0 2436
.00131 November 2, 1988 0
0 2430
.00116 November 3, 1988 0
0 2436
.00164 November 4, 1988 0
0 2436
.00154 November 5, 1988 0
0 2436
.001419 November 6, 1988 0
0 2436
.00136 November 7, 1988 0
0 2435
.00112 November 8, 1988 0
0 2435
.00115 November 9, 1988 0
0 2436
.00109 November 10, 1988 0
0 2436
.00111 November 11, 1980 0
0 2436
.00134 November 12, 1980 0
0 2436
.00123 November 13, 1988 0
0 2436
.00128 November 14, 1988 0
0 2436
.00111 November 15, 1988 0
0 2436
.0013 November 16, 1988 0
0 2436
.00132 c
November 17, 1988 0
0 2436
.000898 Hovember 18, 1988 0
0 2435
.00107 November 19, 1988 0
0 1560
.0012 November 20, 1988 0
0 2431
.00083 November 21, 1988 0
0 2436
.000915 November 22, 1988 0
0 2436
.00103 November 23, 1988 0
0 2435
.060955 November 24, 1988 0
0 2435
.00104 November 25, 1988 0
0 2436
.00131 November 26, 1988 0
0 2436
.00101 November 27, 1988 0
0 2430
.00129 Hovember 28, 1988 0
0 2434
.00182 November 29, 1988 0
0 2432
.00128 November 30, 1988 0
0 2436
.001001 December 1,
1988 0
0 2436
.00101 December 2, 1988 0
0 2436
.0011 December 3, 1968 0
0 2436
.00131 December 4, 1988 0
0 2436
.00146 December 5, 1998 0
0 2436
.00151 December 6, 1988 0
0 2434
,0017 December 7,
1988 107
.09000982 2436
.00119 December 8, 1988 120
.00002762 2429
.001178 Decembsr 9, 1999 566
.00019123 2433
.0011933 Decembea 10, 198F 650
.000235 2300
.00112 December 11, 1988 767
.00124 2436
.000272 December 12, 1988 891
.000t99 2434
.00121 December 13, 1988 1406
,0A0418 2436
.00125 December 14, 1988 2430
.use637 920
.000899 Decet4ber 15, 1988 2427
.00053 0
.000121 Decemeer 16, 1986 2434
.00053 0
.000121 December 17, 1988 0
.00007262 0
.00004694 December 18, 1988 0
.000056474 0
.000048169 December 19, 1988 0
.000012173 243
.000056758 88
3 s.
DAILY CHEMISTRY DATA - 1988 DATE MWT(1) DOSE _E0_I131(1) MWT(2) DOSE _EQ f131(2)
Decemoer 20, 1988 483
.0000048629 500
.06036 December 21, 1988 1757
.0005 1430
.00102 December 22, 1980 2426
.000552 1998
.00104 December 23, 1988 2431
.000649 2340
.0011 l
December 24, 1988 2432
.00066096 2436
.0012437 December 25, 1988 2436
.000652 0
.0000662 December 26, 1988 2436
.0006447 1793
.00064453 December 27, 1988 2436
.000719 2275
.00102 December 28, 1988 2435
.000644 2428
.00107 December 29, 1988 2434
.000751 2429
.000934 December 30, 1988 2436
.000533 2430
.000973 December 31, 1988 2436
.000704 2374
.00114
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f' (Geo<gs F%er CorY pany ~
. 333 P: errant Avenue "
.)
1 Attacia; Georgia 30308 -
.1
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T@ pharm 404 SE3195 -
' Maing ddt ces
'. 40 inver*+ss Ce@ Pstrmsy '
Post Othco Box 129B
' Bam-egbam, Alabam t OS201 Teenhone 2U0 MA5MI th-v e n ep ut a, w n.
W. G. Hairston, lu Sert or Vice Pretornt i
NdCI9iir Oper,3traris n
HL-317 0590I i
J X7GJ17-H510 February 22, 1989 L
F
'.U. S.' Nuclear Regulatory Commission.
~
ATTN: Document Control Dne
. Washington, D. C.
20555 PLANT HATCH - UNITS 1,-2 NRC DOCKETS 50-321,'50-366 OPERATING LICENSES DPR-57, NPF-5 ANNUAL OPERATING REPORT FOR 1988 T
' Gentlemen:-
Enclosed-is the Annual Operating Report lfor. Plant Hatch Unit 1, Docket Number 50-321, and for Plant Hatch Unit 2, Docket Number 50-366.
f This report is submitted in accordance with the requirements of Technical
- Specifications' Sections 6.9.1.4 and 6.9.1.5.
O L
Sincerely, r n
. Q.). }b W. G. Hairston, III SJB/1s
Enclosure:
. Annual Operating Report for 1988
-I ci d ee next page) 1 i
w Ifl
4 t
)
J U. S. Nuclear Regult ory Commission February 22, 1980 Page Two
)
c: Georaia Power Comp _any Mr. H. C. Nix, General Manager - Plant Hatch Mr. L. T. Gucwa, Manager Nuclear Engineering and Licensing GO-NORMS U. S. Nuclear Regulatory Commission. Washington. D. C.
Mr. L. P. Crocker, Licensing Project Manager - Hatch U. S. Nuclear Reaulatory Commission. Reaion II Mr. M. L. Ernst, Acting Regional Administrator Mr. J. E. Menning, Senior Resident Inspector - Hatch 05901
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ -