ML17206A679

Annual Rept 1978.

ML17206A679
Person / Time
Site:	Saint Lucie
Issue date:	02/28/1979
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17206A678	List: ML17206A677 ML17206A679
References
NUDOCS 7903150295
Download: ML17206A679 (343)

Text

1978 P2'NUAL OPEPATING REPORT FLORIDA POhEP. & LIGHT CO."PANY

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ST. LUCXE UNIT l'/1 FEBRUARY, 1979 Abstract: This report is submitted in compliance with Technical Specifications 6.9.1.5, 4.4.11.3 and 10CFR50.59.

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INDEX SUBJECT PAGE NAKER Summary of Design Changes Per 10CFR50.59 Summary of Procedure Changes and 50 Special Tests Per 10CFR50.59 Core Barrel Movement Summary 51

'Steam Generator Tube Inservice Inspection 52 Radiation Exposure Summary 54

Page 1 DESlGN CHANGES On the following pages are descriptions, including a summary of the safety analyses, of the design changes implemented at St. Lucie Unit F1 during the period January 1, 1978 through December 31, 1978 in accor-dance with 10CPR50.59.

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.Page 2 Plant Change/Modification 39-76 PSL Unit Oil "BORIC ACID HEAT TRACING POWER SUPPLY FEEDER" The power supply for the Waste Management Heat Tracing Subsystem of the Boric Acid Heat Tracing System was modified so that it would autonatically be re-energized from the essential, safety related system (MCC lA-5) when powered by the lA diesel generator. This modification prevents boric acid precipitation in the Waste Management Heat Tracing Subsystem due to loss of power.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Peport has not been increased. Failure of the Waste Management Heat Tracing Subsystem has the same probability when automatically re-energized from the lA diesel generator as com-pared to the previous manual re-energizing from the 1A diesel generator.

The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The addition of the automatic re-energizing feature will not produce an accident or malfun'ction that has not already been evaluated.

3. The margin of safety as defined in the basis for technical specifications has not been decreased. The diesel generator has sufficient reserve capacity to accept the additional automatic load imposed by this modification.

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Plant Change/1'fodif ication 2-76 PSL Unit I/1 "INSTR1MENT AIR COHPRESSOR LOADED ON EMERGENCY HCCs" The power supplies for the turbine building instrument air compressors were relocated from the non-vital power distribution system (buses IfCC lAl and 1Bl) to the essential, safety related system (HCC 1A5 and 1B5).

Also, the cooling system for both compressors was modified. These changes allow the instrument air compressors to be powered from the diesel generators with minimal operator action.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal4unction of equipment important to safety previously evaluated in the Final, Safety Analysis Report has not been increased. This design change is similar to other non-class loads on safety related 1ICCs. (The instrument air system is not nuclear safety related.)

2. The possibility for an accident or malfunction of a different any evaluated previously in the Final Safety Analysis Report typ'han has not been created.

The margin of safety as defined in the bqsis for technical specifications has not been decreased. The loading of these air compressors is within the design rating of the diesel generators and is not automatic, but is controlled by the operator in the manual load group.

Page 4 Plant Change/Hodification 5-76 PSL Unit f/1 "BORIC ACID MAKEUP VALVE STATION MODIFICATION" Abandoned electrical boxes were removed and four cables rerouted at the valve station in the reactor auxiliary building. This was done to improve access for maintenance at the valve station.

This change is not an unreviewed 'safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. There was no change in function or quality of any system.

2~ The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has 'not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 5 Plant Change/Modification 01-76 PSL Unit /Il "REACTOR VESSEL HEAD SHIELDING FOR REFUELING" A review of the steady state radiation levels during refueling indicated approximately 10 Rem/Hr around the bottom edges of the reactor vessel head. A ring-shaped radiation shield consisting of structural steel and lead was fabricated and attached to the top of the missle shield. The reactor vessel head is placed on this ring-shaped structure during refueling to greatly reduce the local radiation levels in the vicinity of the reactor head.

This change is not an unreviewed safety questi'on because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the'inal Safety Analysis Report has not been increased. The shield is intended as a biological barrier for personnel during refueling and thus is not'described in the FSAR; therefore, it has no relation to equipment malfunction.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been 'created. No safety related equipment or design features were functionally altered during the installation of the radiation shield.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

age Pg.ant.Change/Modification 115-76 PSL Unit 81 "ADDITION OF TOTALIZER TO PLANT VENT STACK FLOW TRANSMITTER" A totalizer (requiring square root extractor) was added to the Plant Vent Stack Flow Transmitter (FT-26-1). This totalizer will aid operators in determining the activity released through the stack.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change

'is an addition to existing equipment and in no way affects the pro-

'ability of accidents. Consequences of an accident are not increased.

Better resolution of total air volume released is now available.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specif'cations has not been decreased. This change is not nuclear safety related.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 7 Plant Change/Modification 136-76 PSL Unit 81 "PRIMARY SAMPLE SYSTEti VALVE REPLACEMENT" Thirty-nine- valves in the Primary Sample System were replaced due to their unsuitability for the particular application required for primary sampling. These valves were malfunctioning due to damage caused by boric acid crystalization. The replacement valves are Nupro "UG" series bellow valves (or equivalent).

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The replace-ment valves are of stainless steel construction, have suitable pressure and temperature design ratings, and in general are better than or equal to the existing valves for the application.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report.

has not been created. The function of these 3/8 inch diameter sampling valves was not changed.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 8 Plant Change/1fodification 171-76 PSL Unit 81 "D/G CONTROL PANEL GASKETS AND DEMXSTERS" This added gaskets and demisters on relays and doors of diesel generator 1B control panel in proximity of the air intake. This change is to prevent moisture intrusion which could cause deterioration of components.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This modifi-cat'ion is not nuclear safety related. The externally mounted demisters and the gaskets increase the reliability of the control panel and therefore the diesel generator itself.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 9 Plant Change/Modification 199-76 PSL Unit ftl "CONTAINMENT FAN COOLERS VIBRATION ALARMS" A remote reset feature and time delay function was installed on the high vibration switch annunciator for each of the four (4) containment fan coolers. Prior to this installation, the vibration switch would be activated while starting the fan coolers due to the momentary high level and required entry into the containment building to 'ibration manually reset the local annunciator reset button. This installation will reduce the number of entries into the containment building, thus reducing operator radiation exposure.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Analysis Report has not been increased. The vibration switch provides only an annunciation function. Thus functionall'.-,

the modification is not a safety concern.

2. The possibility for an accident or malfunction 'of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report

Page 10 Plant Change/Modification 201-76 PSL Unit //1 "AUXILIARYFEEDMATER PUMP TURBINE GOVERNOR CONTROL BOX'ELOCATIO"I" The auxiliary feedwater pump 1C turbine governor control and central panels were relocated away from their original high moisture area (next to the steam turbine). This will prevent circuitry failure in the panels due to moisture ingression.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eva'luated in the Final Safety Analysis Report has not been increased. The only possible accident that could occur is failure of the steam driven auxiliary feed pump 1C. This has previously been evaluated in Table 10.5-1 of the FSAR.

2. 'he possibility for an accident or malfunction of a different type than any evaluated previously in t'ne Final Safety Analysis Report has not been created.

'he margin of safety as defined in the basis for technical specifi-cations has not been decreased. Only two of three auxiliary feed-water pumps are required to meet the bases of technical specification 3.7.1.2.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page ll Plant Change/Modification 215-77 PSL Unit 81 "HETRASCOPE INPUTS" Due to the new surveillance in technical specification 4.1.3.1.3 requiring a functional test on the CEA block circuit, this modification was submitted to install sliding-link terminal blocks and a multi-pin connector tu allow connecting the CEA position simulator to the metra c'athode ray tube scanner and the oscilloscope scanner on all input channels while allowing the backup scanner to remain in service.

This change is not an unreviewed s'afety question because:

1. The probability of occurrence or .the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The affected systems are not directly safety related and are used for regular surveillance only.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report

. has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 12 Plant Change/Modification 217-77 PSL Unit f/1 "COMPONENT COOLING HATER HEAT EXCHANGER STRAINER COVER" The strain'er cover gasket and stud washers on the component cooling water heat exchanger strainer (Intake cooling water side) were replaced with better quality material. The previous asbestos gasket material leaked requiring torquing of the stud bolts (to stop the leak) to the point where the stud bolt threads might strip and the stud washers would bend. Replacement material is 1/8" thick ethylene propylene gasket.

material and mild steel stud washers.

This change is not an unreviewed safety question because:

probability of occurrence or the consequences of an accident or

'he malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Repor . has not been increased. The replacement strainer cover gasket material -and stud washers are of better quality than the previous material.

The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The only equipment malfunction that could result from this change is a,minor leak at the strainer cover gasket. Table 9.2-2 of the Final Safety Analysis Report already analyses loss of a strainer.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 13 Plant Change/Modification 220-77 PSL Unit 81 "CONDENSER LEAKAGE DETECTION AND RESPONSE MODIFXCATEO lS" The following changes were made to provide immediate detection and response to condenser tube leaks:

a. Relocated blowdown flow controller to the control room.

b. Relocated the condenser quadrant cation conductivity recorder to the control room and added 'an alarm.

c. Modified control circuitry for condenser discharge valves, ~

vacuum breakers, and circulating water pumps, to provide rapid drain capability for the condenser.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change is not nuclear safety related. Xt will provide increased protection against plant equipment damage which could be caused by salt water intrusion.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased. 'i This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 14 Plant Change/Nodification 231-77 PSL Unit //1 SODIUM'i HYDROXIDE CONTAIh~ifENT SPRAY ADDITIVE SUBSYSTB'1 This documented the installation of the sodium hydroxide additive sub-system for containment spray system to satisfy condition I.l of the Facility Operating License. The new subsystem is designed to operate in conjunction with the containment spray system to remove radio-iodines from the containment atmosphere following the postulated LOCA. This system also replaces the function of the trisodium phosphate dodecahydrate storage baskets by providing containment sump ph control.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Refer to section 6.2.6 of the FSAR. Also refer to FPL letter to NRC, L-78-'78 dated 5/19/78.

The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. This change is considered in the FSAR and was a condition of the operating license.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 15 P3,ant Change/Modification 270-77 PSL Unit 81 "SHIELD BUILDING VENTILATION SYSTEM HEATERS" This change involved the installation of auxiliary heaters in each train of the shield building ventilation system. This item resolved Condit.'on of License I.2. These additional heaters provide humidity control to increase the effectiveness of methyl iodine removal by the filters.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. These auxiliary heaters are considered in the FSAR.

The possibility for an accident or malfunction of a different ty,".

than any evaluated previously in the Final Safety Analysis Report has -not been created. The backfit re q uirement for these heaters is described in the FSAR.

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3. The margin of safety as defined in the basis for technical specifi-

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cations has not been decreased.

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This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 16 Plant Change/Modification 295-77 PSL Unit /Jl "CONTAIRiENT HYDROGEN SAMPLING VALVES REPLACEMENT" A review of the purchase order for Unit I hydrogen sampling valves revealed that the installed valves had not been type tested for the post LOCA containment environment per FSAR Section 3.11 requirements.

This PC/M documents the installation of replacement valves to correct the deficiency. (Reference LER 8335-76-28).

This change is not an unreviewed safety question because:

1. The'robability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The replacement valves meet or exceed the applicable seismic class I, nuclear safety class, and post LOCA requirements.

2. The possibility for an accident or malfunction of a different type than any eval'uated previously in the Final Safety Analysis Report has not been created. No functional or quality requirements were changed.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 17 Plant Change/Modification 300-77 PSL Unit //1 "SPENT FUEL STORAGE RACK MODIFICATION" Modified spent fuel storage racks were installed to increase the spent fuel storage capability from 310 to 728 spent fuel assemblies. The increase in on-site storage capacity provided by this modification is required because of the lack of off-site spent fuel storage and/or reprocessing facilities.

A detailed description and evaluation of this change was given in the "Spent Fuel Storage Facility Modification Safety Analysis Report" which was submitted to 'the NRC on August 31, 1977 (FPL letter L-77-273).

Technical Specification Amendment No. 22 dated March 29, 1978, authorized use of the modified storage racks.

Page 18 Plant Change/Modification 304-77 PSL Unit /!1 "APPLICATION OF FLAM&fASTIC TO ELECTRICAL CABLES" Flamemastic 71A, a sprayable fireproof coating was applied to all electrical cables installed in cable trays located in the reactor auxiliary hailding, containment building, and turbine building. This fireproofing material was applied to preclude electrically initiated fires, to prevent fire spread, and to limit cable damage due to a fire in the vicinity of cable trays.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change is in accordance with the fire protection evaluation sent to the NRC by letter No. L-77-102 dated March 31, 1977.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical 'pecifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 19 Plant Change/Modification 323-77 PSL Unit /11 "ADDITION OF STANDOFFS TO NEh'UEL ELEVATOR" Standoffs were installed on the new fuel elevator to positively eliminate the possibility of placing a fuel assembly adjacent to a loaded new fuel elevator.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change enhances safety during fuel transfers.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

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Page 20 Plant Change/Hodification NNS-324-77 PSL Unit Pil "REFUELING HACHINE MODIFICATION" The vendor of the refueling machine recommended modifications to the bridge and trolley sections to improve the reliability and function of the machine. Spacers were added between the wheels and bearings on the bridge and trolley. Also, angle bracing was added under the bridge drive shafts.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The modification to the refueling machine is a minor improvement to existing equipment.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The modification to the refueling machine does not change the basic equipment design or function.

3; The margin of safety as defined in the basis for technical specifi-cations has not been decreased. The modification to the refueling machine improves the operation of the equipment. No structural changes on the machine were performed.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 21 Plant Change/Hodification 326-77 PSL Unit //1 "INSTALLATION OF NEUTRON STREWING SHIELDING" A reactor cavity neutron shield, consisting of nylon-neoprene covered bags holding ordinary light water, was installed. FP&L letter to NRC, L-76-406, dated November 29, 1976 describes the design and analysis for this change. This change satisfied Condition D of the Facility Operating License.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Refer to FP&L letters to the NRC, L-76-406 dated November 29, 1976 and L-77-245 dated April 3, 1977.

2. The ".ossibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been .created. This was addressed in the correspondence referenced above..

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

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Page 22 Plant Change/Modification 327-77 PSL Unit I/1 "RELOCATION OF STARTUP TRANSFORMER SECURITY FENCE" This change moved a security fence to exclude the startup transformers from the Unit 1 security area, added separate fences around each trans-former, and sealed or locked cable duct manholes. This will allow Unit 2 construction work in the area of these non-safety related transformers to be accomplished with adequate security for the transformers but without affecting Unit 1 overall security.

Thil change is not an unreviewed safety question because:

1. The'robability of occurrence or the, consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change does not affect nuclear safety related equipment.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The new fencing, gates, and administrative controls meet existing levels of security requirements.

.3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

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Page 23 Plant Change/Hodification 330-77 PSL Unit f/1 "SERVICE BUILDING 4'AREHOUSE SPRINKLER ALARMS" Alarm circuitry was installed on the service building warehouse sprinkler system to produce local and remote (Control Room) alarms whenever thn war house sprinkler system is activated. Also, an alarm will sound whenever- the isolation valve between the warehouse sprinkler system and the fire main comes off its fully o'en position.

This change is not an unreviewed safety question because:

1. The'robability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This modification does not involve safety related equipment.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been 'created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

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Page 24 Plant Change/Hodification 332-78 PSL Unit f/1 "ONE TON HOIST ADDITION TO REFUELING 1&CHINE" A one ton hoist with monorail was fabricated and installed on the refueling machine. The hoist is used to handle tools and light equip-ment that would otherwise require the use of the polar crane. This would divert the polar crane from other required heavy lifting services and possibly extend a refueling outage as a result of reduced polar crane availability.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of,an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. A report issued by the 'refueling machine vendor indicated that the refueling machine structural integrity is adequate to accept the additional loads imposed on it by the one ton auxiliary hoist and its load ader all conditions including a seismic event.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

0 Page 25 Plant Change/Hodification 335-78 PSL Unit Ol "CEA GROUP INTERLOCK BYPASS" A CEDS internal wiring change was made to permit temporary bypass of automatic functions which inhibit the regulating groups from being withdrawn in the group mode of control when the shutdown CEA's are not at their fully withdrawn position. This allowed implementation of a technical specification change concerning repositioning CEA's to minimize guide tube wear. (Reference FPL letter to NRC, L-78-7, dated January 4, 1978).

This change is not .an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3~ The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

Page 26 Plant Change/Modification 336-78 PSL Unit 81 "ADDITION OF ISOLATION VALVES FOR PRESSURE SWITCHES" Isolation valves were installed in the sensing lines for several pressure switches for the diesel generators. This change allows for calibr tion of the pressure sw tches without taking the entire diesel generator out of service. Small diameter (1/4 inch) tubing and valves were involved in this change.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. There is no change in quality or reliability of the affected @witches and tubing.

2. The pos'sibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. No functional change was made.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 27 Plant Change/i'fodification 339-78 PSL Unit /tl "INTAKE COOLING HATER PUNG'EPAIR" Because of pitting corrosion on the interior wetted surfaces of intake cooling water pump column sections, the pump manufacturer provided recommendations" for cleaning, repair welding, 'and coating the affected areas.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment im'portant to safety previously evaluated in the Final Safety Analysis Report has not been increased. The repair procedures and materials were consistent with the intake cooling water pump design and quality criteria.

2. The possibility for an accident or malfunction of a different typo.

than any evaluated previously in the Final Safety Analysis Report has not been created.

3.~ The margin of safety as defined in the basis for technical. specifi-cations has not been decreased.

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This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 28 Plant Change/31odification No. NNS-340-78 PSL Unit 81 "ADDITION OF gUICK EXHAUST VALVES TO TURBINE NON-RETURN VALVE ACTUATORS" t

Six quick exhaust valves were installed between existing solenoids and non-return valve actuator cylinders. The quick exhaust valves allow for the conventional installation of the exisiing solenoids and provide for a rapid flow of air through the solenoids; thereby, decreasing the valve stroke ti~es.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. ~

3.~ The margin of safety as defined in the basis for technical specifi-cations has not been decreased. ~

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

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Page 29 Plant Change/~ifodif ication 373-78 PSL Unit /ll "INSTALL HARD PIPING FOR SHUTDOVN COOLING PURIFICATION" Permanent piping was installed for the bypass of a portion of shutdown cooling flow through the letdown portion of the chemical volume and control system for purification. Previously, temporary hoses were connected for purification during shutdown cooling operation.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased, The design for the permanent shutdown purification piping is consistent with that of the interfacing systems.,

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. Paragraph 9.3.5.2.2 of the FSAR considers the installation of temporary piping in the CVCS to bypass a portion of the cooling flow during shutdown cooling through the letdown portion of CVCS for filtration and ion exchangers. This permanent design improves the method for implementation.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

Page 30 Plant Change/lIodification 375-78 PSL Unit Pil "REFUELXNG EQUXPtIENT SPEED CONTROL HODXFXCATXON" The motor drive panels for the refueling machine, spent fuel machine and control element assembly change machine were replaced with new motor drive panels containing improved motor drive controllers. The new motor d;ive controllers are more reliable, easier to adjust and have finer speed control capabilities.

This change is not an unreviewed safety question because:

The probability of occurren'ce or the consequences of an accident or malfunction of equipment important,to safety previously evaluated in the Final Safety Analysis Report has not been increased. This modification improves the reliability'f the equipment modified.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. There is no 'functional change involved with this repair. This modification is not nuclear safety related.

The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change does not represent a change to the facility as describ'ed in the Final Safety Analysis Report.

Page 31 Plant Change/Hodification 378-78 PSL Unit 81 "ADDITION OF NOISE SUPPRESSION DIODE FOR FCV-2161" An arc suppression diode was installed across the solenoid wiring for FCV-2161 to alleviate random noise generated when closing the valve.

FCV-2161 is a boric acid makeup flow control valve.

This change is not an unreviewed safety ouestion because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The new component does not decrease the quality of the electric circuitry involved. The reliability of FCV-2161 operation is enhanced.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. There is no functional change involved.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 32 Plant Change/Hodification 390-78 PSL Unit 81 "INSTALL INSULATION ON CCW PIPING" Anti-sweat type insulation was installed on the component cooling water supply piping to the reactor coolant pumps to alleviate corrosion.

Insulation was added to approximately 10 feet of piping near each pump.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The affected piping is not safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The new insulation is non-combustible.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 33 Plant Change/Hodification No. 391-78

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PSL Unit 81 "DISCHARGE CANAL HODIFICATIONS TO PREPARE FOR RAISING DIKE" This change involved maintenance excavation work at the exterior slope of circulating water system discharge canal banks. A gravel drainage layer was constructed and some "dressing vp" of the inside slope of the dike was performed where erosion had reshaped the slope. This work was done in preparation for raising the dikes (Reference PC/ki 430-78).

This'change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change is not nuclear safety related.

2. The possibility for an accident or m"lfunction or a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

Page 34 Plant Change/i~fodification No. 397-78 PSL Unit //1 "WORK STATIONS FOR CEA GUIDE TUBE MODIFICATIONS" Temporary equipment was installed at the spent fuel, pool to support modification of selected fuel assemblies. This PC/~i documents the tools, equipment, and procedures used to implement the modification. (Reference PC/:~ P4Z1-78).

This field modification is discussed in Section V. of CEN-90(F)-P which was submitted to the NRC in April, 1978.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

Page 35 Plant Change/Modification No. 398-77 PSL Unit //1 "REa40VAL OF PART LENGTH CONTROL ELE1KNT ASSEtfBLIES" The eight part length control element assemblies (PLCEA) were removed from the reactor and guide tube plugs were installed in the locations previously occupied by the PLCEAs. These plug assemblies preserve the dynamic operating characteristics of the reactor. A description of this change was provided to the NRC with FP6L letter No. L-78-125 dated April 12, 1978. Technical specification change to support removal of the PLCEAs was authorized in Amendment No. 27 dated Hay 26, 1978.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The remo;al of these PLCEAs has no effect'on the physics characterist,~.cs of the reactor. Also, there is no significant change in thermal or hydraulic effects by the use of the plug assemblies to replace the PLCEAs,

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The PLCEAs -served no nuclear safety function.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased. The use of PLCEAs had been prohibited by the technical specifications, and they had to be locked in the full out position during all reactor operations. Thus, this change was consistent with the technical specification basis.

Page 36 Plant Change/Hodification No. 405-78 PSL Unit Prl "TEtPORARY REACTOR CAVITY FILTRATION SYSTEl" This change provided for temporary tie-in of a reactor cavity filter assembly to improve water clarity in the cavity during refueling operations. (A permanent design for this change could not be imple-mented because of material availability problems. The permanent change is still planned.) The system was restored to its original configuration following refueling operations.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The design and quality of the flanges used to implement this temporary change were consistent with original system requirements.

'he possibility for an accident or malfunction of .a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

Page 37 Plant Change/Hodifiaction No. 406-78 PSL Unit 81 "NOISE REDUCTION ON INPUT TO ESF BISTABLES" A filter capacitor was added to the detector loops to reduce noise input to the bistables in the engineered safety features actuation system.

This change is not an unreviewed safety question because:

1. 'he probability of occurrence or the consequences of an accident or malfunction of equipment. important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change was recommended by the equipment vendor as a noise reduction improvement. New components are .equal in quality to the originally supplied components.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 38 Plant Change/Modification No. 415-78 PSL Unit (/1 "REPAIR FLOH RESTRICTION ORIFICE FOR INTAKE COOLING HATER SYSTEM" Restriction flow orifice SO-21-1B was modified as a flat plate orifice instead of a plate with vortex breaker. The cantilevered extension vortex breaker was deteriorated. This orifice is located in the "B" intake cooling water header downstream of the CCH heat exchangers.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The func-tion and flow characteristics of the orifice were not changed. An engineering evaluation showed that the vortex breaker (flow cone) is

not needed.

The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3~ The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not 'represent a change to the facility as described in the Final Safety Analysis Report.

Page 39 f

Plant Change/Modification No. 418-78 PSL Unit 81 "MAIN STEAM CHECK VALVE MODIFICATION" The main steam check valves disc bacl stop and .shaft bearings were modified to reduce stress levels, thereby augmenting service life. Examination of the valve internals revealed the need to reduce wear.

This change is not an unreviewed safety question because:

1. 'he probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the'inal Safety Analysis Report has not been increased. Melding procedures and materials conform to applicable codes and specifi-cations.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. This modification is consistent with the original valve design 'criteria, and does not alter the function or operating characteristics of the main steam check valves.

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.3. The margin of safety as defined in the basis for technical specifi; cations has not been decreased.

Page 40 Plant Change/1Iodification No. 419-70 PSL Unit f/1 "FUEL HANDLING EJUIPi~fENT CHAVGES FOR CEA PLUG CO:PATIBILITY" The long spent fuel handling tool and the CEA change mechanism guide plate was modified. These minor hardware modifications were required to make the existing fuel and CEA handling equipment compatible with the CEA plugs. These CEA plugs were installed after eight part length CEA's were removed (See PC/H 398-77).

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Only minor hardware modifications were performed. Changes to CEA handling equipment are non-nuclear safety related only.

2.'he .ossibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. On the basis that the modification to the spent fuel handling tool meets the intended design requirements for material compatibility, strength, testing and inspection, this change does not constitute an unreviewed safety question.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not"represent a change to the facility as described in the Final Safety Analysis Report.

Page 41 Plant Change/11odification No. 420-78 PSL Unit ~)l "INCORE INSTRUiiENT THIa1BLE REPAIR" This documents repair of three (3) incore flux detector thimbles (protective tubes). Replacement of 810 coarse threads with ]r'16 UNF threads, and tack welding of the collar to the thimbles reduces the likelihood of future failures. This item was reported to the NRC in Licensee Event Report 335-78-11 Update Report dated August 8, 1978.

This change is not an unreviewed safety question because:

1. The. probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The method of repair increased the reliability of the thimbles. (These thimbles are not pressure boundaries; they serve as "bushings" to support the small, flexible incore assemblies within the larger guide tubes).

The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. There is no functional change involved with this repair. This change is not nuclear safety related.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 42 Pl'ant Change/Ifodification No. 421-78 PSL Unit Pl "RETURN TO OPERATION WITH CEA GUIDE TUBE SLEEVES INSTALLED" This PC/li documents the reviews and approvals for reactor operation with selected fuel assemblies modified with control element assembly guide tube sleeves. (Implementation of the sleeving modification is covered in PC/lf 397-78). The stainless steel sleeves are needed to alleviate a CEA guide tube wear problem reported in Licensee Event Report 335-78-12 dated April 28, 1978.

A detailed report was provided in CEN-90 (F) P, "St. Lucie Unit 1 Reactor Operatio'n with Hodified CEA Guide Tubes", submitted to NRC in April, 1978.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Section VI of CEN-90 (F)'-P addresses this concern.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Peport has not been created. Section VI.C of CEN-90 (F)-P addresses this.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased. Section VI.D of CEN-90 (F)-P addresses this.

Page 43 Plant Change/Hodification No. 429-78 PSL Unit ill "TURTLE PROTECTION NET" A protective net was installed across the ocean intake canal to keep entrapped turtles away from the plant intake screens where they might be harmed, and to confine the entrapped turtles to a small area which will facilitate their removal.

This change is not an unreviewed safety question because:

1'. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change is not nuclear safety related.

2. The possibility for an accident or malfunction of a different tyq'"

than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

age, Plant Change/i~fodification No. 432-78 PSL Unit 81 "HODIFICATIONS TO OVERPRESSURE HITIGATING SYSTEi" The relief setpoint for the power operated relief valve in the over-pressure mitigating system was changed from a "sliding" setpoint for various temperatures to a constant setpo"..nt of 465 psia for all tempera-tures. This modification was requested by the NRC.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. This modification was a requirement of the NRC.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

Page 45 Plant Change/kfodification No. 433-78 PSL Unit i!1 INTAKE COOLING 'WATER STRAINER CHANGE This changed the bodies of the intake cooling'pumps lube water strainers from carbon steel to type 316 stainless steel. This change was made to enhance corrosion resistance.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. There is no change in function or quality of the strainers associated with this modification.

2. The p.~ssibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased by this material change.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 46 Plant Change/Modification No. 446-78 PSL Unit f!1 "RPS TH/LP PRETRIP SETPOINT" The hardware in the Core Protection Calculator //1, TH/LP Pretrip Circuit was'odified to change the TM/LP Pretiip setpoint from 100 pounds above the trip setpoint to 50 pounds above the trip setpoint.

This was done to eliminate the numerous alarms which were occurring.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Neither the TM/LP Pretrip Circuit nor any device to which it sends a signal is safety related or taken credit for in the safety analysis.

2. The p~ssibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined

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in the basis for technical specifi-cations has not been decreased. The TM/LP trip setpoint has not been changed. ~

Page 47 Plant Change/Modification No. 453-78 PSL Unit 81 "MODIFY PRESSURIZER RELIEF VALVES DISCHARGE PIPE VACUUM BREAKER" A one-half inch diameter check valve was installed to replace the one-half inch diameter hole to serve as a vacuum breaker in line 10-RC-822.

This piping vacuum breaker is located inside the pressurizer quench tank above the normal water level. The hole allowed steam to flow into the tank above the quenching water. The new check valve forces the leakage steam into the water, but still provides the vacuum breaker function.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change is not nuclear safety related. The check valve will allow for improved cooling of steam leakage from the safety relief valves.

2.. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The vacuum breaker function was not changed.

3. The margin of safety as defined in the basis for technical specifi-cations has -not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 48 Plant Change/i~fodification No. 455-78 PSL Unit Prl "DIESEL GENERATOR LOSS OF FIELD TRIP CIRCUIT HODIFICATION" The emergency diesel generator loss of field trip circuitry was revised by defeating S2 lockout relay contacts 12 and 12C to eliminate loss of field relay chatter.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the. Final Safety Analysis Report has not been increased. The loss of field trip circuit does not affect any accident evaluated in the FSAR. (The loss of field relay is taken out of the lockout circuit under accident conditions, including loss of offsite power.)

2. The ~ossibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis 'Report

, has not been created. This is a nonfunctional change made to increase the- reliability of the relay.

The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

Page 49 Plant Change/Modification No. 493-78 PSL Unit /!1 "SI TANK FILL LINE BRACE HODIFICATION" A 3 inch by 3 inch angle iron pipe brace for line 1-SI-123 was modified by using bolted flanges instead of welded construction. This facilitates removal of the brace, which is required periodically for entry to a steam ge'nerator manway.

This change, is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The design load bearing capability of the brace was maintained.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.. No functional change is involved.

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3.~ The margin of safety as defined in the basis for technical specifi-

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cations has not been decreased. ~

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 50 The'following summarizes changes in procedures and special tests

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conducted per 10CFR50.59 during the period January 1, 1978 through

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December 31, 1978:

PROCEDURE CHAiNGES I-3 PI/PSL-1 Design Control This procedure was revised to allow in-plant approval of design changes that are nest nuclear safety related. This was needed to avoid the time delay and expense of processing minor, non-safety related changes through the home office power plane engineering department. The method for pro-cessing nuclear safety related changes was not revised. Also, the method of 'control of design document updating for all changes was not affected.

This procedure revision affects only non-nuclear safety related design changes. (The depth of review to determine if a change is nuclear safety related or not was not lessened.) This procedure change does not con-stitute an unreviewed safety question.

SPECIAL TESTS Low Power Feedwater Control S stem Test, Letter of Instruction T-06.

.This test was performed to document the dynamic behavior of the steam generators to small level perturbations in the low power range, and also to verify the response of a proposed control system to changes in settings. The proposed system, which was designed by the NSSS Vendor, is to provide an automatic mode of contxol at low power levels to improve plant availability by minimizing reactor trips caused by steam generator level oscillations at low powers.

The test was conducted at low reactor power levels (less than 15%). Pro-cedural controls were specified to keep operations within normal plant operating limits. The test equipment has been removed. This test was determined not to involve an unreviewed safety question. The test data is being evaluated by the NSSS Vendor to determine if the proposed system will increase the reliability of the low power feedwate:- control operations.

Page 51 CORE BARREL NOVEKiNT

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Section 4.4.11.3 of PSL f/1 Technical Specifications requires the results of all periodic Amplitude Probability Distribution (APD) and Spectral Analysis (SA) monitoring to be included in this report.

Routine monitoring in 1978 included weekly APD processing and SA processing was done in February, June and October. SA measurements in June included analysis at nominal thermal power levels of 20%,

50%, 80% and 100% at the beginning of fuel cycle 2. At no time were the Alert or Action levels exceeded.

As previously observed and reported in 1977 the R."fS levels of all excore neutron detector signals continued to show a gradual increase throughout 1978 with the exception of a slight decrease following refueling.

The increase, confirmed by independent APD and SA analysis, amounts to about 50% above the levels of December 1977. As in 1977 the greatest portion of signal increase was in the frequency range of 1-4 hertz. This phenomenon has been observed at other PWR's and has been attributed to the increase in fuel element vibration worth due to the decreasing boron concentration with core burnup. Like-wise the observed step decrease in the band of 1-4 hertz following refueling may be attributed to the step increase in boron concen-tration at the. beginning of the new fuel cycle.

The observed net increase for 1978, if imputed to be entirely due to core barrel motion would signify less than 4 mils %1S motion at the core midplane.

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Page 52 "STEAtf GENERATOR TUBE INSPECTIONS" An inservice eddy current examination of selected tubes in the No. 1A St. Lucie Unit No. 1 Steam Generator was performed during the period of April 2 through April 7, 1978, by C-E Power Systems, Systems Integrity Services personnel. The inspection was conducted in accordance with C-E Test Procedures Nos. 00000-ESS-105, Revision 00 and 00000-ESS-070, Revision 01, and satisfied the requirements of the St. Lucie Plant Technical Specification 3/4 4-5 and the ASME Boiler and Pressure Vessel Code, Section XI, 1974 Edition through the Summer 1976 Addenda.

The inspection program consisted of 400kHz testing for tube wall anomalies and 25kHz testing for sludge accumulation on the secondary side of the

,tube sheet. Table I details the number of tubes examined in each generator.

Selection of tubes to be examined was based on an evaluation of data taken previously in other steam generators in service. Additionally, when requested by the Data analyst, certain tubes were re-examined at 100kHz for confirmation of the 400kHz data.

The data from the inspection was recorded on magnetic recording tape and strip charts. These recordings were evaluated and the results recorded on Eddy Current Examination Report Sheets. Of the tubes examined none were found to have re ortable wall de radation. (>20% of wall). No tubes were lu ed.

A number of tubes were found to be dented at the point where the tube passes through one of the drilled hole support plates. A summary of the numbers and magnitude of these dents is included as Table I.

The 25kHz inspection indicated up to 4" of sludge on the Hot Side and up to 3.5" of sludge on the Cold Side of the secondary side of the tube sheet.

Information on the results of this first Unit 1 steam generator inservice inspection were reported to NRC in FP&L letter No. L-78-249 dated July 31, 1978. In addition, more detailed information is available at the plant site.

Page 53 TABLE I SlMGKY OF EDDY CURRENT TEST RESULTS STEAM GENERATOR 1A INSPECTION CONDUCTED APRIL 1978 TOTAL TUBES (by design) 8,519 'U'ubes TUBES THROUGH PARTIAL SUPPORT PLATE No. 9 2,225 (26.1%)

TUBES THROUGH PARTIAL SUPPORT PLATE No. 10 771 (9.1%)

TUBES EXAHINED No. yr of. of

.Tubes Total HOT SIDE DEFECT DETECTION 583 6.8% No Tube 4'all Degradation Indications

.COLD SIDE DEFECT DETECTION 100 1.2% No Tube Mall Degradation Indications HOT SIDE SLUDGE MEASUM21ENT 62 0.73% Maximum 4 Inches COLD SIDE SLUDGE MEASUR&1ENT 62 0.73% Maximum 3.5 Inches

P APPEND A STANDARD FORMAT FOR REPORTING NUMBER OF PERSONNEL AND MAN-REIVl SY.IlORK.AND JOB FuNCTION Yunrbcr of Pcrsonncl (> 100 rnrcm) Total htan Rcm-Contract EVorkcrs Contract Workers ls'ork tv, Jnb Function Station Employccs UtilityEmployccs and Others Station Employccs UtilityL'mployccs and Others Reactor Operations h Surveillance maintenance Pcrsonncl 0 0 0 0 0 0 Operatint; Personnel 0 00 0 0 Ilealth Physics Pcrsonncl ~

7 0 0 3.370 0 0 Supcnisory Pcrsonncl 0 0 1.680 0 0 Engineerint; l'crsnnncl 0 0 0 0

Routine htaintenanec htainteranec Pcrwnncl 96 39 138 43.430 18.310 55.220 Operating Personnel 0 0 .330 0 0 llcalth Physics Pcrsonncl 0 6.690 0 .740 Supervisory Pcrsonncl

~ Ent;inccting l'crsonncl .

Insen ice inspection hlaintcnanec Personnel 1.590 2.510 .660 Operalinp Pcrsonncl 0 0 0 0 0 llcalth Physics Pcrsonncl 0 0 0 l. 90 Supervisory Personnel 0 . 050 0 0 Entdneerini, Pcrsonncl 0 00 Special hlaintcnancc Maintcnancc Personnel 0 0 59 0 0 25..200 Opcratint; Personnel Ilcalth Physics Personnel Supervisory Personnel Engineering Pcrsonncl lVastc Processing hlaintcnancc Pcrsnnncl 15 0 4.580 0 .140 Operating Pctsonncl llcalth Physics Pcrsonncl Supervisory Pcrsonncl Engincerint; Personnel Refueling Maintcnancc Pcrsonncl 38 53 0 17.960 38.430 0 Operatinr. Pcrsonncl 10 0 .2.1 0 0 Ilcalth Physics l'crsonncl 10 Supervisory Pcrsonncl Engineering Pcrsonncl 0 . 53.0 0 20 TOTE L hfalntcnancc Pcrsnnncl 105 84 191 67.560 59. 250 81.220 Opcralint', l'crsonncl 31 0 18.390 0 0 llcalth Physics Personnel 0- 1 ~ 500 0 10.7 0 Supervisory Pcrsonncl 1 . 0 .0 0 Enginccrine Personnel 0 56 2.700 0 29. 330 Grand Total 171 86 279 113.500 5 .510 122. 30

C~

L I

1977 ANNUAL OPERATING REPORT FLORIDA POWER AND LIGHT COMPANY ST. LUCIE UNIT f/1 February, 1978 DocketNW co< v. ~ro~oaazp B."a~ZWMWofDocument KCuL07Ci.7 t!Z iH HE

E CI

INDEX I II LE PAGE Narrative Summary Design Changes Procedure Changes 58 Tests 59 Core Barrel Movement 60 Steam Generator Tube Inspections Radiation Exposure 62 List of Abbreviations 63

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,L

Page 1

SUMMARY

OF OPERATING EXPERIENCE The following is a summary of plant operations including pertinent items of interest chronologically for the period 1/1/77 thru 12/31/77.

1/1/77 at 50% power. (Power Ascension Testing) 1'/3/77 PM Rx trip. Loss of "B" S/G feed. pump.

1/4/77 3:40 AM Rx critical.

1/5/77 Unit line-increasing

':15'5 12:07 AM on power to 50%.

1/9/77 11:50 AM Increasing power to 60%.

1/10/77 5 AM Increasing power to 70%.

1/ll/77 1:00 PM Increasing power to 80%.

1/24/77 2:00 PM Decreasing power to 50% to clean cond-ensate pump strainers.

1/25/77 6:52 AM Increasing power to 80%.

2/1/77 (1) 10:48 AM Planned reactor trip. Partial loss of flow test.

(2) 11:48 AM Rx critical.

(3) Unit on line-increasing power to 40%.

2/2/77 (1) 4:04 AM Planned Rx trip. Total loss of flow test.

(2) 7:30 PM Rx critical.

2/3/77 (1) 00:03 AM Unit on line-increasing power to 20%;

(2) 3:05 AM Planned Rx trip-loss of offsite power.

(3) 7:53 AM Unit on line-increasing power to 90%.

2/4/77 (1) ll:15 AM Dropped rod 860-reduced power to 50%.

Recovered rod-increasing power to 90%.

(2) 9:00 PM Rx at%0% power.

2/20/77 (1) 10:15 AM Planned Rx trip-manual turbine trip at 100% power.

(2) 10:55 PM Rx critical.

Page 2

SUMMARY

OF OPERATING EXPERIENCE 2/21/77 (1) 12: 05 AM Unit on line.

(4) 2:25 PM Planned Rx trip. Manual loss of load at 60% power.

2/22/77 (1) 12:16 AM Rx critical.

(2) 8:47 AM Unit on line-increasing power to 100%.

'/4/77 8:15 PM Reduce power'o 50% to clean 1B S/G feed pump strainers.

3/5/77 1:30 PM 1B S/G feed pump back in service-increasing power to 100%.

3/7/77 (1) 2:04 AM Discovered CEA 820 stuck at 125" while performing FLCEA periodic test-decreasing power'to 70%.

(2) 2:50 AM Dropped rod 823 while aligning with Rod i/20-deer'easing power to it 60%.

(3) 4:15 AM Rods //20 6 f323 repaired-increasing power to 100%.

3/ll/77 (1) 5: 00 PM Containment emergency air lock interior door failed leak rate test-reducing power to enter containment and inspect.

(2) 10:04 PM Unit off line to replace airlock gasket.

3/12/77 '1) 6:38 AM Unit on line-increasing power to 30%.

(2) 5:30 PM Completed load swing test at 30%-

increasing power to 50%.

3/18/77 2:45 PM Completed load swing test at 50%--- in-creasing power to 90%.

3/19/77 6:40 PM Completed load swing. test at 90%

increasing power to 100%.

4/3/77 6:08 AM Rx manual trip-fire in generator lead box. Fire caused by H2 leak on generator lead box seal. Seal replaced.

4/11/77 (1) 7:04 AM Rx critical.

(2) 2:09 AM Unit on line-increasing power to 100%.

4/15/77 3 39AM Manual Rx trip due to loss of cooling water to RCP seals caused by loss of containment air compressor. See PCM 258-77.

6 Page 3 1 ~

SUMMARY

OF OPERATING EXPERIENCE 4/29/77 (1) ll:40 AH Rx critical.

(2) 1:10 PM Unit on line-increasing power to 100%.

4/30/77 (1) 1:54 AM Channel "A" pressurizer pressure trans-mitter failed high.

(2) 3:18 AH Started reducing power to 12% to make containment entry to change transmitter.

(3) 7:30 AM Replacement of transmitter completed-increasing power to 100%.

4/30/77 (1) 1:04 PH Reactor tripped on loss of load. Indi-cation in Control Room of loss of volt-age,regulator. Operator opened exciter field-breaker. Indication caused by fail-ure of B heater drain pump L.C.U. which caused extreme vibration in discharge line.

Line is near the north wall of turbine building switchgear room and vibration caused the 6.9 KV undervoltage relays to trip.

RCP's tripped.

(2) 3:30 PH Reactor critical.

(3) 5:42 PM Unit on line-increasing power to 100%.

5/10/77 (1) 6:25 PM Started reducing power to take unit off line due to small leak on 1A2 S.I.T. line.

(2) 9:50 PM Unit off line.

5/12/77 4:15 PM Haintenance completed repair of 1A2 S.I.T.

drain valve line.

5/13/77 (1) 7:58 AH Rx critical.

(2) 1:23 PH Unit on line-increasing power to 100%.'x 5/16/77 (1) 10:23 AM trip due to system voltage fluctuations.

(2) 8:04 PM Rx critical.

(3) 9:58 PH Unit on line-increasing power to 100%.

5/19/77 9:30 PM Reducing power to 50% to clean steam generator feed pumps 1A 6 18 suction strainers.

5/20/77 5:55 PM Started increasing power to 100%.

'5/27/77 (1) 3:10 PH Dropped rod 839-turbine runback to " -70%.

lA SGFP tripped. Rx tripped on low S/G level.

(2) 6:28 PM Rx critical.

(3) 8:13 PM Unit on line-increasing power to 100%.

a Page 4

SUMMARY

OF OPERATING EXPERIENCE I

5/31/77 3:40 PM Turbine/reactor trip-loss of generator excitation caused by failure of permanent magnet generator. Unit shutdown to re-place permanent magnetic generator (PMG).

6/5/77 (1) ll:13 AM 'Reactor critical.

(2) 8:38 PM Unit on line-increasing power to 100%.

6/7/77 (1) 4:16 PM Reactor trip-operator error. Tripped operating motor generator set.

(2) 6:00 PM Reactor critical.

(3) 8:04 PM Unit on line-increasing power to 100%.

6/13/77 (1) 5: 04 AM Reduce load to "-50% to replace 1B turbine cooling water pump seal.

(2) 11:59 PM Seal repaired-increasing power to 100%.

6/25/77 10:39 PM "A" condensate pump expansion joint failed, A condensate pump tripped, A SGFP tripped on low suction, Rx tripped on low S/G level. Replaced expansion joint.

6/26/77 (1) 2:10 AM Rx critical.

(2) 1:17 PM Unit on line-increasing power to 100%.

6/27/77 10:20 PM Reducing power to come off line to iso-late PCV 1100F. (Pressurizer Spray Valve) 6/28/77 3:40 PM Unit back on line-increasing power to 100%.

'7/4/77 10:18 PM Reducing power to come off line to 'repair PCV 1100F.

7/5/77 8:00 AM Unit back on line-increasing power to 100%.

7/8/77 (1) 2:30 AM Reducing power to come off line to repair PCV's 1100 F 6 E.

(2) 6:41 AM Rx shutdown.

(3) 3:23 PM Rx critical.

(4) 4:52 PM Unit on line-increasing power to 100%.

7/10/77 5:52 PM Reducing power to come off line to repair PCV 1100F.

7/ll/77 5:59 AM Unit back on line-increasing power to 100%.

7/12/77 8:56 PM Reducing power to come off line to repair PCV 1100 E & F.

I Page 5

SUMMARY

OF OPERATING EXPERIENCE 7/13/77 6 54 AM On line-increasing power to 100%.

7/29/77 5:25 PM Reducing power to come off line to repair PCV 1100 E 6 F. (See PCM 268-77) 7/30/77 2:18 PM Unit back on line-increasing power to 100%.

8/30/77 8:45 PM Reducing power to 50% to work on 1A feedwater pump.

8/31/77 (1) 1:11 PM Rx tripped on low S/G level. 1B S/G feedwater pump tripped when 1A S/G feedwater pump start was attempted.

(2) 3:11 PM Rx critical.

(3) 4:47 PM On line-increasing power to 100%.

9/24/77 6:25 PM Reducing load to take unit off line for planned outage. See PCM 264-77.

10/10/77 (1) 5:40 AM Rx critical.

(2) 11:59 AM Unit on line-increasing power to 100%.

10/28/77 (1) 12:45 PM Dropped Rod 856.

(2) 1:30 PM Rx power <70%.

(3) 2:38 PM Rod 856 retrieved.

(4) 6:00 PM Due to flux

-"75%.

tilt maintaining power at

'10/30/77 (1) 1:55 AM Reactor Engineering calculated flux tilt to be within Tech Spec limits-increasing 'power to 100%.

11/17/77 7:05 PM Reducing power to 50% to clean lA SGFP strainer.

11/18/77 6:50 AM Increasing power to 100%.

11/22/77 (1) 10:32 AM Rx trip due to loss of 1A SGFP.

(2) 12:40 PM Rx critical.

(3) 2:27 PM Unit on line-increasing power to 100%.

12/20/77 (1) 5:31 AM Rx trip-loss of load caused by loss of excitation. *

(2) 1:49 PM Rx critical.

(3) 3:59 PM Unit on line-increasing power to 100%.

12/31/77 at 100% power.

• Cause later found to be intermittent open in fuse in excitation circuit.

Page 6 DESIGN CHANGES On the following pages are descriptions, including a summary of the safety analyses, of the design changes implemented at St. Lucie Unit bl during the period January 1, 1977 through December 31, 1977.

Page 7 Plant Change/Modification 4-76 Unit 81 "INSTALL ACCESS PLATPORMS AT PRESSURIZER CUBICLE" Access platforms were installed at the pressurizer cubicle to facilitate maintenance and operation of valves and instruments near the top of the cubicle.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The platforms are not required for safe shutdown of the plant.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the final Safety Analysis Report has not been created.

The new platforms were designed and,installed to Seismic Class I requirements to preclude failure as a result of a postulated earthquake.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Pinal Safety Analysis Report.

4ll Cl

Page 8 Plant Change/Modification 37-76 Unit f/1 "STEAM GENERATOR Hl WATER LEVEL TURBINE TRIP CIRCUIT" Control circuitry was added to provide a trip signal on steam generator hi water level to prevent the possibility of water carry over to the steam turbine.

I This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Repoxt has not, been increased. Where safety related components were replaced (level indicator controllers),

devices of the same basic type and equivalent qualification were used.

The low water level trip features were not affected.

2. The possibility for an accident or malfunction of a different type than" any evaluated previously in the Final Safety Analysis Report has not been created.

3. The maxgin of safety as defined in the basis for technical specifica-tions has not been decreased. This modification does not affect any technical specification basis.

Page 9 Plant Change/Modification No. 83-76 Unit f31 "RELOCATION OF 4.16 KV AND 6.9 KV SWITCHGEAR RELAYS" Hinged armature relays were originally mounted on the 6.9 KV and 4.16 KV switchgear door panels in such a manner that inadvertent relay contact actuations could possibly result from slamming closed the switchgear doors. To prevent this, those relays with control or trip functions were relocated to the side walls of the breaker cubicles or to separate boxes outside the switch-gear cubicles. Approximately 51 Westinghouse type "SG" and 6 General Electric type "HGA" relays were relocated.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analy-sis Report has not been increased. Replacement mat-erials meet or exceed the requirements of the original components. The reliability of the subject relays is increased.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This modification does not change the function of any safety related equipment.

3~ The margin of safety as defined in the basis for Technical Specifications has not been decreased.

This change does not represent a change to the facility as de-scribed in the Final Safety Analysis Report.

Page 10 Plant Change/Modification No. 86-76 Unit 81 "BUTTERFLY VALVES CONTROLS MODIFICATION" The control circuitry for 31 motor operated valves was revised to utilize a limit switch rather than a torque switch to stop the motor operator in the valve closed position. This change conforms to valve vendor recommendations that rubber seated butterfly valves be position seated instead- of torque seated.

The torque switch, which actuates when the operator reaches its mechanical stop, was retained as a backup to the limit switch control.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Anal-ysis Report has not been increased. The new limit switches will stop the valves at the same relative position as the original torque switches. To pro-vide a backup to control valve closure, the torque switches were retained.

2 ~ The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

There is no change in control scheme involved. The use of a limit switch with torque switch backup for control will increase the reliability of each valve closure operation and reduce operator maintenance.

3. The margin of safety as defined in the basis for Technical Specifications has not been decreased.

This change does not represent a change to the facility as de-scribed in the Final Safety Analysis Report.

Page 11 Plant Change/Modificationl08-76 PSL Unit 81 "INSTALL SPARE 5 KV CABLES" These cables were installed to meet Technical Specification 4.8.1.1.3.b.2 which requires three (3) spare cables for insulation resistance tests.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Rep'ort has not been in-creased.

These cables, which will not be used for plant equipment, are designed and installed to the same requirements as the original cables. They allow testing required by the NRC to demonstrate the satisfactory condition of the Class 1E underground cable system.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change is required by the Technical Specifications.

This change does not represent a change in the facility as described in the Final Safety Analysis Report.

Page 12 Plant Change/Modification No. 169-76 PSL Unit 1 "EQUIPMENT DRAIN TANK 1A STRAINER MODIFICATION" The single element equipment drain tank strainer (S6904) was replaced with a dual basket type strainer with a differential pressure alarm.

This change allows the cleaning of a clogged strainer without interrupting the ability to pump into the equipment drain tank from the reactor cavity sump or the engineered safeguards room sump.

'Zhis change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change is not nuclear safety related. New components are compatible with the original systems affected.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. This change provides operational flexibility without a change of function.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page Plant Change/Modification No. 170-76 Vnit b'1

'"CODIFICATION TO REACTOR CAVITY SAP LEAK DETECTION SYSTEM" This change involved replacement of 1 inch tubing from collection boxes to wier tank with stainless steel gutters, modification of the weir tank, and relocation of a flow transmitter to outside of the sump. These changes were made because of problems experienced with maintenance and calibration of the reactor cavity sump leak detection system.

This change is not an unreviewed safety question because:

1 ~ The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Replacement components were de-signed and installed to equivalent or better standards as the originals.

2~ The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

There is no change in overall function or quality of the system.

3 ~ The margin of safety as defined in the basis for Technical Specifications has not been decreased.

The modified system is calibrated to detect RCS leakage within Technical Specification limits.

This change does not represent a change to the facility as de-scribed in the Final Safety Analysis Report.

Page 14

'LANT CHANGE/MODIFICATION NO. 175-76 PSL UNIT //1 "REDUCE NEEDLE FLUCTUATION ON RADIATION MONITORING INSTRUMENTS" This change was performed for the plant radioactive waste and process radiation monitors. It involved changing capacitors in the log count ratemeter circuit board to increase the circuit time constant. This eliminated the many spurious alarms caused by needle (signal) fluctua-tion.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The time response of the instruments is not significantly changed.

These instruments are not discussed in the Accident Analysis.

Eliminating spurious alarms enhances equipment reliability and removes a source of distraction for the operators.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

An equivalent part of changed capacitance is installed. No new accidents or malfunctions are created.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change does not represent a change in the facility as described in the Final Safety Analysis Report.

Page 15 Plant Change/Modification No. 187-76 Unit f/1 "ADDITION OF VENT VALVES FOR CCW PUMPS" Manual globe valves were installed at the 3/4 inch threaded vent connections for each component cooling water pump casing. This was done to provide a convenient means to vent the pump casing section. The vent connections were originally provided with screwed plugs only.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analy-sis Report has not been increased. New materials were purchased and installed to standards which meet or exceed original pump design specifications. Because of the small mass of the new 3/4 inch valves and lightweight tubing, original seismic design was not affected.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Failure of a vent valve would be similar to failure of the originally installed vent plugs. The FSAR single failure analysis for the CCW system is not af-fected.

3~ The margin of safety as, defined in the basis for Technical Specifications has not been decreased.

This modification reduces the time required for vent-ing component cooling water pump casings following maintenance activities.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 16 Plant Change/Modification 188-76 PSL Unit 81 "ADDITION OF CHECK VALVES IN ICW Ply LUBE WATER PIPING" Check valves were installed in each 1 inch bearing lube water supply line to the intake cooling water pumps. This was done to prevent the draining of the lube water piping when lube water supply is shutdown.

The draining could lead to the formation of air voids in the piping causing pump bearing overheating.

~'his change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. All new materials meet or exceed the original design requirements for the intake cooling water system. The new check valves will improve the reliability of the bearing lubrication system.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. There are no functional changes associated with this modification.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 17 Plant Change/Modification No. 189-76 PSL Unit I/1 "INSTALLATION OF VALVE FOR ILRT PRESSURE CONNECTION" Gate valve I-V00101(612) and associated supports were installed at containment penetration number 54 in accordance with Amendment No. 2 to the Facility Operating License. This valve was part of original plant design for ILRT pressurization. Installation was deferred because of material availability.

This change is not an unreviewed safety 'question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This containment penetration valve is part of the plant design as described in the FSAR.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described, in the Final Safety Analysis Report.

Page 18 PLANT CHANGE/MODIFICATION NO. 191-76 PSL UNIT 81 MODIFY INTAKE STRUCTURE CRANE PLATFORM The original crane platform and ladder allowed personnel to work too close to transmission lines over the crane. The access ladder and platform, needed to allow maintenance on the crane, were relocated to ensure compliance with OSHA requirements.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The crane is not safety related but is located near some safety related equipment. The new platforms were designed and built to the same as or better specifications as the original.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

'Page 19 Plant Change/Modification No. 193-76 PSL Unit 1 "INSTALLATION OF REACTOR COOLANT PUMP OIL DRIP PANS" Oil drip pans were installed for the reactor coolant pump motors to reduce the possibility of a fire due to uncollected oil leakage. Also, upper oil reservoir gaskets were replaced in accordance with the vendor's recommendations.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change is not nuclear safety related. The oil drip pans were fabricated and installed to Seismic Class I requirements, however, due to the mounting location.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report,has not'been created. This modification does not affect the function or quality of any plant system.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

'This cha'nge does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 20 Plant Change/Modification No. 195-76 PSL Unit 81 "STEAM GENERATOR BLOWDOWN TREATMENT FACILITY ELECTRICAL TIE-IN" This change tied in several Steam Generator Blowdown 'Treatment Facility electrical interfaces to Unit 1. 'ix valve positions indicating lights and annunciation for high radiation levels were added in the Unit 1 control room. The blowdown treatment facility is a requirement of our license and is described in the FSAR.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. These tie-ins implement the design described in the FSAR. b

2. The possibility for an accident or malfunction of a different any evaluated previously in the Final Safety Analysis Report type'han has not been created. The systems affected are not nuclear safety related.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 21 Plant Change/Modification No. 210-76 PSL Unit //1 "ROUTE STEAM DRAINS AWAY FROM AUXILIARYFEEDWATER PUMP" Several drains associated with auxiliary feedwater pump 1C steam turbine were routed away from the pump area. This change was made to decrease

,the possibility of corrosion caused by collection of moisture in control components.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The. drain line piping is not nuclear safety r'elated: No functional changes were made.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The new drain piping is small in diameter and will not carry high energy fluids.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

,L Page 22 Plant Change/Modification No. 211-76 PSL UNIT 81 "LETDOWN CONTROL VALVES SEAL RING CHANGE" The pressure seal rings for letdown control valves LCV-2110 P&Q were changed from teflon coated rings to silver plated rings. This change was recommended by the valve vendor to alleviate leakage problems caused by corrosion of the seal rings.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The replacement parts were provided by the original vendor to the same specifications as the original parts.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The change does not affect the'function or quality of the letdown control valves.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 23 Plant Change/Modification No. 213-77 PSL Unit //1 "TURBINE RUNBACK CIRCUITRY REVISION" .

The heater drain pumps turbine runback input was revised to a setting of 92% of full load upon tripping of both pumps. Also, a feedwater/steam flow mismatch contact was deleted from the steam generator feedwater pumps tur-bine runback circuit because of a relay chatter problem. This modification will provide automatic runback to required levels utilizing a simpler, more reliable circuit.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated .in the Final Safety Analysis Report has not been increased. This change affects only the non safety related turbine runback circuits.

P

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

I Page 24 e

Plant Change/Nodifitation No. 216-77 PSL UNIT ill "PRESSURIZER SPRAY VALVE PLUG. TO ST&f"CONNECTION REPAIR" This change added a retention weld at the plug to stem connection of the pressurizer spray valves. This was recommended by the valve vendor to alleviate problems with plug and stem separation caused by internal vibration in the valves.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment:important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The function and operating characteristics of the valves were not changed.

Qualified welders and procedures were used and the valve pressure boundary integrity was not affected.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

I

\

'V I

Page 25 Plant Change/Modification No. 217-77 PSL Unit 1 "INTAKE COOLING WATER STRAINER COVER GASKET CHANGE" This change allo~s the use of alterate materials for gaskets and washers on the Component Cooling Water Heat Exchanger 'salt water side inlet strainers.

Rubber gaskets were installed and mild steel full washers were used in place of "C" type stud washers. These changes were made to eliminate minor leakage problems.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The quality and function of the replacement materials are the same as the original components.

2~ The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety A'nalysis Report has not been created. This is a minor nonfunctional change.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

r ci

Page 26 Plant Change/Modification No. 218-77 Unit f/1 MOMENTARY PICKUP "C" AUX. FEEDWATER PUMP TRIP SOLENOID This change modifies the trip solenoid controls so the solenoid is picked up momentarily instead of continuously. This allows solenoid to relatch so,visual observation better informs operators pump is operable and clears a continuous annunciator window which blocks another needed alarm.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety in the Final Safety Analysis Report has not been previously'valuated

, increased.

No functional changes are made. Failure of the steam inlet valve controlled by this solenoid is already evaluated in the FSAR.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments in 1 above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 27 91ant Chango/Nodff 1oatdon No. 219" 77 PSL Unit 81 "RESIN DEMATERING PUMP PIPING MODIFICATION" The resin dewatering pump discharge piping was rerouted from the Holdup Tanks to Equipment Drain Tank lA. This change was made to prevent chemical contamination of the Holdup Tanks with waste resin sluice water.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident orr malfunction m of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increase d .

The new piping, valves and fittings used to implement this change are similar in quality and design to materials originally installed in the spent resin handling system. This modification does not affect safety related components.

2. The possibility for an accident or malfunction of a different tyypee than 'any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 28 Plant Change/Modification 223-77 Unit f/1 "CHARGING PUMP CONTROL RELAY MODIFICATION" This change added diodes to the charging pump pressurizer level control relay coils to prevent electronic noise which was giving interference with the Reac-tor Cooling System temperature indications.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evalu-ated in the Final Safety Analysis Report has not been increased.

There is no change to the design intent or function of the in-volved circuits.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3.'hefications margin of safety as defined in the basis has not been decreased.

for technical speci-This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 29 Plant Chango/Modifioation No. 224-77 PSL UNIT 81 "MODIFICATION TO RESTRAINT FOR BLOWDOHN VALVE FCV-23-6" The restraint for blowdown valve FCV-23-6 was modified to improve allowance for thermal growth by line I-2-B-2. This change was needed to prevent the possibility of subjecting the operator structure of the valve to binding.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Components used were designed and built in accordance with applichble codes and FSAR seismic design criteria. This change reduces the probability of a valve failure already analyzed in the FSAR.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. No functional changes are involved.

3.'he margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 30 Plant Change/Modification No. 226-77 PSL Unit 81

-"EQUIPMENT DRAIN PUMPS DISCHARGE PIPING MODIFICATION" Piping, valves and fittings were installed to connect lines 2-MH-S and 1$ -WM-B02 to provide the capability to process liquid waste from the 1A Equipment Drain Tank and 1A Chemical Drain Tank through waste ion exchangers and the waste filter while a liquid release from the 1B Equipment Drain Tank is in progress. These operations originally required use of the same piping and thus could not be conducted simultaneously.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Anilysis Report has not been increased.

No safety related components are involved in this change. All new materials used are compatible with the original design and quality group classification of the waste management'ystem.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

4 I Page 31 Plant Change/Modification No. 227-77 Unit //1 "FEEDWATER PUMP DISCHARGE VALVES CONTROL MODIFICATION"

These valves are required to close on both Safety Injection Actuation Signal (SIAS) and Main Steam Isolation Signal (MSIS) . It was noted (Refer to Licensee Event Report 335-77-5, dated February 25, 1977) that if a MSIS were. present with the feed pump running and no SIAS present the valves would close but then would reopen, close again, reopen, etc. This PC/M corrected that situation so the valves would not reopen on MSIS (or SIAS).

=

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an acci-dent or malfunction of equipment important to safety previously evaluated in the Final Safet'y Analysis Report has not been in-creased.

This change does not alter the valve functions; it gust corrects the as-built system to comply with original design intent. It should be noted that the probability of receiving MSIS without SIAS is low.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under 1 above.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

I Page 32 Plant Change/Modification No. 228-77 PSL Unit /31 "DELETION OF DROPPED CEA INITIATION OF TURBINE RUNBACK" The automatic feature of a turbine runback on the occasion of a full length dropped CEA was eliminated by disconnecting the runback circuit.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The runback circuit is not nuclear safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The dropped CEA analysis in FSAR Section 15.2.3 was made on the basis of, no turbine runback.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 33 Plant Change/Modification No. 229-77 PSL UNIT f/1 "REPLACEMENT OF VENTS AND DRAINS ON INTAKE COOLING WATER HEADERS" This PC/M allows the replacement of 1 inch vent and drain valves with caps or plugs on the 30-inch intake cooling water headers. This change was needed to promptly repair leaks that have occurred in the 1 inch vent or drain connections.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The vent and drain valves involved are not required to safely shut the plant down or to prevent or mitigate the consequences of accidents as evaluated in the FSAR. Replacement materials are consistent with original FSAR design criteria. This change does not affect the seismic design of the intake cooling water headers.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The failure of a cap or plug is not different than the failure of a 1 inch vent/drain valve.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 34 Plant Change/Modification No. 232-77 PSL UNIT //1 "MODIFICATIONS TO INTAKE COOLING WATER PKP" The following repairs were made to an intake cooling water pump. These changes were required because of corrosion and wear problems:

a. The pump shaft was repair welded and straightened.

b. Gutless rubber.,bearings were replaced with polypenco acetal bearings of similar design.

c. A small crack in the pump impeller was repaired.

d. Corrosion attacks in bearing support struts, casing welds, and the o-ring area of shaft tube were repaired using materials selected to increase corrosion resistance.

This change is not an unreviewed safety question because:

1. The probability of occurrrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The repair materials and methods were consistent with original equipment specifications and FSAR criteria for the intake cooling water pumps.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The changes do not affect the function, design bases, reliability, or single failure analysis of the intake cooling water system and components as evaluated in the FSAR.

3. The margin of the safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 35 Plant Change/Modification No. 236-77 PSL Unit 1 "RELOCATION OF RCP INSTRUMENTATION ON RTGB" The reactor coolant pump vibration reset switches and motor ammeters were relocated on the control board (RTGB 103). The original placement of these instruments was inconsistent with the location of other reactor coolant pump instruments and was a source of unnecessary confusion.. The control wiring schemes for these instruments were not changed.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The affected components are not nuclear safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. No functional change was made.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

\ I Page 36 Plant Change/Modif ication No. 242-77 PSL Unit !31 "REACTOR REGULATING SYSTEM CEA CONTROL MODIFICATION" The RRS was designed to automatically control RCS temperature by insertion/

withdrawal of CEA's as needed. Due to the vendors rec'ommendations to operate with all CEA's normally fully withdrawn and not to use CEA's for power increases it was decided to modify the system by removing the CEA automatic withdrawal function. Also, an audible indication was added to inform the operators of an automatic CEA insertion. This change in consistent with the vendor's fuel operating guidelines.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The Reactor Regulating System is not a nuclear safety related system.

This change provides a conservative approach to prevent exceeding the T-inlet Limiting Condition of Operation(LCO 542oF).

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 37 Page Plant Change/Modification No. 246-77 Unit //1 "CONTAINMENT SPRAY PUMP AMMETER REPLACEMENT" The containment spray pumps ammeter was changed from a 0-75 amp scale to a 0-100 amp scale to conform with original design spec-ifications. This change provided an increase in accuracy by ex-tending the range of pump motor amps indication.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an ac-cident or malfunction of equipment important to safety pre-viously evaluated in the Final Safety Analysis Report has not been increased. The replacement components conform to original specifications.

2. The possibility for an accident or malfunction of a dif-

'erent type than any evaluated previously in the Final Safety Analysis Report has not been created. The pump control system was not altered. This change only in-volved the range of an indicator.

3. The margin of safety as defined in the basis for Techni-cal Specifications has not been decreased.

This change does not represent a change to the facility as de-scribed in the Final Safety Analysis Report.

Cl Page 38 Plant Chango/Moddgicatton No. 251-77 PSL Unit 81 "RCP SEAL COOLER VALVES REMOTE OPEN SWITCHES" Remote open switches were installed in the, control circuit of the Reactor Coolant Pump seal cooler supply valves.. This change was made to avoid unnecessary entries to the containment building to re-open these valves and to reduce the time duration for loss of seal cooling water due to inadvertent valve closures.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This modification decreases the possibility of malfunction of a Reactor Coolant Pump by allowing timely remote resetting of the seal cooler supply valves. This control circuit change does not affect any automatic control features.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility's described in the Final Safety Analysis Report.

Page 39 Plant Change/Modification No. 258-77 PSL Unit //1 "CONTAINMENT BUILDING INSTRUMENT AIR SYSTEM MODIFICATIONS" Modifications were made to improve the reliability of the instrument air system inside containment. Changes include the addition of spring-loaded check valves in the compressor discharge lines, installation of a pressure regulator in the turbine building instrument air supply to containment, and the addition of a low pressure alarm.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Containment.

instrument air is not a safety related system.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. No changes in function of the instrument air system were made. These modifications improved the reliability of the system.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 40 Plant Change/Modification No. 259-77 PSL Unit 1 "INSTALL CURBING AROUND STATION SERVICE TRANSFORMERS" Six inch high concrete curbs were installed around the station service trans-formers to contain any possible leak or spill of askarel (PCB). This change was made because of recent concerns about the adverse affects on the environ-ment by the askarel (PCB's) .

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This is a non-nuclear safety related change.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. No nuclear safety related equipment is affected by this change.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as decribed in the Final Safety Analysis Report.

Page 41 Page Plant Change/Modification No. 260-77 Unit //1 "INSTALLATION OF POWER SUPPLY TEST JACKS FOR NUCLEAR INSTRUMENTATION DRAWERS" 15-volt power supply test jacks were installed in the linear range and wide range nuclear instrumentation drawers. This change was made to provide easier access for monitoring and maintaining equipment and to eliminate a potential personnel safety hazard.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Materials used to implement this change are consistent with the original vendor supplied equipment.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This modification does not affect the function or reliability of the Nuclear Instrumentation System.

3. The margin of safety as defined in the basis for Tech-nical Specifications has not been decreased.

This change does not represent a change to the facility as des-cribed in the Final Safety Analysis Report.

Page 42 Page Plant Change/Modification No. 264-77 Unit /ll "MODIFICATION OF A'DfOSPHERIC D&IP VALVES AND CONDENSATE STORAGE= TANK" During an engineering design review, an error in the sizing of the atmospheric steam dump valves was discovered (Reference Reportable Occurrence 335-77-22). This could have caused problems in plant cooldown to the shutdown cooling window in the unlikely event of loss of offsite power for an extended time period. The following modifications were made to cor-rect the deficiency:

a ~ The valve internals for I-HCV-08-2 A 6 B were changed from 6x3 to 6x4 to provide increased flow capacity. New valve actuators, silencers, and modified restraints were installed to ac-commodate the increased flow characteristics.

b. The Seismic Class I capacity of the Condensate Storage Tank was increased to 160,000 gallons by relocating a piping nozzle.

c. Digital indicators for RCS temperature were

• added in the Control Room to provide increased accuracy for transition to shutdown cooling.

d. The shutdown cooling piping support designs A*

were re-evaluated for a proposed increase in shutdown cooling entry temperature. Several minor support modifications were made. Some additional piping support changes are expected to be made in 1978.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Replacement components are purchased and installed to the same or better spec-ifications as the original components. New materials are compatible with the as-built systems. Present FSAR analyses envelope the consequences of any accident or malfunction related to this modification.

Page 43 Page Plant Change/Modification No. 264-77 (Cont)

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

There is no change in function of any safety re-lated system. Replacement components are of the same generic design as the originals.

3. The margin of safety as defined in the basis for Technical Specifications has not been decreased.

Design bases are met with changes to shutdown cooling entry temperature. (Technical Specifica-tion revisions are under consideration at this time.

• The digital temperature indicators are not yet operable due to material (RTD's not yet available).

The RTD's will be installed and the indicators made operable during the April 1978 refueling.

The remainder of the restraints will be adjusted/

modified during the 1978 refueling. A request for Technical Specification change to allow operation at higher temperatures has been submitted but not yet approved by the NRC.

The system will not be operated at the higher temperature until the items above are completed and the Technical Specification change is approved by the NRC.

l Page 44 Plant Change/Hodlfioatdon No. 268-77 PSL UNIT 81 "PRESSURIZER SPRAY VALVE INTERNALS MODIFICATION" Redesigned pressurizer spray valve plugs were provided to correct lateral vibration problems experienced when operating the valves partially open. The design of the new valve plugs redistributes the forces of the process the valve plug such that the net force on the plug at low lifts will be fluid'n directed in one direction and thus eliminate the vibration.

This change is not an unreviewed safety question because:

1. 'he probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The new valve plugs were provided by the valve vendor and were built to equivalent specifications as the original parts.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The redesigned plug has equivalent capacity and characteristics as the original plug.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

The change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 45 Plant Change/Modification No. 274-77 PSL Unit 81 "SECURITY LIGHTING STANDARD MODIFICATION" Installation of a Unit 2 construction tower crane deadman required removal of a Unit 1 lighting pole foundation and re-routing of conduit for lighting. It was determined by measurement that this light is not required for security illumination.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the. Final Safety Analysis Report has not been increased. This change does not affect nuclear safety related systems.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 46 Plant Change/Modification No. 275-77 PSL Unit /11 "RTGB ANNUNCIATOR FLASH RATE CHANGE" A resistor was changed in the Rochester Instrument Systems DC Momentary Alarm and Flasher Module to increase the flash rate. This was done to improve operator awareness of the alarming window.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The annunciator system is not nuclear safety related. The replacement resistor is of equivalent quality as the original part. The increased flash rate will make this system more effective to the operators.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. No changes in function were made.

3. The margin of safety as defined in the basis for technical specifications has not been decreased. The affected components are not discussed in the technical specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 47 Plant Change/Modification 280-77 Unit 81 "RCS TORNADO PROTECTED MAKEUP WATER SOURCE" Installed piping and valves between the Safety Injection Tanks drain line and the Volume Control Tank to allow the use of safety injection tank borated water inventory as an emergency shrink makeup water source to the Reactor Coolant System. This modification resolves condition I.4 of the St. Lucie Unit 1 Operating License.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety prev'iously evaluated in the Final Safety Analysis Report has not been increased. This change does not affect actuation or functional performance of plant safety related systems. The new piping and valves are located in the auxiliary building, which is designed for tornado missiles. Material selection is consistent with the design of the interfaced systems and with FSAR quality group classifications.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. All materials added by this modification are passive low pressure and low temperature components.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 48 Plant Change/Modification No. 281-77 PSL Unit 1 "ANNUNCIATE INVALID DISCHARGE CANAL LEVEL AND TEMPERATURE ALARMS" Circulating Water System discharge canal level and temperature indication circuitry was revised to provide alarms in the control room upon loss of data. This change was made to alert the operators that the data channel from the discharge canal is inoperative.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. No safety related components are affected.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 49 Plant Change/Modification No. 285-77 PSL Unit //1 "MODIFICATION OF CONTROL ROOM DOORS" Five control room doors were modified to improve the operation of security system card reader controls and to improve leak tightness. The direction of swing was reversed for four doors and magnetic type seals were installed for three doors.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This is not a nuclear safety related change. The leak tightness of the doors was improved.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This is a non-functional change made to improve the reliability and operation of the doors.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 50 Page Plant Change/Modification No. 286-77 Vnit 1/1 "ANNUNCIATION OF EMERGENCY COOLING CANAL VALVE POSITION" The annunciator scheme for the isolation valves in the flow barrier between the intake structure and Big Mud Creek was modified to pro-vide control room annunciation when the valves start to open. Orig-inally annunciation occurred only when the valves reached the full open position.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an ac-cident or malfunction of equipment important to safety pre-viously evaluated in the Final Safety Analysis Report has not been increased. This is a minor wiring change for an-nunciation of valve position only.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analy-sis Report has not been created. This change does not affect valve control scheme, characteristics, or function.

3. The margin of safety as defined in the basis for Technical Specifications has not been decreased. Annunciation when a valve starts to open will assist in meeting the Environ-mental Technical Specification limits on flow through these valves.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

t I Page 51 Plant Change/Modification No. 293-77 Unit  !/1 "TEMPORARY TEST CONNECTIONS FOR CHARGING SYSTEM" Pressure taps were installed at the suction and discharge of each charging pump and at the common pump discharge line to facilitate testing. These test connections were used to monitor and record pressure pulsations in the charging system, and were removed fol-lowing completion of testing. (See Page 59 of this report for a summary of the test procedure).

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or mal'function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Although these test connections were temporary, materials used (valves, fittings, etc) were consistent with the design and quality of the charg-ing system.

2. The possibility for an accident or malfunction of a dif-ferent type than any evaluated previously in the Final Safety Analysis Report has not been created. Charging pump controls and permanent indication and alarms were not affected by this temporary change.

3. The margin of safety as defined in the basis for Tech-nical Specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 52 Page Plant Change/Modification No. 299-77 Unit 81 "INTERIM RCS OVERPRESSURE MITIGATING SYSTEM" The pressurizer power operated relief valves control scheme was revised by installation of a variable low pressure setpoint. This modification is designed to mitigate pressure transients in the Reactor Coolant System during plant startups and shutdowns. This installation is considered to be an interim fix until NRC approval is received and verification of the water relief capability of the power operated relief valves'is complete. A description of this interim solution was forwarded to the NRC in our letter of August 23, 1977 (L-77-257).

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analy-sis Report has not been increased. Electrical inter-locks and administrative controls are provided to prevent an inadvertent PORV actuation signal. The mechanical characteristics of these valves are not changed.

2~ The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The equipment added by this modification is not nuc-lear safety related. Where a safety related circuit is interfaced, suitable isolation is provided.

3. The margin of safety as defined in the basis for Tech-nical Specifications has not been decreased.

Page 53 Plant Change/Modification No. 301-77 PSL Unit /31 "REPLACEMENT OF INTAKE STRUCTURE WARNING LIGHT" The intake structure warning signal and supporting components were removed and a new lighted marker buoy was installed. This change was approved by the United States Coast Guard.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction. of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change is not nuclear safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

A Page 54 Page Plant Change/Modification No. 302-77 Unit Pl "GOULD INVERTER CAPACITOR JUMPER WIRING MODIFICATION" The commutating and power factor jumper wires were revised to sep-arate the parallel connected capacitor banks into smaller parallel loops in order to reduce the maximum currents in each jumper wire.

These wires had shown signs of premature aging. The 10 KVA and 15 KVA inverters were affected.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an ac-cident or malfunction of equipment important to safety pre-viously evaluated in the Final Safety Analysis Report has not been increased. New materials were specified as equal to or better than those originally installed. Wiring was sized adequately to increase component reliability.

2. The possibility for an accident or malfunction of a dif-ferent type than any evaluated previously in the Final Safety Analysis Report has not been created. The func-tion of the inverters was not altered.

3. The margin of safety as defined in the basis for Techni-cal Specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 55 Plant Change/Modification No. 308-77 PSL UNIT /31 "MAIN STEAM CHECK VALVES BACKSTOP MODIFICATION" This change added a valve disc backstop to the main steam check valve cover plate to relieve loading on the original stop. Examination of the original backstop area revealed excessive wear. This modification improved the mechanical advantage and increased the backstop surface area to prevent deterioration of the valve internals.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety. Analysis Report has not been increased.

This modification is consistent with the original valve design criteria, and does not alter the function or operating characteristics of the main steam check valves. Calculated loadings on the cover and bolts are within allowable Ximits of the applicable code.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The new stop plate is a non pressure boundary apertenance welded to a pressure boundary part. Welding procedures and materials conform to applicable codes and specifications.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 56 Plant Change/Hodification No. 309-77 PSL UNIT 81 "INSTALLATION OF PIPING RESTRAINT FOR SAFETY INJECTION LINE I-24-SI-506" As a result of our program for inspecting concrete anchor bolts used in pipe hanger installations, it was discovered that support number SIH-57 was not installed in conformance with the as-built design drawings. This PC/M provided for installation of that support. (Reference FPL letter to NRC, f/L-77-312 dated October 7, 1977).

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change restored line I-24-SI-506 to the original design configuration.

The support materials, location, and design loading, satisfy the FSAR design criteria.

2. The possiblity for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been .

created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 57 Plant Change/Modification No. 311-77 PSL Unit 1 "REACTOR CAVlTY SUMP LEVEL ENDECATION MODlFXCATION" The reactor cavity sump liquid level transmitter was changed from a float type to a bubbler type instrument. Also, the instrument was relocated to outside the secondary shield wall. These changes will facilitate maintenance and improve the reliability of the level indication (LT-07-06).

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The instrumentation affected in not nuclear safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created. The change does not affect the function or quality of this level measuring instrument.

3. The margin of safety as defined in the basis for technical specifications has not been decreased. This modification improves the reliability of the reactor cavity sump level monitoring system.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 58 PROCEDURE CHANGES Januar 1 1977 December 31 1977 The following summarizes changes to procedures as listed in the FSAR in accordance with Title 10, Code of Federal Regu-lations, Part 50.59.

EMERGENCY AND OFF NORMAL PROCEDURE NO. 0030142 RCS COOLDOVN DURING BLACKOUT This procedure provided instructions to ensure the capability of reaching cold shutdown in the event of loss of offsite power considering the design deficiency of the installed atmospheric steam dump valves as reported in LER 335-77-22 dated April 22, 1977. Instructions were provided to bypass the low vacuum inter-lock and to remove a flange in the secondary steam dump system for additional steam dumping capacity. Also, this procedure pointed out that other sources of condensate storage tank make-up water are useable if necessary (city water transferred by fire pumps and hose). Permanent corrective actions for the atmospheric steam dump design error are summarized in the de-sign change section of this report (See PC/M /f264-77). This new procedure does not impact the safety analyses of the FSAR or the Technical Specifications and was determined not to in-volve an unreviewed safety question.

Page 59 TESTS The following list summarizes special tests performed under the provision of Title 10, Code of Federal Regulations, Part 50.59 during the period January 1, 1977 through December 31, 1977.

Char in S stem Pulsation and Vibration Testin Pressure pulsation, strain and vibration data were recorded at different RCS pressures and charging pump operating configurations (1,2 and 3 pump operation) to provide an extensive data base for determining the causes of charging system pulsation problems.

Damage to instruments had been experienced during low pressure operating modes.

~

This test. was performed on September 24 .and September 25, 1977 prior to and during plant cooldown for a scheduled maintenance outage. Southwest Research Instiutte (SWRI), a consultant organization assisted in the collection and evaluation of data.

Recommendations to resolve the apparent problems will be considered following receipt of a full report from SWRI.

The charging system was operated within design limits and in accordance with approved procedures at all times. This test did not constitute an unreviewed safety question.

Concrete Ex ansion Anchor Bolt Verification At the request of the NRC'a program for inspecting anchor bolts and concrete expansion anchors utilized in seismic class I pipe hanger installations was completed. The inspection procedure included checks for anchor bolt size and engagement as well as verification for proper installation. (Letter of Instruction T-03, revision 1). The results of this program were summarized" in our letter to the NRC, L-77-312, dated October 7, 1977.

This inspection program did not involve an unreviewed safety question.

RDT Sensor Res onse and Process Noise Data Testin Signal measurements were recorded for control channel temperature instruments to obtain steady state and transient data. on RTD signals to develop a technique for on-line verification of RTD sensor time response. Also, the test was to obtain steady state data on process input signals, such as feed the core protection calculator, to provide the NSSS vendor with typical noise levels.

This was done to support RTD response time testing as described in the FSAR and required by Technical Specificatiohs.

Test cables were attached to selected points in the C-E patch panel. Data was collected on May 10, 11 and 19, 1977.

The plant was operated within design limits and in accordance with approved procedures at all times during this test. This test did not involve an unreviewed safety question.

Page 60 CORE BARREL MOVEMENT Section 4.4.11.3 of the PSL 81 Technical Specifications requires the results of all periodic Amplitude Probability Distribution (APD) and Spectral Analysis (SA) monitoring to be included in this report.

In March,'977 the baseline measurements were completed and a full report of levels observed was submitted to the Commission in April, 1977.

This report also detailed the determination of alert and action setpoints pursuant to specification 4.4.11.1.

During routine monitoring on June 21, 1977 the RMS levels of noise on four of the excore detector channels exceeded their baseline values by 10%. A special report describing measured levels and identifying likely causes for the increases was submitted pursuant to specification 3.4.llc.

As anticipated, the RMS levels of all detector signals have continued to increase on a gradual trend throughout 1977. This trend has been observed on both APD and SA monitoring and represents an increase of about 30% over the baseline levels. The greatest portion of this in-crease remains in the lower frequency band (1 4 Hz) (normally observed due to changes in coolant temperature coefficient of reactivity and fuel element vibration coefficient of reactivity).

The observed increases in RMS values, if attributed entirely to core barrel motion, would indicate an increase of less than .001" RMS motion at the snubber level.

Page 61 STEAM GENERATOR TUBE INSPECTIONS Section 4.4.5.5.b of the PSL 81 Technical Specifications requires reporting all Steam Generator Tube Inspections in this report.

For the period January 1, 1977 throu'gh December 31, 1977 no tube E

inspections were performed.

It is planned that tube inspections, as specified by Section 4.4.5.3.a of the Technical Specifications, will be performed during our first refueling in 1978 and reported in the Annual Operating Report for that year.

I  %, <<s k ),,vl el k*Aee s 4 As s<<4 s',R %st lk <<ass ~ s,sash e es 0

~ s s ss> ~ s NUMBER OF PERSONNEL ( >100 mrem} TOTAL ~-REM STATION UTILITY CONTRACT STATION UTILITY CONTRACT OYEES EHPLOYEES WORKERS 6 EMPLOYEES EMPLOYEES WORKERS

. EMPI WORK 6 JOB FUNCTION Reactor 0 erations & Surveillance:

Maintenance Personnel 0 0 0 0.00 0.00 0.00 Operating Personnel 32 0 0 7.90 0.00 0.00 Health Physics Personnel . 14 0 0 4.51 0.00 0.00 Supervisory Personnel '1 0 2 0.15 0.00 0.21 Engineering Personnel 2 0 0 .0.26 0.00 0.00 Routine Maintenance:

Maintenance Personnel 78 0 67 43. 69 0.00 12.53 Operating Personnel 2 0 0 0.33 0.00 0.00 Health Physics Personnel ll .2 1

1 6

5.57 0.15 0.22 0.11 0.13 0.79 Supervisory Personnel 1 Engineering Personnel 0 0 1 0.00 0.00 0.14 Inservice Ins ection 6 S ecial Maintenance:

Maintenance Personnel 7 0 21 3.35 0.00 5.94 0 0 15.63 0.00 0.00 M Operating Personnel 27 0.00 O Health Physics Personnel 8 0 0 6.47 0.00 Supervisory Personnel 2 1 3 0.71 0.11 0.41 Engineering Personnel 2 0 3 1.24. 0.00 1.65 Maintenance Personnel 36 0. 0 10.55 0.00 0.00 Operating Personnel 18 0 0 2.91 0.00 0.00 Health Physics Personnel . 17 0 0 4.91 0.00 0.00 Supervisory Personnel 0 0 0 0.00 0.00 0.00 Engineering Personnel 0 0 0 0.00 0.00 0.00

~Refuelin 0 0 0.00 0.00 0.00 Maintenance Personnel 0 0.00 Operating Personnel 0 0 0 0.00 0.00 Health Physics Personnel 0 0 0 0.00 0.00 0.00 Supervisory Personnel 0 0 0 0.00 0.00 0.00 Engineering Personnel 0 ~ 0 0 0.00 0.00 0.00 TOTAL:

Reinrenenee Personnel 80 0 84 57.59 0.00 18.26 Operating Personnel 37 0 0 26.79 0.00 0.00 Health Physics Personnel 21 2 1 21.46 0.22 0.13.

Supervisory Personnel 4 2 10 1.01 0.22 1.41 Engineering Personnel 4 0 3 1.50;, 0.00 1.79 146 98 108.35 ~

0.44 21.59 GRAND TOTAL

'lt ~ ~

Cg 0

Page --"

ABBREVIATIONS USED A/C Air Conditioner Air Operated Valve B.A. Boric Acid CCP Coolant Charging Pump CCW Component Cooling Water (for Rx plant components)

Channel (i.e. one of four channels of the RPS)

CVCS Coolant and Volume Control System (Charging and letdown)

Control Wiring Diagram Disch Discharge Flow Control Valve

'eedwater Feedwater'ump Header-HPSI High Pressure Safety Infection Heat exchanger ICW Intake Cooling Water (sea water cooling for CCW, Turbine Cooling Water)

ISO or ISOL Isolation (valve)

Ion exchanger (demineralizer)

Level Control Valve LPSI Low Pressure Safety Injection MOV or MV Motor Operated Valve MS IV Main Steam Isolation Valve NI Nuclear Instrumentation PCV Pressure Control. Valve

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ABBREVIATIONS (cont)

, PRZR or PZR Pressurizer

'RCP Reactor Cooling Pump RV Relief Valve Reactor

'hutdown Cooling (decay heat removal system)

S/G or S.G. Steam Generator SIT or Sl Tank Safety Injection Tank (Accumulator) m/LP Thermal Margin-Low Pressure TK Transmitter VCT Volume Control Tank V/I Voltage to Current (signal) converter

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1976 ANViJAL OPERATING REPORT FLORIDA POWER AND LIGHT COMPAiVZ ST. LUCIE UNIT 81 qg ~ ~gb gC February, 1977

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INDEX TITLE PAGE Narrative Summary Outage Summary, Including Safety Related Maintenance

(

Changes 24 I

Design Changes 25 Procedure Changes 138 Tests 139 Failed Fuel Indications 140 Core Barrel Movement 141 S/G Tube Inspections 142 Radiation Exposure 143 Containment Penetration Leak Rate Test Results 144 Abbreviation List 147

Page 1

SUMMARY

OF OPERATING EXPERIENCE The following is a summary of Plant Operations including pertinent items

-of interest chronologically for the period 3-1-76 '(operating license issu 6) to 12-31-76. The plant did not reach 100X power during this period.

3-1-76 1) Full security plan implemented

2) Commenced preparations for initial core load

3) Operating incense issu e d, effective 3-1-76

'-2-76 1) Dummy w"th neutron source loaded in Z-ll for instxument response check 3-3-76 1) 2:45 A.H. core load commenced 3-4-76 1) Fuel loading secured (11:45 P.H. 3-3-76 2:37 A.H. 3-4-76)

Containment integrity was breached due to decxeasing level in transfer canal canal refilled to proper level. Reported per LZR 335-76-1, dated April 2, 1976.

3-5-76 1) Surveillance check of Sp'ent Fuel Hachine performed 6:35 A.H.

, Ovex'load did not function properly fuel loading secured.

Reported via LER 335-76-2, dated April 5, 1976

2) 1:30 P.M. - NRC issued (pex telephone) Amendment 81 to license DPR-67 allowing operation without the overload for a period of two weeks (for 'non-irradiated fuel).

3) 3:30 P.H. - Fuel loading recommenced 3-11-76 2:26 P.M. Last fuel assembly loaded into core 3-12-76 1) 5:50 A.H. Upper guide structure installed

2) 11:45 P.M. Completed latching CEA's 3-14-76 12:22 A.M. Vessel head in place 3-15-76 6:45 A.M. Entered Mode 5, Vessel head tor'qued 3-16-76 Commenced filland vent and heat-up preparations 3-20-76 4:40 P.H. CEDM cable connections completed 3-21-76 1) Fill and vent - 3:50 A.M. water issued fxom Pzr vent-vent closed, increased RCS to 200 psi.

2) 10:59 P.H. - Commenced 30 sec. pump runs (RCP) for venting Reactor Coolant System (RCS).

3-23-76 3:13 P.M. RCP runs'omplete for filland vent.

3-25-76 4:30 P.H. - Commenced heating Pzr to draw bubble

Page 2

'-26-76 1) 2:35 A.M. Started 1B1 and 1B2 RCP's for heat-up

2) 5:45 A.M. - RCS 0 at 2.8 ppm 8 200 F - stopped 1B2 RCP, began reducing 02 conc.

3) 4:15 P.M. - Commenced Pre<<Op 0110081 CEDH/CEA Testing

4) 10:11 P.M. 02 8 .05 ppm heat-up recommenced.

3-30-76 1:00 A.M. - 1.23 GPM Leak rate from RCS found to be RCP Bleedoff relief - repaired and returned to service 11:30 A.M.

4-1-76 1) Entered Mode 4

2) 4,'30 P.H. - Declared CEA 44 inoperable during preoperational

., test

3) 11:00 P.H. Drew vacuum 9 26" in Turbine condenser 4-2-76 Continued CEDM testing 4-3-76 6:00 A.H. Entered Mode 3 RCS 8 300oF

/

4-4-76 1) 6:00 A.M. Commenced heat-up to 470oF, 1800 psia

2) Performed OP 0110081B, CEDM/CEA 44 Lower Gripper check

3) Completed filling fuel pool 4-5" 76 1) 1:15 A.H. Commenced Heat-up to 510 F and 2150 psia

2) 9:15 A.M. CEA. 844 declared operable "3) 2:45 P.M. Started 4th RCP Commenced heat-up to 532 F - 2250 psia 4-6-76 Pre-Op 0110081,.CEDM Drop Testing in progress at 532oF, 2250 psia.

4-10-76 7:00 A.H. CEA Drop testing per Pre-Op 0110081 complete 4-11-76 ll:23 A.H. Conducted RCS Flow Coastdown Test 4-12-76 9:00 A.M. - OP 0110081 In progress Fuse Block on lA2 CUP Breaker caused discharge valve on lA2 CWP to close with pump operating Rapid action by NPS (opening breaker locally - manually would not open from control center) .

avoided damage to lA2 CWP.

4-13-76 1) 12:35 P.H. Secured all RCP's for no-flow CEA testing

2) 8:32 P.M. Started 3 RCP's

3) ll:30 P.H. Commenced cooldown to Mode 4 for testing CEA 44.

4-14-76 1) 5:00 A.H. - Entered Mode 4

2) 11:30 A.H. CEA 44 Inoperable

3) 4:00 P.M. Heat-up to 420 F to Mode 3

4) 9:30 P.M. - CEA 44 operated 8 420 F 1260 psia 4-15-76 1) 12:55 A.M. Entered Mode 4 RCS 280oF, 470 psia

2) 4:05 A.M. Retested CEA 44, Test Satisfactory

3) 4:45 P.M. Heat-up, entered Mode 3

4) 11:36 P.H. - 4 RCP's in service 4-17-76 Conducted S/G Feedwater Hammer Testing 4

Page 3 4-18-76 Normal Plant Ops (NPO) - making preparations for initial criticality 'I 4-20-76 7:00 P.M. - Commenced CEA testing IAW App 3, Addendum I of Initial Criticality Procedure 4-21-76 Continued Rod Testing

1) 7:00 P M. - Completed App. 3 of Initial Crit. Procedure.

2) CEDM 44 would not operate properly cold but did function-hot., Amendment 4 to license DPR-67 has been issued, deleting approval to go critical cold for testing, thus allowing hot operations and criticality with CEDH 44 to be repaired/replaced at a later date.

3) 7:56 P.M. Began diluting RCS IAW Initial Crit. Procedure 4-22-76 1) 7:10 A.M. - Entered Mode 2

2) 8'30 A,M St Lucie /Il Critical, Boron Cene., 935 ppm

3) 3:40 P.M. Commenced Low Power Physics Testing 4-26-76 1) 6:00 A.M. - Rx subcritical for CEA drop test

2) 7:43 A.M. Rx critical

3) 10:51 A.M. Use of part length CEA's to control power Inserted belo~ 90% withdrawn limit. Reported in LER 335-76-18, dated May 26, 1-9-76

4) 3:10 P.M. Reactor shutdown In Mode 3 for repair of CEDM Reed Switches 4-27-76 1) 11:15 Rx critical - Entered Mode 2

2) 8:27 P.M. Recommenced Low Power Physics Testing 4-30-76 1) 2:55 P.H. Completed Low Power Physics

2) 3:30 P.M. - Rx shutdown, Mode 3 5-2-76 ~

Attempted to withdraw CEA's Timers bad on several manual trip for .repair.

5-3<<76 11:37 P.M. Reactor critical 5-4-76 1) 5:35 A.M. Rx Trip - Operator error while testing Nuclea2 Instrumentation

2) 9:32 P.M. Rx Critical 5-5-76 1) 1:30 A.M. Rx S/D to repair CCW Line to RCP motor

2) 7:51 A.M. Rx Critical

3) 10:42 A.M. - Commenced power ascension to 5%

5-7-75 1) 4:00 A.M. Rx Trip while conducting Steam Bypass Control System (SBCS) Test due to Low S/G level

2) 6:00 A.M. - Rx critical Note: The FW Control

3) 6:24 A.H. Rx Trip (Low S/G level) System was later adjusted

4) 7:05 A.M. Rx Critical (at power) to minimize

5) 12:24 P.H. Rx Trip (Low S/G Level) recurrence of these

6) 1:55 P.M. Rx Critical problems.

7) 8:48 P.H. PSL Unit 1 on Line 8 100 5-8-76 12:15 P.H. Rx Trip (Low S/G Level) while performing Turbine overspeed tests

• Indicates forced power reduction of 20% or more per Reg. Guide 1.16

Page 4 5-8-76 2) 1:40 P.M. - Commenced Rx Start-up

3) 1:44 P.M. - Dropped CEA 32 due to blown SCR, replaced SCR. (See Note 2)

4) 5:18 P.M. - Rx Critical 5-9-76 2:18 A.M. - Performed turbine overspeed test - Sat 8 1971 RPM-Unit on Line 8 100 %f 5-10-76 1) 3:51 P.M. - Manually tripped unit due to closure of 1B MSIV - cause DC ground

2) 7:04 P.M. Rx Critical

3) ll:05 P.M. - Unit on line 5-12-76 1) 2:55 P,M. Conducted 20% Trip test

2) 8:40 .P.M. - Rx Critical 5-13-76 1) 2:27 A.M. - Unit 1 on line 8 160 E4 (20%)

2) 6:00 P.M. Power ascension to 30%

3) 7:30 P.M. 8 30%

5'-09 A.M. Rx Trip due to loss of Feedwater (feed pump tripped), performed S/G Water Hammer Test (See Note 1)

2) 3:ll P.M. - Rx Critical

3) ll:30 P.M. Unit on line Ascension to 50/ plateau commenced 5-15-76 1) 9:30 A.M. 40/ plateau

• 2) 8:00 P.M. - Problems with water hammer in feedwater heaters while placing moisture separator/reheaters (MSR's) in service Unit manually shut down from 50% Entered Mode 3 for addition of restraints to secondary plant (60 hr. S/D) 5-16-76 Start-Ups performed by hot license candidates for training 5-18-76 9:10 Rx Critical - performed training start-ups 5-19-76 6:10 A.M. Secondary plant restraint addition complete, Unit 1 on line and going to 40/

5-20-76 Ascension to 50/ power 5-21-76 *5:23 P.M. Rx Trip - Low S/G Level, 1A i~fP Tripped, lB MFWP did not start. (See Note 1) 5-22-76 1) 4:20 A.M. Rx critical

2) 8:25 A.M. Rx Trip Low S/G Level During turbine startup

3) 10:22 A.M. Rx Critical

4) 1:41 P.M. - Unit on line, increasing power to 50%

5-23-76 1) 4:00 P.M. Lightning strike in switchyard caused 4 loss of load pretrips 5-24-76 Normal Plant Operations (NPO) 8 50/

6-1-76. NPO 8 50%

• Indicates forced power reduction of 20/ or more per Reg. Guide 1.16

Page 5 6-6-76 NPO 8 50X Set up for Moderator Temperature Coefficient (MTC) and Power Coefficient Test

'-7-76 N) 9:40 A.M. - Manual Trip of Rx 9 50X due to dropped CEA 826 during MTC and Power Coefficient Test,'. Cause - power switcH on CEA - repaired. (See Note 2)

2) 5:24 P.M. - Rx Critical

3) 7:44 P.M. on the line 6-8-76 1:30 A.M. - Reached 50X plateau 6-12-76 4:56 P.M. - Tripped Reactor 8 50% - To perform Control Room Inaccessibility Test 6-13-76 Cooling'own RCS for a scheduled maintenance shutdown I 1) 8:00 A.M. Entered Mode 4

2) 12:52 P.M. On Shutdown Cooling (SCD)

3) 6:35 P.M. - Entered Mode 5 6-17-76 Fill and Vent of RCS

1) Shen first RCP was started, there was a pressure transient 0

to 300 psi above the pressure temperature limit (at 100 F) for less than 5 minutes. Reported in LER 335-76-30, dated 7-17-76.

2) 4:30 P.M. Commenced drawing Pzr Bubble 6-18-76 11:53 A.M. Entered Mode 3 6-19-76 1) 9:53 A.M. Rx Critical

2) 5:25 P.M. - On the line Commenced Power Ascension to 80%

6-20-76 1) 2:02 P.M. - Increasing Power to 60%

2) 6:40 P.M. 8 60%

4 6-21-76 9:10 A.M. Reached 70%

6-22-76 1) 9:27 A.M. Dropped CEA /r'20 Turbine load reduced to match Rx power 9 60/ - recovered CEA

2) 2:28 P.M. Reached78Z plateau

• 3) 5:00 P.M. Reduced load to 50% to clean Feedwater Pump (FWP) strainers 6-23-76 *ll:28 P.M. Rx trip 8 50Z - Cause low S/G level due to closure of 1 A MSIV MSIV closed due to opening of its DC Breaker. The breaker was opening while investigating cause of loss of AB DC Bus. AB,DC Bus was lost while trying to secure A Battery Charger which was oscillating.

6-24-76 1) 9:45 A.M. - Rx Critical

2) 12:18 P.M. - On line, increasing power to 70/

6-25-76 6:15 A.M. 8 70Z 6-26-76 *3:30 A.M; - Reduced load from 70X to 50X to clean 1B PWP strainer

>Indicates forced power reduction of 20% or more per Reg. Guide 1.16

Page 6 6-28-76 Increased power to 78%-

7-1-76 1) 1:35 A.M. Commenced load reduction from 78/ due to high usage of H 2 in Generator (Note leak found 6 repaired)

• 2) 2:55 A.M.. Reactor Trip (TM/LP) Thermal Margin/Low Pressure-awhile Borating to" r-duce power and tiansferring from single to seguential valve control, generator picked up 60 - 70 MV

- rapidly decreasing RCS pressure caused trip.

7-2-76 1) 3:05 A.M. Rx Critical

2) 8:10 A.M. - Unit on line

3) 8:33 A.M. - Rx Trip 9 10/ Low S/G level

4) ll:20 A.M. - Rx Critical

5) 1:ll P.M. Rx Trip - DEH" Malfunction 8 10%

6) 3:25 P.M. - Rx Critical 7-3-76 Increased power to 78/

7-4-76 3:23 A.M. - Turbine Runback 78% - 70% due to malfunction in Runback circuit - repaired 7-5-76 NPO 8 78%

7-6-7 6 *1) 7:20 P.M. Reduced Power from 78% to 50% to clean condensate pump strainer. At this time we first became aware of a possible flux distribution anomaly. See the write-up attached to PCM 176-76 in this report. Also reported in LER 335-76-35 dated July 23, 1976.

7-7-76 NPO 8 48%

7-9-76 NPO 8 48%

• 1') Reduced power at 1:15 A.M. to off line for testing. This was to help determine cause of the flux distribution anomaly.

2) 11:27 A.M. Generator off line

3) 8 10 2% power for physics testing 7-14-76 NPO g 10 2%

2:06 A.M. Rx Trip - Operator error during RPS Logic Matr'ix Test 7-15-76 1) 8:02 P.M. . Rx Critical 9 10 7-18-76 NPO 8 10 2%

1) 1:19 - Tripped Rx Borating to Refueling Concentration

2) 9:45 A.M. OQ SDC

3) 12:30 P.M. Entered Mode 5 7-26-76 Decision to remove some. fuel for inspection confirmed by Company management.

7-27-76 6:00 A.M. PZR solid

• Indicates forced power reduction of 20/ or more per Reg. Guide 1.16

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Page 7 8-1-76 Entered Mode 6 8:30 A.M.

8-3<<76 Uncoupling CEA's 8-4-76 Dual CEA problems - See PCN 203-76 8-5-76 ,Drilled the "7" Slots on Dual CEA sh'afts. All CEA's uncoupled.

8-8-76 Commenced removing fuel for inspection of fuel assemblies 8-18-76 Commenced defueling core 8-23-76 12:14 A.M. Core defueled 8-23-76 11-15-76 Fuel. Reconstitution - See PC/M's 176-76, 192-76, 200-76 and discussion attached to PC/M 176-76 11-4-76 . Commenced core loading Cycle lA 11-15-76 Fuel reconstitution complete 11-16-76 Core loading complete, started reassembly of Reactor vessel 11-25-76 Reassembly complete, entered Mode 5 11>>26-76 On Shutdown Cooling (SDC) 11-28-76 6:30 P.M. - RCS Solid 11-29-76 Started RCP Runs for filland vent 11-30-76 Fill and Vent

1) 8:15 A.M. Commenced drawing Bubble in PZR

2) 3:37 P.M. Entered Mode 4

3) 10:30 P.M. Entered Mode 3 12-2-76 Mode 3-CEA Testing including testing of CEDM 44, replaced dur'ing the fuel reconstitution shutdown.

12-3-76 RCS Flow Test and low flow trip setpoint verification 12-4-76 1) 2:10 A.M. Commenced diluting to critical

2) 3:37 P.M. Rx Critical 12-5-76 810  % CEA Symmetry Test, In progress 12-6-76 910"2X CEA Testing 12-7-76 910 I CEA Testing 12-8-76 810 ~X Isothermal Testing 12-9-76 10:25 PM - increased Power to 3/

Page 8 12-10-76 1) 5:58 A.M. Mode 1

2) 7:53 A.M. On the line at power, 20Z plateau 12-11<<76 1) 6:00 A.M. 9 30Z plateau 12-16-76 1) 3:50 P.M. 9 40Z plateau 12-17-76 1) 8:56 P.M. 9 50Z plateau 12-22-76 Still at 50Z. - Declared PSL 81 Commercial at 12:01 A.M.

12-27-76 NPO 0 50Z. Conducted Moderator Temperature Coefficient and Power Coefficient test 12-31-76 Continuing operations at 50Z power Note 1 We have experienced some difficulty with HFW pumps tripping while running and tripping off immediately after automatically starting when the running pump has tripped. This appears to be caused by signals from the feed pump protective trip circuits (low suction pressure and flow). We have modified the Feed Pump recirculation valve controls and are continu-ing to gather data and evaluate the problem so any further corrective action necessary can be defined and implemented.

Note 2 Shortly before licensing the CEDM power supply vendor and the NSSS vendor recommended a modification to a power supply module to improve reliability. This was approved and draw-ings issued before licensing. As modified modules became available, throughout the power ascension, the new modules were installed. Since return to operation in December with all modules replaced, there have been no dropped CEA's due to these modules. The modification primarily consisted of up-grading the voltage rating from 400 to 600 volts.

Page 9 OUTAGE

SUMMARY

MARCH 1, 1976 - DECEMBER 31, 1976 The following summarizes the three plant shutdowns performed for maintenance from initial criticality on April 22 through December 31, 1976.

5/15/76 5/19/76 This 60-hour outage was for installation of restraints on the second-ary plant due to vibration and water hammer experienced during the first attempt to place the Moisture Separator/Reheaters in operation.

The restraints prevented plant damage until sufficient operating experience allowed formulating and implementing permanent corrective action which has been satisfactorily completed. No major corrective safety related maintenance was performed.

6/S2/76 6/19/76 This 7-day shutdown was to clear up several problems, mostly in the secondary plant, in anticipation of power ascension to 80X and beyond.

Major safety related corrective maintenance consisted of: repairing the pressurizer spray valves (leaking through as discs had loosened from the stems); zeplacing the Auxiliary PWP crossconnect valve stems; repacking the SDC loop isolation valves, the power operated relief valves and 1B MSIV; replacing Cell f/31 in 1B battezy; and performing PC/M 116 to reduce the RPS temperature indication noise.

7/9/76 12/'0/76 This 155-day shutdown was for fuel poison pin replacement (fuel recon-stitution). See PC/M's 192-76 and 176-76 with attached discussion for details. Other major safety related items were: repair of 1C ICW pump bearings, repair of two small leaks in ICW lines and replacing the motor on V-2501, Volume Control Tank outlet isola-tion valve.

Pollowing is a summary of other Safety Related corrective maintenance.

MECNANICAL CORRECTIVE HAINTENANG ON ETY-RELATED EQUIPMENT E UIPHEWT PWO 0 HAI.PUNCTION CORRECTIVE ACTION 1 A CCM ICM Strainer 0187 Nigh AP Cleaned Fuel Pool Purif. V07188 0194 Bonnet leak Replaced gasket B Diesel Gen. Air Start Sys. Lubricator 0201 Low oil flow Ad)usted B Diesel l.ube Oil Pump 021.1 Oil leak Replaced seal B Charging Pump 0213 Seal Leak Replaced plungers & packing 1 B 1 Diesel Gen. l.ube Oil Pump 0214 Cracked brg. housing ~ Replaced housing A Diesel Gen. Air Comp. (Diesel) 0218 Ran out of fuel Prime fuel system A Diesel Gen. Alr Start Sys. Valve 0284 Faulty Air Relay valve Cleaned A CCM P Inboard Seal 0298 Seal leaks Replaced seal.

Control Rm NVA-3A, -3B, -3C 0313 Dust filters dirty Changed filters A Dics01 Can. Fuel Filters 0315 Nigh AP Changed filters A Charging Pump Dinch. RV2326 0320 Plug leaking Replace gasket B Cont. Spray Pump Suet. V7124 0321 Bonnet leak Replaced gasket R.A.B. SAPSDS Rm. Drain V25-6 0322 Packing leak Packing tightened IC Gas Decay Tank Disch V6703 0323 I.eaks past seat Replace diaphgram A Emcrg. Diesel Fuel Transfer Pump 0328 Packing leak Rcpackcd A CCN Nx. JCU Strainer 0333 Nigh AP Cleaned Przr. Spray Bypass V1236 0380 Packing leak Rcpacked Przr. Spray Bypass V1236 0381 Packing leak Repacked B I.C.M. Pump Piping 0432 Seal Water Line Plugged Cleaned B Diesel Gen. Cooling RV 0450 Leak Past seat Removed, reset, replaced B CCW Nx ICM Strainer 0456 Gasket leak Replaced gasket B CCM Pump Inbrd Brg. 0461 Oil leak Replaced gasket Fuel Pool Purif. Pump Suction V07170 0466 Body to bonnet leak Replaced gasket Letdown LCV-2110Q 0468 Body to bonnet leak Replaced gasket & seal A Cont. Sump. Check V07174 0479 Leaks past sear. Cleaned seats Cont. Escape Natch 0482 Various loose bolts in operating mechanism Tighten and stake threads B ICN Pump 0483 Packing leak Repackcd Aux. NPSI Ndr. to CVCS, V-2340 0486 Bonnet leak Replaced gasket ACC 3 A, B, C Control Room hir Cond's. 0490 Condenser cooling air dampers seized Freed and lubcd C ICH Pump 0492 Bearings failed Installed spare pump Pressurizer Vent V-1239 0493 . Valve leaks past seat Ncw plug & stem - lapped

lfEClfhHICAL CORRECTIVE MAIHTEHAN OH APETY-RELATED EQUIPMEHT E UIPlfEHT PRO 0 MALPDNCTIOli CORRECTIVE ACTIOH Containmcnt Personnel Door 3273 Hisalignment - hard lo operate Ad)usted Pressurirer Valves 1436 6 1440 3289 Bonnet leaks Tighten bonnet lA Aux. Feed Pump Valve HV-9 3290 Bonnet leak Replaced bonnet gasket RCP Bleed Off to Quench Tank R.V.2199 3293 Leaks past seat Remove,relap, retest, replace RV 3483 to llold Up Tank 3294 Flange leaks 'I Replaced gasket 1A Purif.I Ex. V2380, V2372 3465 Valves not closing Replaced diaphragm in air operator H2 to Gas Comp. V6059 3504 Body to bonnet leak Tighten bonnet RCP Controlled Bleed Off V6107 3508 Valve leaks thru Rclapped seat lA llastc Gas Comp. V6573 3522 Valve leaks thru Adjusted stem stop nuts IA Charging Pump Coupling 3523 Excessive Crease Removed excess grease 1C HPSI Inboard Seal 3524 Leaks Replaced Regcn llx V2810 3559 Valve leaks thru Rclapped seat 1B 6 C HPSI Pump 3570 Seal cooling lines luak Tightened unions A Loop S.D.C. Isolation V3480 3580 Packing leak Repacked Letdown Iso. Valve 2515 3582 Packing leak Tightened packing lA Shutdown Hx RV3431 3593 Leaks past seat Rcmove,rework, retest, replace Letdown llx RV2345 3595 Leaks past seat Rcmove, relap, retest, replace Gas Decay Tank V 6592 6701 3605 Stems pulled 'loose Replaced diaphragms C.V.C.S: PCV 2201 3606 Packing leak Repackcd valve B I.C.M. Pump 3609 Packing leak Ad)usted packing Aux. Feed Pump Cross Connect HV09-14 3610 Stem damaged during lfOV test Replaced stem 1 C Gas Decay Tank V6597 3614 Leaks past seat Replaced diaphragm C.V.C.S. Let Dwn I,.C.V. 2110P 3620 Packing leaks Repacked valve C.V.C.S. I.ct Dwn L.C.V. 2110 Q 3621 Packing leaks Rcpacked valve B Cas Decay Tank V6578 3635 Leaks past seat Replace diaphragm IB Charging Pump Vent V2805 3638 l.caks past seat Installed blank flange B.A. Sys, F.C.V. 2161 3643 Packing leak Tightened packing 1 A Chargfng Pump 3646 Plunger packing leak Replaced center plunger<<repacked 1 A Bh "fake-Up Pump 3683 Seal leak Replaced seal Przr. Spray V1100E 3685 Bonnet leak Replaced gasket 1 A Charging Pump 3699 Brass Chips in brg. housing Cleaned and inspected o.k.

1 B Emer. Diesel Coolant Tk. RV 3705 Leaks past seat Lapped and reset CVCS l.etdown LCV2110P 3720 Bonnet leak Replaced gasket 1 A llSIV Check llinge Pin 3730 Steam leak on hinge pin cover Repaired by Purmanite process

. B SG Blowdown Orifice Ping. 3I115 Steam leak Replaced flexitallic gaskets l ~

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~0 HECIIANICAL CORRECTIVE MAINTENANCE CN SAFETY-RELATED EQUIPHENT E UIPMEtlT PWO lI HALFUIICTION CORRECTIVE ACTION Flash Tank Cas Vent Trap 3820 Trap leaks past Cleaned internals 1 8 Aux. Feed Pump Pfping 3878 Pipe plug leak Tightened CVCS Letdown I.CV2110Q 3892 Steam leak Rcpacked CVCS Letdoun LCV2110P 4002 Bonnet and packing leak Replaced gasket & repacked CVCS I.etdoun PCV 2201P 4015 Bonnet leak Replaced gasket 1 C Gas Decay Tank V6597 Ai020 Diaphragm leak Replaced diaphragm Flash Tk. Cas Vent T6909 4030 Trap leaks past Replaced plug & seat Cont. Personnel llatch 4042 I.oose pin in operating mechanism Replaced pin 1 A Diesel 12 Cyl. Starter Valve 4070 Air start supp.=valve Cleaned HSIV Bypass I-MV-08-18 4082 Stem broken Repfaccd MSIV Bypass I-MV>>08-IA 4083 Stem broken Replaced Przr. Relief Iso. V1403 4092 Packing leak .Repacked & neu gasket Przr. Relief Iso. V1405 4i 093 Packing leak Rcpacked 18 MSIV IICV-08-18 4097 Steam leak Repacked Aux. Feed Pump Cross Connect MV-09-14 4098 Damaged stem during HOV test Replaced stem

~- Aux. Feed Pump Cross Connect HV-09-13 4099 Damaged stem during HOV test Replaced stem 1 8 Not Lcg S.D.C. V3651 4100 Packing leak Repackcd 1 8 llot Lcg S.D.C. V3652 4101 Packing leak Repackcd Stcam Flou Tx FT-08-18 Iso. V8135 4105 Packfng leak Rcpacked Stcam Flou Tx FT-08-18 Iso. V8136 4i 106 Packing leak Rcpackcd Steam Flow Tx FT-OS-18 Iso. V8137 4107 Packing leak Repacked 1 8 Charging Dfsch. R V2325 4125 Leaking plug Inspected & cleaned Stiam Flou Tx. FT-08-18 Root V8134 4126 Packing leak Rcpacked Przr Conilensarc Pot. Iso. V1437 4129 Steam Leak Tf ghtcned plug Gas Decay Tank V6592 4141 Stripped stem Replaced diaphragm C Cliarging Pump Seals 4145 Seals leaking Replaced plungers & packing A Charging Pump Seals 4161 Seals leaking Replaced plungers & packing 1 A Charging Pump Disch. R V2326 4186 Leaks past scat Removed, lapped, retcstcd, replaced Przr. Safety V1201 4187 I.eaks past seat Removed, relapped, retested, replaced .

A Loop S.D.C. V3470 4193 Packing leak Repackcd A Imop S.C.D. V34igl 4194 Packing leak Rcpacked 8 CCII/ICW V21250 4244 Broken weld Tlircaded (temp fix) See PWO 712 Control Room V<<nt Fan HVA 3G (Damper) 4250 Seized damper Freed and lubed 1 A CCW Pump Ind. Brg. 4260 Excessive'il usage Inspected, replaced seal - PWO 4621 1 A CCW Pump Orbrd Seal k 4261 Seal leak Replaced seal BA Hakeup Relief RV2141 4276 Defective stem R'eplaced stem BA Hakeup Relief RV2133 4277 Defective stem Replaced stem Przr. Spray Bypass V1236 4281 Packing leak Repacked

~0 MEC))ANICAL CORRECTIVE MAINTENANCE ON SAFETY RELATED EQUIPMENT E UIPCfENT PWO 0 HALF):NCTION CORRECTIVE ACTION Przr. Spray Bypass V1237 4282 Packing leak Repacked B.A. Make-Up Sample Sys. A V2128 4407 Seat leak Cleaned - o.k.

1 A HSIV 44i09 Steam ).eak at plug Repackcd 1 B Charging Pump Packing 4413 Packing leak Replaced plungcrs & packing 1B1 SI Tank V3631 4i422 Body to bonnet leak Rccori)ucd bonnet bolts Sample llc. Ex. Iso V1211 4426 Body to bonnet leak Replaced bonnet gasket ACC. 3A, 8, C Control Room h/C's 4600 Conclcnscr cooling air dampers seized Freed-up 6 lubed 1 A Charging Pump Disch. R V2315 4602 Leaks past seat Re>>loved, rclappcd, reset, replaced V.C.T. Piping Casket 4647 Bad gasket - gas leak Replaced gasket V.C.T. Piping Casket 4649 Bad gaskec-gas leak Rcplaccil gasket 1 h G.D.T. Inlet V6584 4i 650 Leaks past seat Replaced diaphragm V.C.T. Inlet Chk. V2112 0002 Bonnet leak ~

Replaced gaskcc C I.C.W. Pump 0003 Packing leak hd]usted packing lB ICW Pump Lube Water Str. Valve 0004 1.caked thru badly Replaced valve

)lVS-1C Cont. Cooling Fan 0007 Bad bearing Replaced Przr. Sliray 1100E 0009 Leaks thru seat Checked o.k.

ccw llx. salt. water str. v21339,21334 0010 Not closing properly Repaired << ad)usted A CCW llx Salt Water Str 0011 Strainer plugged Clean Cont. Pcr llatch Outcr 0012 Casket leaks Replaced 1 A I.PSI Suction R V34i83 0018 Leaks past scilt Removed, rctested, replaced 1 B LPSI Suction R V3468 0019 Leaks pose scac Clachlnc and lap seat Przr. Quench Tank Rupture Disc. 0024 Damaged Replaced Diesel Cen. Lube Oil Pump lh2 0027 Noisy bearing Realigned Dlcsc!1 Cen. l.u)>c Oil Pump 1Bl 0029 Noisy bearing Realigned Przg. Spray Vl)OOF 0030 Leaks pas't scat 1>ound loose scat repaired A Waste Cas Comp. 0039 Leaking Diaphragm Replaced B.A. Malce-Up Strainer S2903 0061 Flange leak Clcancd, replaced flange 8 CCW I.C.W. Strainer 1B 0063 lligh bp Cleaned C Charging Pump lldr V2504 0166 Leak under lagging Replaced bonnet gasket Cont. Per'sonncl )latch Mech. Dr. Shaft 0171 Misaligned bearing Realigned flange bearing Cas Decay Tank Valve 6579 . 0174 Leaks past seat Replaced diaphragm h Diesel Gen. Air Start Sys. Lubricator 0176 Low oil floM Ad)usted 1 C CCW Pump 0178 Seals leak Replace seals L

~0 HECllANICAL CORRECTIVE HAINTENANCE ON SAFETY-RELATED EJUIPHElff

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E UIPMENT PWO 0 HAI.FUNCTION CORRECTIVE ACTION SD Ilx Relief V3431 0498 Leaking plug Removed, cleaned, replaced Radiation Monitoring Sys. FCV-26-02,04 0499 l.eak past seats Cleaned seats B Diesel Ccn. Air Comp (Diesel) 0501 Will not run Primed fuel system B Diesel Gcn. Fuel Transfer Pump 0502 Packing lank Repacked pump C ICW Pump 0505 Packing leak Repacked pump A Charging 1'ump Disch. RV2315 0514 Body to nozzle leak Replaced gasket & reset ll2 Return Chk. VI-27102 0515 l.eak past scat Cleaned seats Primary Water Chk V-15328 0516 Leak past seat Hachlned and lopped S.I.T. Test Line VI-07009 0518 Packing leak Repacked B CCW llx. ICW Scrainer 0520 lligh D.P. Cleaned Letdoyn LCV2110P 0525 Bonnet Seal leak Replaced seal & gasket Aux. 1W Crossconnect Valve, HV-09-13 0527 Bonnet leak Replaced gasket Conc. Vacuumgeliefi V 2520 0661 Springs damaged Mhile installing leak test flange Replaced springs A CCW llx Salt Water Disch. Line 0663 Requires inspection by Engineering Opened line and cleaned A CCW llx Salt Water Disch. Line 0664 Branch connection leaks Removed for repair (See 678)

A CCW llx Salt Water Disch Line 0666 Line cracked Welded crack A CCW llx Salt Water Disch Line 0667 Connection leaks Welded in neM sockolet lA I.PSI Pump Suction Relief RV-3483 0668 Body to nozzle leak Installed gasket Equilr. Drain Tank Inlet Strainer 0669 Plugged Cleaned HSIV Bypass MV-08-1A 0676 Improper torque Reset A CCW llx Salt Water Disch Line OCi78 Branch connection leaks Installed ncu sockolct B CCW llx 0682 Inspected & cleaned Cleaned I l .

A CCW llx Sal.t Water Strainer 0684 lllgh hP & loose bolts Cleaned, replaced bolts 1 B Diescls Air Scarc Relay V's 0690 Suspect dirt Cleaned B CCW llx. Salt Water Scrainer 0696 lllgh hP Cleaned 'I A Bh Make-Up Tank RV2132 Flange 0699 Flange leak Replaced gasket B I CWP Disch. Chk. V21208 0700 Flange leak Tighccn bolts 'I'WO B CCW llx Salt Water Discli Line I.ine leak Removed for repair (See Cont. Personnel llatch 0703 0709 Shafc bent in operating mechanism i.'i. Straightened shaft B CCW llx. Salt Water Disch. Line 0712 llolc in pipe Welded (Sce 0703).

ICW Pump Disch. Ildr. 0713 llole in pipe Welded Letdown LCV2110g 0714 Bonnet leak Replaced seal ring . ~ ~ ~

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E UIPHENT PMO 8 HALF N(CTION CORRECTIVE ACTION B CCM 1lx. Salt Mater Line 0717 Pin hole leaks Meldcd 1 C Aux. Feed Pump Turbine 0718 Low oil flow Change filter 1 C ICW PUIDp 0722 Seized Replaced bearings A & B SDC Loop Inst. Iso. Valves 0731 Packing leaks Rcpacked A Charging Pump Disch. RV2326 0734 Rclievcs at too low a setting Removed, inspected, retested B Charging Pump Dlsch RV 2326 0735 Leaks through . Reset, retestcd Przr. Spray VllOOE 0707 Stem pulled from plug Replaced pin C ICMP "I" Straincrs 0814 Lo ~ flow of lube watirr Cleaned Przr Spray Valve V-llOOE 0815 Pac..'ng leak Rcpacked Letdown Nx Inlet RV2346 0817 Lifts early Replaced valve B Charging Pump Packing 0843 1'eeking leak Replaced packing 6 plungers Charging. Line to 181 l.oop 0898 Blind flange leak on V2805 Tighten flange CEDH 13,44 NA 44 didn't operate when cold Replaced, rctested (by CE) 13 showed signs of pnssible failure (didn't withdraw properly on last startup)

ELECTRICAL CORRECTIVE MAINTENAN ON ETY-RELATED EQUIPMENT E UIPMENT PWO 11 HALFI.NCTION CORRECTIVE ACTION Diesel Cen. lh Annunciator (hnn.) 4i122 Spurious Alarm Replaced relay 1A CCW Pump hnn. S-51 4301 No isolate switch alar>>> Replaced lifted leads lA Aux. FWP 3324 In/Bd Bearing heating i p Replaced motor filters 1B CCW Pump 3971 Sealtite conduit broken Repaired conduit 1B IIPSI Pump 3395 Ovarcurrent relay not functioning AJ)usted contacts lA 6 18 LPSI Breakers 4015 Isolate switch deleted from design Lifted lead to agree uich latest CWD revit>ion MV-09-13 Aux.. FW Ilcader Cross-Connect 4036 Indicates both open and closed 'Replaced Belcvieu washer in operator lA Charging Pump 3915 Motor leaking oil Repaired piping SDC I"olation Valves HV-3480/3481 3387 Received spurioils not open>i alarm Corrected uiring Shield Bldg. OucsidcAir Supply Valve 3989 No op<< indicating ligllt hd)usted lbmit switch FCV-25-11 i HOV 3617 (UPSI) 2754 Damaged conduit Repaired conduit 1A Battery Charger 3391 hmp meter indicating low Ad)usted charger voltage lA Battery Charger 3320 Spurious alarm Replaced phase sequence relay lh Bactcry Charger 4307 No amp meter indicatiou Change circuit bd. MBC 1971 tlv-3656 IIPSI 2765 Spurious valve not open alarm Corrected wiring to agree with CWD tN-21-3 ICW Non-Emerg. Ileadcr Isol. 3380 Won' open 6 close Ad)usted operator clutch mechanism Aux. IIPSI/l.oop 181 IICV 3737 3396 Opens beyond throttling setting Ad)usted limit jhl fuse switch'eplaced SI Tank Disch. Valve HV3624 3381 Won' operate HV 07-lA RWI'uclet '253 Defective corque switch Replaced switch Containmi'.nc Sump 4009 Valve won't close fully Reset tofqilc switcll Boron Control Valve 2525 3957 Overload tripped Repel.red limit suitch tDI 2504 B.A. Hake-Up 2775 Indicator Wrong Ad)usted indicator Volume Control Tank Iso. V-2501 4063 Motor bad - excessive cycling Replaced motor (See PC/H 181-76)

VC Tank Discharge V2501 3322 Overload tripped Reset (Sec PC/M 181-76) 1B Rx Sump Pump 4311 Penetration bad electrically tlove to spare penetration MV 3645 RTOB Indicator - NPSI 3389 Position ind. defective Repaired wire in meter B Battery Charger 3314 Spurious alarm Reolncril pliasr. failure relay Control Rm A/C ACC3C 3949 Disconnect switch bad Replaced switch Aux. FllP 1C 334i 3 Tripped (ov<<rspeed) Rcpaiied wiring tIV09-ll Aux. FW Pump 1C Disch. 3917 Won't close ~ ~ Cleaned contacts I 18 Battery 3988 Cell No. 31 bad Replaced cell 1A LPSI Pump 4109 I.caking oil sight glass motor Rel>laced sight glass 1C Aux. FW Pump Iso. HVOB-3 4282 Valve starts open then stops Replaced torque suitch Aux. FIP tIV09-ll 4293 Valve won'c close R<<moved sand from torque SW B.A. Neat Trace CVCS ~ 413l llcater off Repaired shorted section B.A. Ilcat Trace CVCS 3333 Grounded Repaired grounded section B.A. Hake Up Pump 3907 Low temp. alarm (spurious bad thermocouple) Repaired tlier=ocouple 1C Aux. FWP 4142 Control )unction box hna excessive moisture Sealed control box 1C hux. FWP Steam Valve HVOB-3 3346 Will not open Repaired wiring lug

E UIPi/EHT p/Q 0 MALVlJNCTION CORRECTIVE ACTION Brkr 60308 Aux. FWP 1C 2789 Cround Rcmovcd pumper to make circuit agree with CWD 1C Invertcr 3960 Won't take load Chance oscillator board 1C Invcrtcr 3996 Low voltage alarm (spurious) Ad)usted alarm relay DC Bus lA 3321 Cround on Ckt D114 Repaired 1C Aux. FWP 4317 Solenoid stuck Lubricated 1st'ch 1C Invcrtcr 4094I No AC output Replaced burnt wiring 120 VAC Vital Bus Invcrter 3950 Tripped ~ Replaced blown fuse 120 VAC Instrument Bua "HG" Inverter 4303 Ni Voltage Alarm Ad)usted contact on relay Charging Pump Seal Lube Pump 1B 3397 Motor won't run Replaced motor bearings (See PC/H 75-76)

Charging Pump Seal Lube Pump 18 4008 Motor won't run Replaced motor (See PC/H 75-76) i i ~

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INSTRUMENT & CONTROL CGRRECTDF. MAINTENANCE ON SAFETY-RELATED E UIPMENT E UIPMENT PHO 0 MALFUNCTION CORRECTIVE ACTIOiV Nuclear Instrumentation Detectors 3083 Incorrect detector hk gh voltage Reset & measured the voltage on wide range and linear power channels Remote Wide Range Start Up Rate Meter 3084 Ch. "B" not responding to input signal Replaced transistor and op. amp. in sigma meter Nuclear Instrumentation 3693 Wide range "A" failed functional test Period meter ad)ustment pot. was ad)usted I

Nuclear Instrumentation 4035 Inoperative rate ckt. on Ch. "C" wide range Reset pretrip set point-rechecked period ckts.f Wide Range Nuc. Pwr. Ch. "B" 4059 Indicator inaccurate Calibrated indicator "B" NI Channel C 4256 SUR pretrip and DPM appeared to be off setpoint Checked set point - no abnormal deflections Nuclear Instrumentation 4463 Replace feedback resistor, Reset extended wide range bi-stable-replaced resistors on all four channels I NI 4476 Readjust linear amplifier gains Read)usted upper & lower gains for Ch. B,C & 10 NI 6039 Replace all.trip test pots. on linear pwr drawers Replaced pots and resistors in all drawers NI 6055 Replace R-7 & R-26 on all linear pwr drawers Replaced pots and ad)usted the output on R-7 Wide Range Log Channels 6143 Ch. A & B off on cal. check Ad)usted DPH meter and bi-stable NI 6151 Linear pwr. control Ch. 89 reads 2X pwr 8 no Ad)usted span resistor, verified proper .readings voltage input Wide Range Ch. "C" 6205 All calibrate positicns read high Ad)usted and calibrated Wide Range Ch. "C" 6512 Repair faulty high voltage connector Replaced bad male connector & ray chem heat shrink tubing Pressurizer Temperature 3278 Failed RTD - requires replacement Installed spare RTD RCS - 1B1 Cold Leg Temp. Element 3288 Erratic signal from detector TE 1125 Replaced RTD - calibrated Prcsswirizer, llIC-1100 Spray Valve Control 3533 Continuous open signal to spray valves Ad)usted lower limit on controller Stm, Cen. AP Transmitters 3558 Valves leaking PDT-llll A,B,C Replaced valves and leak checked RCP lBl Oil Lift Pumps 3591 181 oil lift pumps running in auto (speed Ad)ust sensitivity to provide proper ckt. action switch out)

Pressurizer Level Control 3735 LIC lllOY does not give indication Replaced deviation amp, cleaned meter, calibrated RCP 1A1 3848 T1A 1159 & 1158 inputs are reversed Reversed elements-verified normal indication RCP 1A2 Upper Seal 4009 Low alarm failed PIA 1162 & 1173 ~ Installed rctro-fit kit & re-calibrated Pressurizer Spray 182 4017 Temp. output varying causing alarm Tightened lugs on RTD-Satisfactory performance RCS Ni Cold Lcg Temp Alarm is set too high 4024 Replaced temp. indicator-& rccal. of setpoint RCP 182 4056 Spurious hi vibration alarm (Ann. J-28) Ad)usted sensitivity of switch RCS AT Power Ch. "D" AT Power signal is abnormal 4089 Installed pwr supply-checked against Ch. A,B & C Quench Tank Pressure 4109 Failed to alarm at setpoint Replaced blown fuse Containment RTD's 4119 Check for loose leads at RTD's Tightened leads at containment penetration I RCS Cold J.eg Tcmpernture 4142 Different readings on temp indicators Calibrated sigma temp. indicator RCS Cold Leg Temperature on Hot Shut- .'4458 Out of calibration specs. Failed calibration - replaced.with spare down Control Panel oa'. i I m],  !

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E UIPHENT PWO ff MALFUNCTION CORRECTIVE ACTION RCS Pressure Lou Lou Sctpoint 4462 Press. ind. will not nllou SDC valves to open Replaced press. ind. controllers-cali-brated & installed. Vcrificd indications agreed. f RCS-PDI 1124X 4468 Failed specs during RCS flow test Replaced servo meter & calibrated meter RCS-PDI 1112 4469 Failed specs during RCS flow test Heter calibrated RCS Lo oP lA Cold I.eg Temp. 4473 TIC-llllYis non linear & out of tolerance Replaced scale 6 calibrated-rcinatallcd,meter RCS Loop lA Cold Leg Temp. 4480 Transmitter output cycles to max & locks in Replaced circuit board and recallbrated unit Quench Tank 6030 Check cal. on quench tk instrumentation Ad)usted zero on PT-1116-replaced oscillator on CT-lll6 I RCS-PDT 1121A 6040 Bad zcnor diode Replaced diode and tented unit l

RCS Transmitter Inst. Valves 6369 Leaking past packing inspected & repacked around A & B S/G DC Pouer Supplies 6376 Change uiring to agree with CMD Ran neu wire RCS-PDT-1121D (Loop hP) 6385 Repair transmitter Could noc be repaired-returned to factory Pressurizer Pressure 6470 PIC 1100Y Indicating incorrect installed replacement oscillator amplifier Pressurizer Spray Valve 1100E 6479 Remove 6 replace instrumentation for valve rpr Replaced instrumentation & stroked valve Quench Tk. Level Indicator 6483 Level fluccuaces between 42Z & 58X Replaced faulty meter with calibrated spare 1'rcssurizer Level Control LIC lllOX 6489 Output meter sticks Rebuilt output meter and reinstalled Engineered Safcguards Cabinet Mod. 3133 Isolation module wiring changes Performed uork pcr PC/H ff31-76 Enginccrcd Safcguards 3549 Accuation module SB cab. trip light dim Corrected bad solder point Engineered Safcguards 3550 ATI przr. press. function not resetting Rcpaircd bad connection 6 module in SIAS Actuacion lfodule (safcguarda) 3575 Hodule does noc reset Replaced IC V3 - rccested sacisfactory ATI Module (safegoards) 3576 Docs not sequence through logic check Replaced IC Vl rctcsted satisfactory Lou Pl.zr. Prcssure Trip Ann. 3654 press. trip comes in before pre-trip alarm

'ow Hade setpoint change pcr CE letter F-SF-835 ESC-lfh Bi-stable lfodule 4065 Setpoint dial pot. has broken locking device Rcplac<<d poc. 6 read)usted bi-stable ESC Cab "A" 6077 Current ad)ust for test ckt. erratic Replaced pot. - tested satisfactory FSC 1'rzr. Press. Hater PIA-1102 All 6353 Low alarm is in-operative Installed retro-fit kic - checked cal & alarm PPS Stm. Cen. Level 3255 LT-90138 Secpoint drift Installed retro-fit kit - cal'd. satisfactory RPS CIP Panel 3687 T Not Digital readout too high Repaired loose leads at TE 1112 llB RPS Indicating Relay 3589 PIS-07-2A Pegged Doun scale Repaired bad torque suicch-test satisfactory RPS Adder 6 Multiplier Modules 3659 Voltage greater than 10 HV Replaced modules - tested satisfactory RPS Trip Unit Ch. "D" 3845 Resistors overheated Replaced and rctcsted RPS-T"fLP Ch "A 3870 THLP Bi-stable uould not open Reuorked module-re-test operation satisfactory RPS - Pouer Supply Ch. "D" 3871 -15 V failed to -21.00 V Ad)usted pouer supply RPS Cabinet "D" 4043 ffires have wrong polarity co metro scope Reversed leads s

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INSTRUMENT & CONTROL CORRECTIVE HAItlTENANCE ON SAFETY-RELATED EjlUIPMENT E UIPMENT PWO jt HALFUNCTION CORRECTIVE ACTION RPS 4124 Noise on instrument loops (Check) C'hecked-for grounds on lifted shields RPS T Not Ch. "D" 4130 Temp. transmitters erratic Removed grounded lend on TE-1122 llD RPS 4248 Read)ust core protection calculator setpoints hd)usted coefficfencs-Set pwr. racio calculator RPS Ch. "A" 4251 Comparator module failed Replaced defeccive dfode RPS Cab. "D" 4451 Positive diffcrcntial alarm is inop. Installed recro-fit kits-cal'd-ad]. set-points.

Reactor Protection Syst. Cabinet "h" 4488 tluclear Pwr. AT pwr. X meter failed Installed retro-fit kit and calibrated Reactor Protection Syst. 6046 Relay socket onhn2 'logic matrix relay Replaced socket - resoldered shields Reactor Protection Syst. 6206 S/C low press. Bi-stable will not activate Installed spare bi-stable-checked trips Reactor Protection Syst. 6312 'IT-1122 llh Produces no output Replaced zener diode & cal'd within specs. J RPS Ch. "B" 6284 CPC-1 C-2 module cxhi'bfts 10 MV offset Replaced module and retested RPS Chz HA TM/l.P RPS tilde'Range Ch. C 6347 lli alarm inoperative Installed retro fit kits, cal'd & set alarms 6366 DPM Heter indicating fncorrect Replaced meter and re-cal'd.

RPS ljide Range Ch. C 6377 llf. start-up race trip setpoinc ie above litjite hd)usted setting on crfp and pre-trip RPS Cnb. C 6304 Llspulr 15V power suppLy C-1 Busuldurel lusdu-ehsch!jl Control Element Drive Motors (CEDM) 3527 Replace defective reef switch ass'y. Replaced two defective switches with spares CED.'I CCY Timer Modules 3530 Replace Defective diodes in 12V pwr. supply Replaced diodes and tested Control Element Drive System (CEDS) 3727 CEA jj33 intermittently drops rearranged wire to eliminate noise problem CEn.l 3859 I.ower limit light on Rod jjlB did not clear Removed position transmitter & replaced with spari CEDS CPP 4120 Repair damaged connector sockets Removed modules-replaced damaged sockets CEDM CCP Timer Hodules 4492 Replace 12V pwr. supply diodes Performed per CE Letter F-SF-920 CEA jj50 6047 Dropped cannot retrfeve Replaced 15V power supply CEDM and Reed Switches 6092 Replace damaged connectors Replaced connectors per procedure 8770-8202 R-0 CED."l 6093 jjlO blows fuses-lj13 wfll not raise. correctly Located grounded pfn-replaced damaged cable CEDM Coil Stacks 6289 Remove for repair (position indication) Repaired coll stacks & reed switches CEDS. Control System 6404 Troublcshooc & repair in support of start up Replaced CEA module jj7 & 34 Timer Module ]55

& 61 CEA 3163 Adjust reed switch position transmitters Heasured resistance & adjusted (8770-6947)

CEA jj29 3704 Lower electrical Immit switch shorted Replaced defective switch-tested Ops. Proc.

0110081 CEA j"31 Reed Switch 3748 Improper positions indications & accuations Soldered broken wire in position transmitter Boric Acid Make Up Tank lh 3191 Low low level alarm failed Install retro-fit kit & cal'd. LIA 2206 Charging Pumps llcader Pressure 3640 PT-2212 Internal leak Replaced bourdon tube & re cal'd.

Charging Pumps llcader Pressure 3657 PT-2212 leaking Replaced with Spare-installed snubber-re cal'd.

lA Charging Pump 3690 Seal Water Tank Level Indication Inoperative Freed float in the indicating assembly Volume Control Tank '6002 Pressure Regulator not working correctly Reset regulator

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Boric Acfd Hake-up Tank 1B 6089 Level indicator failed high Disconnect fittings-cleaned flow meter- rcconnectu lA & 1B BAM Tanks 6140 React low level alarm Installed retro-fit kits-reset alarms 18 BAM Tank 6236 Recirc. valve indicates intermediate position Ad)usted limit switch lh BAM Tank LIT/LIA2206 6418 Level transmitter air line plugged Flushed lines-returned to service BAH Isolation Valve. 6459 No indication of valve position Snap lock switch sticking-freed arm oj

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INSTRUMENT & CONTROL CORRECTIVE HTENANCE ON SAFETY-RELATED EQUIPMEHT E UIPM'tT puo ii MALFlNCTION CORRECTIVE ACTION Volume Control Tank 6509 Low alarm not functfc ning Miring error on sigma contact-corrected I "A" Charging Pump 6523 Low level alarm on st el tank Corrected indicator alignment Volume Control Tank 6524 Low level alarm not operating Replace sigma instrument with calibrated spare.

Safety Infection Tk. 1Bl 3146 Abnormal fluctuation on LIA-3331 Replaced with spare rccalibratcd S.I. Tank 1Bl 3564 llfgh-high level alara. Vented transmitter LT-3331 I S.I. Tank lA2 & 181 3565 Nigh and low level a)arms Vented transmf tters-read)usted setpoint'n LIA-3331 S.I. Check Valve Leakage to RUT 3566 Valve leaking by seat. Re-zeroed valve positfoner Shutdown lleat Exch. 1B 4484 Temperature indicatirn incorrect servo-motor-re-calibrated I'eplaced Safety Infection Tank 1B2 6497 llfgh-high level alarm Ad)usted transmitter-zero Gas Depay Tank 3554 Plow indicator incorrect PIT-6648 removed unit-cal'd. spare & installed Haste Gas Compressor 3599 Setpofnts need ad)ustment Ad)usted PS-6647 2 & 3 Gas Decay Tank 4025 Flow meter gives inaccurate flow rata Added 3/8" orifice in gas release line llaste Gas Comp. "A" 6495 Hot cycling properly Re-calibrate pressure switches Haste Cond. Tank LIC-6640 & 6641 6290 Cannot pump tanks Joi er than 22X Checked calibration & verified setpoint

'ctuation limit switch & valve stem collar Containment Spray FCV 07-lB 3706 Valve indicates closed when open Ad)u .ted Reactor Containment Bldg. Prcssure 6026 PIS-07-2D oscillates Tiglitencd all terminal screws Reactor Contafnmcnt Bldg. 6357 PIS-07-2A Servo sticks Replaced servo - ca'ld. & ad)usted setpoints Reactor Containment Bldg. 6364 PIS is connected wrong for lower alarm Corrected wiring lla f n Stcam Dump Valve 3632 PlC-08-181 reads above indication Repaired and re-installed lA Main Stm. Isolation Valve 3671 Dfsphragm leaking Replaced diaphragm and tested lA Main Stm. Isolation Valve 3672 Solenoid light stays on Repaired switch arm & tested lB Ilain Stm. Isolation Valve 4032 Ground in controls Ground in LS Replaced switch "A" <tain Stm. Dump Valve 4108 PIC-08-1A Setpoint problem Balanced controller to hold sctpoint input 18 Plain Stm. Dump 6122 PIC-08-181 indicates too high Replaced capacitor & re cal'd. unit Main Stm. Prcssure 6235 PT-08-lA Output saturates up scale Checked m<<ch. linkage-replaced detector coil Hain Stm. Prcssure 6476 PT-08-1B fails intermittently Replaced with calibrated spare transmitter Feedwater Reg. ByPass S/G 1B 3115 LIC-9006 inoperative Corrected remote 6 local process linearly

. problems Stm. Gen. 1A Level 3251 Unable to determine actual level Filled ref. legs on transmitters & vented D/P blocks "B" Aux. Feed Pump 3510 No flow indication with pump in operation Replaced transmitter force motor & tecalfbrated "C" Aux. Feed Pump 3590 Control problem-overspeeds when started Termfnal connection on servo actuator loose Aux. Fcedwatcr Flow Header "B" 3738 Gauge reading zero with flow through header Reset zero to 4 ma DC 6 returned to service lA Main Feedwatcr Plow 4118 Indication failure Replaced PE-09-1Al- Cal'd - returned to scrvfcq S/G Level."B" 4192 Level reads high on L-9023 A & L-9021 Tightened fittings & valve stems Feedwatcr ByPass Control 4490 LIC-9006 has erratic operation Replaced resistor 6 ad)usted-stroked valve 18 B.A. Makeup Tank 6469 Spurious level alarms Replaced remote amplifier, calibrated i I

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INSTRUNEHT & CONTROL CORRECTIVE INTENANCE ON SAFETY-RELATED EQUIP}1EHT E UIPHEHT PWO 0 HALFCHuTION CORRECTIVE ACTION Stm. Cen. 1B Level 6370 Hi level controller actuation inoperative Installed mod. kit-cal'd. & re-installed Stra. Gcn. 1A Level 6371 lii level controller actuation inoperative Installed retro kit-cal'd. & re-installed Stm. Cen. 1A Level 6372 Transmitter out of cal. Removed,aligned 6 cal'd., replaced 6 tested 1C Aux. Fccduaccr Pump 6387 Controls are rusty and full of water Cleaned, removed moisture, lubricated f Stm. Ccn. 1B Level 6468 Incorrect level indication on 1.IC-9023G Vented instrument & tightened equalizer-valve Stm. Cen. 1B Level 6484 1li controller contacc is on constantly Installed retro kic and calibrated-tested Stm. Cen. 1A Level 6521 1li level alarm uith normal level Installed retro-kit and cal'd - tested CCW HX-1B Ouclec Pressure 3098 PIS 14-SB indication is erroneous Replaced force motor in transmitter & cal'd CCW NX-1B Outlr.t Pressure 3887 PIS 14-8B In'ication is erroneous Replaced force motor and re-cal'd.

CCW From Letdown 11X Flou 6390 Flow indication on FIS 14-6 & valve is closed Replaced oscillator amplifier CCW llcader D Flou 6403 Ill/Lo Alarm on uith normal flow Installed retro-fit kit and cal'd.

Shutdown Iieet Exch. CCW Outlet 1B 6450 Air supply line to valve broken Repaired broken line (11CV-14-3D)

Primary Water Valve to Containment 4271 Annunciator for HV15-1 noc working Checked 6 installed arming acre'w IA Emergency Diesel 6356 Starting air lou press alarra uon't clear Suitch scicking- replaced-functionally checked Intake Water Level 3574 Alarm does not actuate on low level test Recalibrated ICW Disch. lleadcr A 3691 Transmitter failed Replaced internal parts uith spares Intake CW Disch. llcader A & B 6065 Ho bleed off valves betueen isolation valves Installed valves as req'd per Tech Specs and press. transmitter CCW Hx Inlet D/P Ann. 6088 Docsn't alarm D/P Indicator beyond repair-replaced indicator Incake Cooling Water Pump A 6303 Flou indicator inaccurate Fl-21-3h Removed, cleaned & .repair flou meter CCW Nc. Exch. 1B 6367 Spurious Lou Flow Alarm FIS 21-9B Lifted leads, cleaned corrosion, alarm cleared IntaRc Cooling Pump Lube Water "Cu 6378 Failed Lou Flow Alarm PS 21-46 Cleaned sui tch, opera c ion satisfactory CCW.llc. Exch. 1D 6480 Barton pegged low FIS-21-9B Contacts closed - replaced switch S-G 1A Bloudown PS-23-6 4481 Press. Suitch has ruptured bellows Installed neu switch-tcstcd satis'factorily Containment. Vac. Relief Valve 3114 FCV 25-7 failed open Removed water from sensing lines-test ok ECCS Roora D/P 3168 Transmitter failed high Replaced oscillator amplifier assembly Shield Bldg. Vent D/P Alarm 3228 Ann. has alarm-indicator reading normal Installed retro-fit-reset setpoints 6 cal'd.

Containment Personnel/Escape Locks 3286 Repair monitors Replaced bad I/O and cal'd. monitor Pcrsonncl llatch 3854 Outer seal suitch actuation arm loose Fabricated new pin & installed-tested sat.

ECCS Emerg. Fun llVE-9A '4255 Spurious Alarms on lou flow or motor overload Ad)usted arm contact acreu on flow switch

- retro-kic - operation satisfactory a

Slrield Bldg. Vacuum Alarm "A" 6240 Low vacuum alarm condition Replaced Fuel Pool Exhaust Fan llYE<<16A 6415 Low flou/overload alarm Ad)usted flow suitch Containment Purge Sample Valves 6421 Sample valves indicate midway position Read)usted limit suitches - ceated satisfactory I

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INSTRRfENT & CONTROL CORRECTIVE HAINTENANCE ON SAFETY-RELATED El}UIPHENT E UIPMENT PWO 0 MALFUNCTION CORRECTIVE ACTION Control Rm. A/C NVA-3C 6460 Received trouble failure alarm Ad)usted FS-25-10C (flow switch)-test sat.

Shield Bldg. Vent. Vacuum Alarms 6514 Setpoint too low on PDIS-25-7A & B Increased setpoints ad)ustment per set point list (was 4" Wg vacuum; now is 4.5" Mg Vac.)

Containment Isolation Monitor d4 3086 Non-functional Replaced defective V/I unit test signal Containment Isolation Ch. 3 4050 Nigh MA readings Problem in HA detector-cal'd. & returned to service Ch. "A" ARHS Recorder 6232 Recorder & ECCS meter indication differ Ad)usted V/I to correspond with proper input

& output voltage ARMS Monitor Cab. "A" 6299 Take up reel broken cn chart recorder .Soldered broken take-up reel Cont. Rad Honitor RIS-26-3-2A 6345 HR/Ilk Reading varies from ESFAS panel Replaced V/I <<ad)usted & cal'd new V/I to Containment Isolation FCV-26-04 6352 Valve failed leak test limit switch-stroked valve loop'd)usted Containment Isolation Ch. 5 6399 Nigh alarm light not working Repaired lead & resoldered N2 Analyzer I-FSE-27-6 3265 Valve Stuck in closed position Adjusted stroke N2 Analjjzer Sample Panel 6401 No closed indication for valve 27-09 Ad)usted micro switch & replaced fuse U2 Analyzer Sample FSE-27-10 Does not indicate open when cycled '451 Repaired bulb holder-ad)usted micro sw..

Refueling Hachine 6170 Setpoint load ad)usta.ent Ad)usted setpoints per CE Letter F-SF-0981 Spent Fuel Handling Hachine 6348 Moist stuck in. upper elevation position Read)usted overload & fuel load limit switches Refueling Machine 6375 Setpoint change Reset underload setpoint Upend carriage 6388 Overload occurs during transfer Corrected static zero on load cell-checked setpoints Post Accident Panel Recorders 6239 Chart drive mechanisms not functioning Ad)usted all chart retainer assy-replaced properly motor in 8159

Page 24 DESIGN CHANGES On the following pages are descriptions, including a summary of the safety analyses~ of the design changes implemented at St. Lucre Unit 81 during the period Mi rch 1, 1976 (issuance of operating license) through December 31, 1976.

Page 25 Plant Change/Modification'-76 Unit 81 "CEA GUIDE TUBE IRRADIATION TEST PROGRAM" This Plant Change/Mbdification provides for installation of 3 Zircaloy test specimens in the St. Lucie Unit 81 core. The specimens are patterned after the neutron source assemblies employed in several currently operating Combustion Engineering reactors. The test installed in fuel assembly guide tubes in an arrangement identical specimens're to that used for the neutron source assemblies. The program will confizm the growth of cold worked Zircaloy matezial under actual "in-core" conditions. This was done at the request of Combustion Engineering.

This change is not an unreviewed safety question because:

I (1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Design, construction and installation were done under the same criteria and procedures as were othez in-core assemblies such as neutron souzces, surveillance capsules and in-coze instrument thimbles. The Nuclear Regulatory Commission has reviewed the Combustion Engineering program on a generic basis and concluded that "the health and safety of the public and plant personnel will not be affected by the program." (Letter Olan B. Parr to A. E. Schezer dated November 21, 1975).

(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

(3) The mazgin of safety as defined in the basis for technical specifications had not been decreased.

This change does not represent a change in the facility as described in the Final Safety Analysis Report.

Page 26 Plant Change/Modification No. 7'76 PSL Unit 81 "SAFETY SH01KR AND EYHVASH STATION AT CVCS CHitfICAL ADDITION STATION" This change installs a safety shower and eyewash station at the Chemical and Volume Control System chemical addition station to comply with OSHA reguations.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The, water supply being tapped is not safety related and the area (Reactor Auxiliary Building) involved has been evaluated regarding possible flooding.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

t The non-safety zelated electrical junction box in the area was waterproofed as part of this PC/M.

3. The margin of safety as defined in the basis foz technical specifications has not been decreased.

This change does not repzesent a change to the facility as described in the Final Safety Analysis Report.

Page 27 Plant Change/Modification No. 10-76 PSL Unit 81 FUEL HANDLING BUILDING RADIATION MONITORING FLOW CONTROL VALVE" This bleed-off valve motor was wired so that it was always energized.

Thus, when the sampling pump was turned off, the valve was run against its travel stop trying to raise system flow by shutting off bleedoff flow.- The motor 'was rewired so it is de-energized with the sampling pump and won't damage itself. Also an extra ground on a shield cable was removed by this PC/M. The one ground recommended by the vendor is still installed.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accidental or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Any leakage past this valve is filtered in the sampling unit and is returned to the filtered ventilation system anyway. This ventilation sampling system is not safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No functions or design intents were changed and no new components were added.

3. The marg'in of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 28 Plant Change/Modification 11-76 PSL Unit 81 "WIRING CORRECTION FOR PDIS-25-16A and B" This Plant Change/Modification changes the terminals'or alarm relay wiring to give a closed contact for the low vacuum (high pressure) alarm. As originally wired the alarm was set up for low pressure instead of low vacuum.

Changing the terminals for two wires causes the alarm to function properly.

This was discovered during preoperational testing shortly before receiving our operating license and correction reviewed/processed via the Plant Change/

Modification program.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change does not alter any functions of the instruments; it gust corrects the as-built system to comply with the original design intent (alarm on low differential pressure between Emergency Core Cooling Systems pump rooms and remainder of Reactor Auxiliary Building).

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specificatiom:

has not been decreased.

This change does not represent a change in the facility as described in the Final Safety Analysis Report,.

Page 29 Plant Change/Modification 12-76 Unit 81 COVERS FOR EX CORE ~a TRON DETECTORS'luminum covers were fabricated and installed on the top of the wide range excore neutron detectors to prevent dust and debris from entering the detectors through the top opening. The debris had been reducing the insulation resistance (outer shield to ground) below the required value (1 x 106 ohms). The detectors were cleaned before installation of the cover plates.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The cover plates are passive elements which prevent the entrance of dust or debris and thereby reduce the probability of detector failure due to electrical grounds and insulation deterioration. These failure modes are already considered in the Final Safety Analysis Report.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The materials and methods of construction used're the same as the original detector/housing/lift mechanism assembly which has not been changed by this addition.

3. The margin for safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change in the facility as described in the Final Safety Analysis Report.

Page 30 Plant Change/Modification No. 13-76 PSL Unit 81 "ALLOCATION OF A PERMANENT CABLE TO CONTROL ROOM FOR THE RADIATION CALIBRATION FACILITIES" This change designates a previously abandoned (spare) cable, which goes to the Control Room Radiation Monitoring Cabinet, as a permenent "Testing Facility Cable" and reroutes the last 30 feet of the cable from a cable tray in the Electrical Penetration Room to the testing facility via conduit.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The cable will not be permanently terminated but used only for test/

calibration purposes.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This item is not discussed in the Technical Specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 31 Plant Change/Modification 14-76 PSL Unit 81 "SOLENOID VALVE REPLACEMENT ON REACTOR DRAIN TANK VALVES" This Plant Change/Modification changed the air line solenoid valves for Reactor Drain Tank Isolation valves V-6301 and V-6302. These 2 yalves are Containment Isolation valves and must close within 5 seconds. The orifices in the original solenoid valves were too small to meet this specification so solenoid valves with larger orifices were installed and tested satisfactorily. Need for this change was determined during preoperational testing before receiving our operating license and the change processed via the Plant Change/Modification procedure.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Pinal Safety Analysis Report has not been increased.

The new solenoid valves are direct functional replacements for the original valves with the only difference being larger orifice size needed to meet closure time specification on V-6301 and V-6302.

2. The possibility for an accident or malfunction of a different type than any eva'uated previously in the Pinal Safety Analysis Report has not been created.

The new solenoid valves are direct functional replacements for the original valves with the only difference being largerorifice size needed to meet closure time specification on V-6301 and V-6302.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change was necessary in order to meet technical specifications regarding closure time for 2 containment isolation valves.

This change is not a change to the facility as described in the Final Safety Analysis Report.

Page 32 Plant Change/Modification 15-76 PSL Unit 81 "REMOVAL OF RELIEF VALVE RV-2185" This change removed RV-2185 and blind<<flanged its inlet and discharge.

This relief sometimes lifted when a boric acid makeup pump started and passed boric acid solution into the primary water system (non-borated system). While evaluating the valve and its setpoint it was determined that the valve was not needed due to, another relief, not discharging to.

primary water, 'located in the same section of line (within the same isolation valves). This was determined shortly before our operating license was issued and reviewed/processed per the PC/H procedure.

This change is not an unreviewed safety question because.'.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report, has not been increased.

The section of line involved still has overpressure protection.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety, Analysis Report has not been created.

The section of the boric acid makeup system involved is non-safety related.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 33 Plant Change/Modification 16-76 PSL Unit 81 "ADD HJNCTIONAL TEST CIRCUIT TO AREA RADIATION MONITORING SYSTEM"

. This change added a pushbutton, fixed resistor and a variable resistor to the ARMS Containment Isolation Signal circuitry. This allows internal generation of a test signal on demand and varying it to test the alarm functions and setpoints. Previously, testing of the circuit would have required opening the circuit and supplying a test signal from some external source.

This change'is not an unreviewed safety, question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The components involved are supplied by the original vendor to the original design specifications. Proper testing of the circuits will improve overall reliability.

2~ The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This modification does not cut the detector out of the circuit. As was the original circuit, it is designed to'fail safe". If the test circuit "fails off" it has no effect; if it "fails'n" it gives a high (conservative) signal.

3. The margin of safety as defined in the basis for techn'ical specifications has not been decreased.

Technical Specifications require testing of this circuitry to verify its operability.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 34 Plant Change/Modification No. 19-76 PSL Unit Pl "CONTAIKiEHT INSTRUMENT AIR DRYER OUTLET VALVES CONTROLS" As originally wired the valve remained open at all times (never closed).

This change corrected the problem so the valves close when the associated compressor/dryer is not running. The valves still fail open.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in theI Final Safety Analysis Report has not been increased.

This change corrects a wiring error to meet original design intent and, this system is not safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Pinal Safety Analysis Report has not been created.

See comments under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

See comments under (1) above.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 35 Plant Change/Yudif ication 20-76 PSL Unit 8l "RPS POWER RATIO AND CORE PROTECTION CALCULATORS" This change reverses the polarity of inputs to the summing amplifiers, increases the gain coefficient on the multiplier circuit, and changes the labels on 3 terminals in the Reactor Protective System Power Ratio and Core Protection calculators. This change was determined to be necessary at a similar plant of another utility, reported to us before granting of our operating license and reviewed/processed by the Plant Change/Modification program.

This change is not an unreviewed safety question bacause:

1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eval-i uated in the Final Safety Analysis Report has not been increased.

This change ensures the system will function as intended by the original design.

2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This change does not add any new design features or functions; it t

just corrects the design implementation so the system will function per the original design intent.

3) The margin of safety as defined in the basis for technical speci-fications: has not been decreased.

This change does,not represent a change to the facility. as described in the Final Safety Analysis Report.

Page 36 Plant Change/Modification 21-76 PSL Unit 81 "REACTOR REGULATING SYSTEH CEA STATUS LIGHTS" This change removes 2 interposing relays (RTGB-104) in the "high" and

".low" CEA insertion rate status light circuitry due to electronic noise interference with other instruments. These relays were intended for functions which previously had been deleted from the design and now serve no function.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The affected circuits do not have any safety functions.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

A superfluous active element has been removed from each circuit and no new functions have been added.

3. The margin of safety as defined in the basis, for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 37 Plant Change/Modification 22-76 PSL Unit 81 "SAFETY INJECTION LIHES (2) PRESSURE TEST" This PQi was for connecting )umpers and documenting pressure testing of two (2) vent valve lines on the SI headers to the Reactor Coolant System. Jumpers were installed (and later removed) by a PM as .the lines were isolated from the RCS only by check valves and the RCS had to be hot and pressurized to perform the pressure test. This was done after core loading but before initial criticality.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

System was restored to normal after pressure test. Had the (one inch) lines or Jumpers failed, the SI head'er check valves would have prevented any leakage from the RCS itself. Safety injection tank levels were monitored during the test to ensure they remained within limits.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Change was temporary; jumpers were removed and restored to normal after the pressure test.

This change does not represent a change to the facility as described in the Final Safety Analysis Report

Page 38 Plant Change/Hodif ication No. 23-76 PSL Unit 01 "REROUTE DRAINS TO REACTOR CAVITY SlgiP" This change reroutes certain drains from the Reactor Drain Tank to the Reactor Cavity Sump. The drains involved (Reactor Coolant Pump seal cooler CCW relief valves and containment instrument air compressors moisture separators) contain chromated Componerit Cooling Water which could eventually be returned to the Reactor Coolant System when Reactor Drain Tank water was processed. The drains now go to the Reactor Cavity sump which is waste water and is not reused.

This change is not an unreviewed safety question because:

I

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The drain lines are non-safety related and have no effect'on the non-safety related components served by the lines (RCP and instrument air compressor Component Cooling Water).

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

t This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 39 I

Plant Change/Modification No. 24-76 PSL Unit 81 "TEMPERATURE INSTRUMENTATION NOISE REDUCTION" This change adds capacitors to the Reactor Regulating System (RRS) temperature circuits to remove electronic noise interference. The noise created undesirable oscillation in the output indications (up to +5oF).

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

/

The RRS is not safety related. And, this change does"not alter any design functions or setpoints of the involved system.

2. The possibility for an accident or malfunction of a diffe'rent type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

The RRS is not mentioned in the Technical Specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 40 Plant Change/Modifcation 25-76 PSL Unit 81 "BYPASS CONTAINMENT EVACUATION ALA&i DURING CIS A R.M. CHANNEL TESTING" This change installs a bypass switch to allow functional checking of Containment Isolation System Area Radiation Monitoring channels without actuating the Containment Evacuation alarm. Functional testing is done by injecting a high radiation signal to test the entire circuit and, previously, this actuated the Containment Evacuation alarm. Frequent actuation of this alarm during testing distracts personnel from their duties and may create a complacent attitude toward the alarm.

This change is not an unreviewed safety question because:

I

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The containment evacuation alarm is not safety related. In addition, use of the key operated bypass is administratively controlled.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report

has not been created.

The alarm itself is not safety related and the modification was designed and tested to ensure no other circuits were affected.

3. The margin of safety as defined in the basis for technical specifications i

has not been decreased.

The alarm is not discussed in the Technical Specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Rbport.

NOTE: This PC/M is closely related to and was processed/

performed at the same time as PC/M 26-76.

I "Page 41 t '

Plant Change/Modification 26-76 PSL Unit 81 "CONTAQMNT EVACUATION ALUM BYPASS DURING A.R.M. PANEL ALARM TEST" The Area Radiation Monitoring Panel alarm annunciators are tested by opening the circuits to ground which causes the annunciators to alarm. This change connects previously installed relays which bypass the test contacts for the containment evacuation alarm to 'prevent false evacuation alarms. Frequent actuation of this alarm during testing distracts personnel from their duties and may create" a complacent attitude toward the alarm.

This change j.s not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The Containment Evacuation alarm is not safety related. Also, this change does not alter any functions; it just corrects the as-built sy tern to include the relays already installed to meet the original design intent.

2. The possibili,ty for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The alarm itself is not safety related and the modification was designed and tested to ensure no other circuits were affected.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

The alarm is not discussed in the technical specifications.

This change does not represent a change to the facility as described in the Final Safety Ana1ysis Report.

NOTE: This PC/M is closely related to and was processed/

performed at the same time as PC/M 25-76.

Page 42 Plant Change/Modification 27-76 PSL Unit 81 REPLACRfENT OF V/I CONVERTERS ON AREA RADIATION MONITORING CONTAINMENT ISOLATION SIGNAL CKVWiELS This change installed larger capacity V/I converters, supplied by the original equipment vendor on the A. R. H. - C, I. S. channels. The system impedance was too large for the removed converters and loaded them down excessively, This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunc-tion of equipment important to safety previously evaluated in the Final Safety Analysis Report. has not been increased.

Replacement part meets or exceeds the standards of the original part.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Channel failure is already covered in the Final Safety Analysis Report.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 43 Plant Change/Modification No. 28-76 PSL Unit 81 "TURBINE CONDENSER dT RECORDER WITH ALARM" This change installs instruments (RTD's) with printout and alarm in the control room to monitor and record condenser AT.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident, or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change adds temperature indications and alarms to a non-safety related system.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See 1 above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change aids operators in meeting requirements of the environmental Technical Specification on condenser AT.

NOTE: This equipment has been installed but is not yet fully operational as functional testing is not satisfactorily completed.

Page 44 Plant Change/Modif ication 31-76 PSL Unit Pl "ESFAS CABINET MODIFICATION" This change modifies the Engineered Safety Features Actuation Signal Cabinets so that upon loss of power to a cabinet all channels except Recirculation Actuation and Containment Spray Actuation trip (formerly all channels bypassed).

This change is not an unreyiewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change does not alter any functions of the system; it corrects the as-built system to agree with original design intent.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in. the basis for technical specifications has not been decreased.

This change does not represent a change to 'the facility as described in the Final Safety Analysis Report. The change is necessary to conform to the Facility Safety Analysis Report.

NOTE: This change was reported as the corrective a'ction for Licensee Event Report 335-76-3.

Page 45 Plant Change/Modification 33-76 Unit 81 "INSTALLATION OF GAGE Llew% SÃJBBERS" This change installed snubbers to protect gages from pressure surges on pump starting and stopping. The safety-related pumps involved were the flash tank pumps and reactor drain pumps. Other pumps involved were in the waste treatment systems.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eval-uated in the Final Safety Analysis Report has not been increased.

TheI pulsation dampers (snubbers) used met the same or better standards as the original tubing and gages and are located adja-cent to the pressure gages outside the system isolation valves.

2. The possibility for an accident or malfunction of a different type than any evaluated previously 4n the Final Safety Analysis Report has not been created.

These systems are not evaluated in the FSAR accident analysis and, in addition, addition of passive components in a location isolable from the system is very unlikely to add a malfunction mode

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

Page 46 Plant Change/Modification No. 34-76 PSL Unit 81, "INSTRUMENT AIR PRESSURE SWITCHES-HAIN STEAM ISOLATION VALVES" This change replaces the present switches with new ones. The new type has a stainless steel bellows. Tne old type failed several times due to vibration induced bellows failure.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The new bellows will improve reliability of instrument air to the MSIV's as it is less susceptible to failure.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The new switch meets or exceeds the original specifications and is a direct functional replacement for the original switch.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

See comments under (1) and (2) above.

This change does not,represent a change to the facility as described in the Final Safety Analysis Report.

I

Page 47 PLANT CHANGE/MODIFICATION NO. 36-76 PSL UNIT 81 "COOLING FANS FOR- STATIC UNINTERRUPTIBLE POWER SUPPLY CABINETS" This change adds cooling fans and high temperature alarms to the SUPS cabinets to supplement the'natural circulation cooling originally provided for. The natural circulation cooling was adequate but caused shorter than desired preventative maintenance schedules to ensure SUPS reliability.

This chance is not an unreviewed safety question because:

1~ The probabilitv of occurrence or the consequences of an accident

,or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This chanae does not chancre anv design intent or function of SUPS and will enhance operational reliability.

2~ The possibility for an accident of malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Reliability of SUPS will be enhanced due to better service life of components. If a fan should fail, the cabinets will still be cooled by natural circulation as originally intended.

3~ The margin of safety as defined in the basis for technical specifica-

,".." .tions has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 48 Plant Change/Modification 40-76 PSL Unit 81 "MOVABLE INCORE FISSION CHAMBER AND .A,PLIFIER" This change installed a fission chamber instead of a self-powered rhodium detector. for the movable incore neutron flux monitoring system.

The fission chamber is smaller and more sensitive and can better measure detailed flux patterns and fuel densification gaps if any should exist.

The amplifier was necessary to power the chamber and electronically process the chamber output.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Ho design intents or functions were changed. The new equipment better implements the original design intent.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifica-tions has not.been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 49 Plant Change/Modification 41-76 PSL Unit /fl "CURRENT TRANSFORMER GROUNDS" This change removed superfluous ground connections (in RTGB-101) on the CT (amperage monitoring) circuits for 4160V and 6900V switchgear. The vendor supplied grounds on the CT circuits at the switchgear. Tt is company policy for personnel safety reasons to have one and only one ground on these circuits so the redundant grounds were removed.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Only redundant ground wires were removed. No . functions or circuits were changed. The remaining grounds are adequate for personnel/equipment protection.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Only redundant ground wires were removed. No functions or circuits were changed. The remaining grounds are adequate for personnel/equipment protection.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

These grounds are not discussed in the technical specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 50 Plant Change/Modification 43-76 PSL Unit 81 "MODIFICATION OF FUEL POOL PURIFICATION LOOP SIPHON BREAKER" This change plug welds the 4" siphon breaker in the spent fuel pool purification loop. This will allo~ interim use of the spent fuel pool as a source of RCS makeup in the event the RVZ is unavailable due to a tornado or other causes. A permanent makeup source will be provided and the siphon breaker reinstalled before spent fuel is placed in the fuel pool.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The siphon breaker will be reinstalled before spent fuel is placed in the fuel pool. This is administratively controlled per the plant backfit list.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The siphon breaker will be reinstalled before spent fuel is placed in the fuel pool. This is administratively controlled per the plant backfit list.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

The siphon breaker will be reinstalled before spent fuel is placed in the fuel pool. This is administratively controlled per the plant backfit list.

Page 51 Plant Change/Modification 47-76 PSL Unit 81 "CHANGE MECHANICAL SNUBBERS FROM LOCKING TO NON-LOCKING TYPE" This Plant Change/Modification changes the original INC locking type mechanical snubbers to Pacific Scientific non-locking type. The locking snubbers will not release until,the force which caused the acceler'ation is removed. If a pipe experienced "jerky" movement due to thermal growth, the locking snubbers could lock and thereafter act as a restraint (until the next cooldown).

Although we had a thorough monitoring program to ensure no problems on initial heatup, the snubbers were changed to non-locking type to eliminate this concern for future operations.

This change is not an unreviewed safety question because:

1. The'robability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The replacement snubbers meet or exceed the requirements of the original specification.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The replacement snubbers meet or exceed the requirements of the original specification.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

These snubbers are not discussed in the technical specifications.

This change does not/does represent a change to the facility as decribed in the Final Safety Analysis Report.

k

Page 5?

Plant Change/Modification 50-76 PSL Unit 81 C E A CHANGE MECHANISM TRANSVERSE DRIVE MOTOR This change removed the 4 horsepower transverse drive motor and replaced it with a Q horsepower motor (and larger motor overload protection). The original motor was too small and continually tripped out during preoperational testing. This was determined before receiving our operating license and reviewed/processed by the PC/M program.

This change is not an unreviewed safety question because;

1. The'probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This system is not safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This system is not safety related.

3. The margin of safety as defined in the basis 'for technical specifications has not been decreased.

This system is not discussed in the technical specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 53 Plant Change/Modification No. 51-76 PSL Unit 81 "INCREASE SENSITIVITY OF FLOW ALARM SWITCHES-CONTROL ROOM VENTILATION FANS'his change installed new flow switch paddies and balance springs on the Low Flow Alarm switches for the Control Room Ventilation fans. This increased the sensitivity so the switches respond to the air flow and clear the (previous) continuous low flow alarm.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The switches sense low flow; they cannot cause it. The new parts were ordered from the original vendor per the original specifications.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No functional changes were made and the switches now meet original design intent.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

The low flow alarm is not required by the Technical Specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 54 PSL Unit 81 MODIFICATIONS TO CONTROL ROOM AIR CONDITIONING COMPRESSORS To get the desired capacity the vendor originally converted standard "(Se"smical-ly qualified) 6 cylinder compressoxs into 5 cylinder units (one cylinder was blanked). When repairs weze needed on one unit we discovered, that this approach was no longer used by the vendor and repair parts were unavailable. The units were therefore restored by the vendor to 6 cylinder units using his conversion kits, which give slightly shortened stroke and therefore the same capacity as the original 5 cylinder units.

This change is not an unxeviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipmentimpozxant to safety previously evaluated in the Final

,Safety Analysis Report has not been increased.

Change was done by vendor using many original parts and replacement parts equal to or better than original parts.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Failure of all 3 units is already considered in the Final Safety Analysis Report. The pxesent type of unit has a longer and better satisfactory op-erational record than the 5 cylinder units.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

There are still 3 units of same capacity as the original compressors (two are required by Technical Specifications).

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 55 Plant Change/Modification 53-76 Unit 81

'Zhot INPUT TO REACTOR PROTECTION SYSTMi

'his PC/M reverses 2 pairs of leads foz each channel of Thot input to the RPS. The former arrangement provided a positive input signal and the present arrangement gives the required negative input signal.

This change is not an unreviewed safety question because:

(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eval-uated in the Final Safety Analysis Report has not been increased.

This change does not alter any functions of the RPS; it just corrects the as-built system to comply with the original design intent.

(2) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

(3) The margin of safety as defined in the basis for technical spec-ifications. has not been decreased.

This change does not represent a change in the facility as described in the Final Safety Analysis Repozt.

Page 56 Plant Change/Modification No. 54-76 PSL Unit Pl "RESTRICT ORIFICE IN CONTAINMENT PURGE SUCTION LINE" This change further'restricts the installed orifice in the Containment Purge Suction ductwork. System flow was originally set at design flow under dirty filter conditions. With clean filters, system flow is higher and creates a greater than desired vacuum in containment. The design dirty filter condition is considerably higher than the ac<<al AP's reached before normal filter replacement.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This system is not safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Further restriction of the original installed orifice enables this non safety related system to better meet original design intent. The filter will be changed well before the design (maximum hP) dirty filter condition is reached so there is no need for the excess flow.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This system is not discussed in the Technical Specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 57 Plant Change/fdodifioatlon 'go. 55-76 PSL Unit 81 "CHANGE SET PRESSURE ON RELIEF VALVE (RV) 5124" This relief valve is located -.on the pressurizer steam space sample line down-stream of the flow throttle valve and sample cooler. The normal flow rate in this 3/8 inch line created sufficient back pressure to exceed the con-servative valve set pressure. This change increased the set pressure 15 psi to prevent lifting the relief valve.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the FinaX Safety Analysis Report has not been increased.

2. The possibility for a'n accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

A,break of this line is already evaluated and design pressure of the line is higher than the new valve set pressure (90 psig).

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 58 Place Chaoge/Moddficacdoa No. 57-76 PSL Unit 81 "PERSONNEL AIR LOCK SEAL LEAK TEST ALARM MODIFICATION" This change modified the test unit and alarm circuitry so if the air 1ock door were opened during a leak test, the test would be terminated, pressure vented off and the alarm would not occur. Immediately venting the pressure increases seal life and eliminating spurious leak rate test failure alarms avoids unnecessary distraction of the operators.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

No airlock related accidents are discussed in the Final Safety Analysis Report.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments in (1) above. Also, the circuitry was tested for proper op-eration after the modification.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change meets present Technical Specification requirements.

This change does not represent a change to the facility as described in the Final Safety Analys'is Report.

Page 59 Plant Change/Modification 58-76 PSL Unit 81 "CHARGXNG AND LETDOWN FLOW TO DATA PROCESSOR" This change corrects the wiring for the charging and letdown flow signals to

'he data processor (for calorimetric calculations). The data processor required the b p signal directly from the flow. transmitters but previously was receiving its signal from the indication portion of the circuit. This potential problem was identified in 1975 but could not be resolved until full information on the data processor was received from the vendor in early 1976.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The monitoring instruments involved are not required to safely shutdown the plant and are not discussed in the Final Safety Analysis Report in relation to accident conditions.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The monitoring instruments involved are not required to safely shutdown the plant and are not discussed in the Facility Safety Analysis Report in relation to accident conditions.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 60 Plant Change/Modification 59-76 Unit 81 "WASTE CONCENTRATOR PRESSURE SWITCH REPLAC~NT" This change installs new pressure switches with adjustable differential ranges on the waste concentrator. The new switches will automatically start and stop the concentrator feed pumps on low and high levels re-spectively. The old switches would not do that properly due to their lack of adjustable differential range.

This change is not an unreviewe'd safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety px'eviously evaluated in, the Final Safety Analysis Report has not been increased.

The equipment involved is non-safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No functional changes are made; the system is modified to meet original design intent.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

C This change does not x'epresent a change to the facility as described in the Final Safety Analysis Repox't.

Page 61 Plant Change/Modification 60-76 PSL Unit Pl "C.E.A. DROPPED ROD CONTACTS FOR DATA PROCESSOR" This change xemoves )umpers in the CEDS cabinets for dropped rod signals to the data pzocessor for 4 CEA's. Previously the contacts for the 4 CEA's were in parallel with the data processor and it did not sense these 4 CEA's if they drop. Now the circuit is a series circuit and the data processor functions pzoperly. This was discovered during pre-critical preoperational testing.

This change is not an unreviewed safety question because:

1. The probability of occurzence or the consequences of an accident oz malfunction of equipment important to safety previously eval-uated in the Final Safety Analysis Report has not been increased.

This change does not altex any circuit functions; it gust corrects the "as-built" system to confirm to original design intent.

2. The possiblity for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created'.

See comments under (1} above.

3, The margin of safety as defined in .the basis, for technical specifi-cations has not been decreased.

This change does not repzesent a change to the facility described in the Final Safety Analysis Report.

Page 62 Plant Change/Modification 63-76 Unit 81 "HEAT TRACING CIRCUITS 'BPEDANCE CHANGE" 0

Four heat trac'ing circuits originally had less than 2 ohms impedance.

The heat tracing circuits'ontrollers will not operate properly be-low 2 ohms. This change modified the circuits to have greater than 2 ohms impedance while still providing virtually identical heating capacity.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in I the Final Safety Analysis Report has not been increased.

No function changes are made; the 4 circuits are modified to perform per the original design intent. Mi terials used are excess from the original supply used for heat tracing circuit.

This change will improve reliability for the affected 4 circuits.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See 1 Above.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 63 Plant Change/Modification 64-76 PSL Unit Pl "ADD ORIFICE UNIONS & ORIFICES TO CONTAINHENT INSTRPifENT AIR COifPRESSORS" This change adds orxfices (and replaces existing unions with special unions) to the compressor unloading lines. These rotary, water seal ring compressors spray water on each unloading (discharge back to suction) cycle and the orifice will prevent th's. As part of this PC/H, the nipple supporting the suction air filter/muffler was increased in length from 2 inches to 24 inches to prevent accumulated moisture from dripping out.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This is not a safety reLated system.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 64 Plane Change/Mcd151cac1cn Nc. 65-76 PSL Unit 81 "CHANGE I.V.M. PG!KR SUPPLY A1tD ADD PZBKNCIATION ON RTGBe102" This change supplies power to the reactor Internals Vibration Monitor panels fzom an uninterruptible supply and adds a constant voltage transformer to eliminate voltage fluctuations. Also, it adds a visible and audible alarm to alert the operators that the I.V.M. has detected an undesirable condition.

This change is not an unreviewed safety question because:

1. The pzobability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

No I.V.M. related accident is discussed in the Final Safety Analysis Repozt.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comment under (1) above. Also, this change will increase the reliability of the I.V.M. by eliminating unnecessary low voltage shutdowns.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 65 Plant Chango/Modifioatdon No. 66-76 PSL Unit 81 "POST-LOCA PAWL RESISTANCE T~ERATURE DETECTORS" Due to availability, 2 uncompensated RTD's were installed as an interim solution.

They would not calibrate to the desired degree of accuracy. Hew, compensated type RTD's were installed and properly calibrated by this PC/H, which is based on the Field Report which tracked/documented the original problem.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change removes the interim solution and restores the equipment to meet the original design accuracy.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facili,ty as described in the Final Safety Analysis Report.

Page 66 Plant Change/Modification 69-76,

.Unit 81 "PERSONNEL AIR LOCK INDICATOR LIGHTS" This change added warning lights by the interior and" exterior personnel air lock doors to 'indicate that an airlock doorairlock seal leak test is in progress or results were out of tolerance and doors should not be opened until the test is completed (light goes out) or problem is resolved. Prior to this modification, personnel approaching the airlock doors had no indication of seal tester status or of a possible alarm condition.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Design function of system has not changed- relay and indicating lights were added for indication only.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comment under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 67 Plant Change/Modification No. 70-76 PSL Unit 81 "COOLING WATER CANAL LEVEL AND TEMPERATURE IÃ)ICATION" This change adds discharge and intake canal level indication and adds a pre-alarm to installed discharge canal temperature instrumentation.

The discharge level will give operators indication to help avoid over-flowing the canal banks. The pre-alarm will give operators warning that condenser AT is approaching the limit given in the environmental Technical Specifications.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident oi malfunction, of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This circulating water canal level and temperature instrumentation is non safety related and is for indication/alarm only.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The new instruments give indicaiton and alarms only.

3. The margin of safety as defined in the basi's for technical specifications has not been decreased.

The temperature pre-alarm will aid in meeting environmental technical specifications but has no effects on the margin of safety.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 68 Plant Change/Hodification No. 72-76 PSL Unit 81 "WASTE GAS HEADER DRADl CONNECTXONS" This change installs. drain lines and valves at various low points in the waste gas header and adds slope to part of the header. This eliminates pockets in the header and allows draining of accumulated moisture from the header.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased..

The new drain connections and valves are designed and fabricated to the same as or better standards as the original header, and this system is not safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 69 Plant Change/Modification 74-76 PSL.Unit 81 ADD FLANGES IN CHARGING PPiiP SEAL WATER VENT LINES This change adds flanges in the vent lines so the seal water pump can be removed for maintenance w5.thout cutting and rewelding the vent line.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Failure of this line/change would not prevent operation of the seal lube system and in addition, the charging pumps can be operated without the seal lube system.

2. The possibility for'an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

There is no change in function of the seal lube water system or the vent lines.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as'escribed in the Final, Safety Analysis Report.

Page 70 Plant Change/Hoddficatlon No. 75 76- PSL Unit 81 "ADDTTIONAL WATER SEAL ON CCP SEAL LUBE WATER PUttP SHAFTS" This change installed an additional water seal on the shafts of the Coolant Charging Pumps'eal Lube Water Pumps. These pumps are a one-piece pump/motor unit and the one existing seal allowed moisture to enter the motor area caus-ing motor/motor bearing failure.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The seal lube water system is not safety related and, this change improves the reliability of the system.'.

The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This change does not affect any functions of the seal lube system and improves system reliability.

3. The margin of safety as'efined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 71 Plant Change/Modification 77-76 Unit 81 "PAST DEAD BUS TRAiRSFER FROM AECILIARY TO START-UP TRANSFORMERS" This change adds time delay relays to the two (A and B) transfer circuits which will prevent tz'ansfer of plant auxiliary loads from the Auxiliary to the Startup Transformers if greater than .17Transformers.

seconds (10 cycles) has, This will pre-elapsed since loss of power to the Auxiliary vent out of synchronization transfer of in-plant switchgear to the system grid.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in'the Final Safety Analysis Report has not been increased.

The involved circuits are non-safety related, non<<Class IE equipment. This change involves only power for non-vital auxiliary loads, and, this will improve overall reliability by preventing out of synchronization transfer which could result in equipment damage.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Pinal Safety Analysis Report has not been created.

Total loss of all off-site power is already evaluated.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 72 Plant Change/Modification 78-76 PSL Unit 81 "1C AUXILIARY FEEDWATER PUMP CONTROLS" This change modifies 2 contacts at one control station for 1C AFW pump (steam driven). Previously, when the 1B steam supply header was selected at the remote operating station, the pump shut down. The wiring was changed to agree with that at the main (Control Room) control station so the pump would shut down only when desired. This was discovered during hot functional testing before initial criticality.

This change is not an unreviewed safety question because:

1. The. probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final. Safety Analysis Report has not been increased.

This change does not alter any functions of the system; it just corrects the "as built" system to agree with the original design intent.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 73 Plant Change/Modification 80-76 Unit 81

'..'ACCUMJLATORS ON WASTE GAS COMPRESSOR DISCHARGE LINES" This change installs accumulators on the discharge lines of each waste gas compressor between compressor and discharge check valve. Formerly the check valves chattered at the end of each discharge stroke. The accumulators will prevent this, thus saving time, money and radiation exposure by reducing valve maintenance.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The accumulators are desi'gned, fabricated and installed to the same as or better specifications than the original equipment.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under (1) above.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

Page 74 Plant Change/Modification 81-76 PSL Unit 81 "MAIN STEAM ISOLATION VALVES AIR SUPPLY MODIFICATION" This change installed additional air accumulators connected to the existing ones for air supply to the HSIV's. The existing accumulators were adequate to hold the valves open for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (minimum) without supply air as described in the Final Safety Analysis Report. The new accumulators were added to ensure ability to close the MSIV's within technical specifications limits without supply air.

This change is not an unreviewed safety question because:

I

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

There are no changes of function and materials/installation criteria used were equal to existing design.

The consequences of accidents evaluated in the Final Safety Analysis Report remain the same or are decreased. The closure time of an MSIV is improved with this change.

2. The possibility for an accident or malfunction of a different type than any" evaluated previously in the Final Safety Analysis Report has not been created.

There are no changes of function and materials/installation criteria used were equal to exisint design.

The consequences of accidents evaluated in the Final Safety Analysis Report remain the same or are decreased. The closure time of an MSIV is improved with this change.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

There are no changes of function and materials/installation criteria used were equal to existing design.

The consequences of accidents evaluated in the Final Safety Analysis Report remain the same or are decreased. The closure time of an MSIV is improved with this change.

Page 75 Plant Change/Modfflcatdon Ão. 82-76 PSL Unit tl "STILLING TUBE MODIFICATION" This stilling tube is a sensing line for two level switches which give a low level alarm for the intake cooling pumps suction source (intake well). This change shortens the tube and relocates a bracket to prevent interference with the intake well trash rake.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-of equipment important to safety previously evaluated in the 'unction Final Safety Analysis Report has not been increased.

/

The tube still extends belo~ the low level alarm setpoint and the instru-ment gives only an alarm - there are no control features associated with these instruments.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This change does not change any function of the involved instruments.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

The tube still extends below the required low level alarm point.

This change does not represent a change to the facility as describ'ed in the Final Safety Analysis Report.

Page 76 Plant Change/Modification No. 84-76 PSL Unit 81 "COMPONENT COOLING WATER CHEMICAL ADDITION TANK DRAIN LINE" This change reroutes the CCW Chemical Addition tank drain line to the chemical drain system. This will aid in waste treatment, avoid the health hazard exposed chromated water poses and prevent contamination or exposure problems should CCW become contaminated.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. Neither the tank nor the drain line is safety related. The drain comes from thh CCW chemical addition tank through an originally installed valve and then the drain line is rerouted from a sump to the chemical drain system.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report, has not been created.

The CCW chemical addition system is non safety related.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

The chemical addition system is not discussed in the Technical Specifica-tions.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 77 Plant Change/Modification 85-76 PSL Unit /Il "MODIFICATION OF TUBING SUPPORTS ON HOT LEG lA SAMPLE LINE" This change modified the tubing supports for this sample line. The original supports were sliding collars which had an inside diameter slightly too small to allow the tubing to slide when it expanded due to plant heatup. The new supports allow free thermal expansion but still provide seismic restraint.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change does not alter the functions, numbers or locations of the tubing supports; it corrects installation to allow full thermal growth.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

,This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 78 Plant Change/Modification No. 88-76 PSL Unit 81

'MAIN STEAM ISOLATION VALVES BYPASS VALVES CONTROL CIRCUITS" This change bypasses the seal-in feature and the limit switch lockout feature.

Previously, if the valves were open about 5X or less (to warm up steam lines),

they could not be (electrically) closed without first going to greater than 5X open.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety AnaIysis Report has not been increased.

This change enhances the reliability of bypass valve closure under all conditions including Main Steam Isolation Signal.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Pinal Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does,not represent a change to the facility as described in the Final Safety Analysis Report.

Page 79 Plant Change/Modification No. 89-76 PSL Unit 81 "COLLAR ON PIPE PENETRATING THE CONTAINMENT SUMP SCREEN" One pipe experienced sufficient thermal growth to cause greater than the maxi-mum 1/2 inch gap where it penetrated the screen entering and exiting the sump area. This change added a stiff wire mesh collar outside the present screen to cover this gap.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Foal Safety Analysis Report has not been increased.

The decrease in screen flow area is negligible (less than 1/2Z) and the collars are located outside the present screen.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The collars are added to ensure we meet the requirements of the Final Safety Analysis Report (no gap in containment sump screen greater "than 1/2 inch).

3. The margin of safety as defined in the basis for technical specifica-tions has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 80 Plant Changehfodif ication No. 97-76 PSL Unit ftl "SMOKE AND HEAT (FIRE) DETECTION SYSTEM AUDIBLE ALARM" The original alarm for the fire detection system was not loud enough to be heard easily over normal control room noise levels. This change added a louder alarm on the system console to replace the original fire detection console alarm.

This change is not an unreviewed safety question because:

l. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

I This system is not nuclear safety related and the change does not alter any functions; it replaces the original alarm with a louder one. This ensures the system meets the original design intent of notifying the operator of any smoke/heat detector alarms.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under 1 above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 81 Plant Change/Modification No. 101-76 PSL Unit 81 "POWER OPERATED RELIEF VALVE RELAY COILS" The original relay coils were rated at 115-125 Volts DC which is unit battery voltage. However, normally the battery chargers are in operation to ensure the batteries remain fully charged and system voltage is about 135V D.C. The new relay coils are rated for continuous operation in the higher voltage.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction. of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The new relays were purchased to the same as or better specifications than the original relays and will enhance system reliability by preventing coil burnup.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This change does not alter any design intents or functions.

3. The margin of safety as defined in the basis zor technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 82 Plane Change/Modfffcacfon go. 102-76 PSL Unit 81 "INSTALLATION OF HYDRAULIC SNUBBER TESTER" Technical Specifications require periodic performance testing of hydraulic snubbers. This change allows installation of the testing machine in the Reactor Auxiliary Building.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

I The machine is not itself safety related and it is not located in the vicinity of any safety related equipment.

2. . The possibility for an accident or malfunction .of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Neither the machine nor the area in which it is located are discussed in the Technical Specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 83 Plant Change/Modification 103-76 PSL Uni.t 81 0

"REMOVAL OF CYCLE T'DIE WIRE IN CEDM CPP'S" This change removed a wire from the CEDM Coil Power Programmers which had been acting as an antenna, picking up electronic noise and causing/con-tributing to inadvertent rod drops. The wire involved ran from a connector terminal to another terminal and was not connected to any other part of the circuits. It originally had been for monitoring the cycle time signal for test purposes but this function previously was deleted from plant design.

I change is not an unreviewed safety question because:

This

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased'.

I The wire does not have any functional purpose and the design intent of the circuit and system are not changed.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This change reduces the probability for dropped rods and cannot create any new accidents.

3. The margin of safety as defined in the basis for technical specifi.cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 84 Plant Change/Modification 104-76 Unit 81 REPLACE ~+AYS IN CONTROL ROOM OUTSIDE AIR INTAKE RADIATION MONITORS The original relays vere 120 VAC relays modified for use in a DC circuit.

The vendor has informed us this is not suitable for.1.ong term use. New, DC relays have been installed per this PC/H.

This change is not an unrevieved safety question. because:

'. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Pinal Safety Analysis Report has not been increased.

This change alters no functions; it corrects the as built non-safety related system to meet original design intent/function.

/

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comment under l. above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

E This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 85 Plant Change/Modification No. 109-76 PSL Unit Pl "MODIFY FILTER OUTLET PIPES" This change shortens the outlet standpipes of the CVCS, fuel pool and waste management filters by 3/4 inch. This is done to accommodate the "throwaway" filter/cage assemblies 'previously approved {before licensing) as a means of significantly reducing personnel radiation exposure.

This change is not an unreviewed safety question because:

l. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change does not alter any functions of the affected systems.

2.. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This is not a functional change and does not affect any filter performance monitoring instrumentation.

3. The margin .of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change, to the facility as described in the Final Safety Analysis Report.

Page 86 Plant Change/Modification No. 112-76

~

PSL Unit !Il "CONTROL ROOM AIR CONDITIONING THERlQJ EXPANSION VALVE CAPACITY REDUCTION" This change installs new internals of lower capacity in the (freon) thermal expansion valves in the control room air conditioning system.

The original larger valves cycled excessively and due to low freon velocity appeared to allow compressor oil "hideout" in the system.

The smaller valves will cycle less, promoting greater stability and create higher velocity freon flow to avoid oil "hideout".

This change is not an unreviewed safety question because:

1. The probability of'ccurrence or the consequences of an accident

. or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. "

The new internals are a modification kit designed and fabricated by the vendor of the original valves. And, failure of one (of 3) air conditioning units is evaluated. The greater stability of operation will improve system reliability.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Complete failure of one air conditioning unit is already considered'n the Technical Specificatj,ons.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 87 Plant Change/Modification 113-76 Unit 81

'DD VIBRATION RESTRAINT TO CHARGING PUMP SUCTION LINE"

'his change added a vibration restraint with snubber to the charging pump suction line. This avoids the possibility of long term operation'ith slight vibration causing damage to the pipe.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident. or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The'luid boundary of the system is not changed. The restraint is designed, built and installed to Seismic Class I standards as good as or better than original specifications.

'; The possibility for than any evaluated an accident or malfunction of a different type previously in the Final Safety Analysis Report.

has not been created.

This change reduces the possibility of a piping failure.

3. 'The margin of safety as defined in the basis for technical specifications has not been decreased.

The fluid boundary of the system and the piping configurations are not changed.

This change does not represent a change to the facility as described in Final Safety Analysis Report.

Page 88 Plant Change/Modification 114-76 PSL Unit 81 "EXCORE NUCLEAR LNSTRUMENTATION LINEAR AMPLIFIER GAIN CHANGE" This change increased the gain of the subchannel 3 (upper detectors) linear amplifiers for Channels 8, C and /$ 10 (Control Channel 82) of the Power Range Linear Channels. These amplifiers were designed and built with the flexibility to change the gain if needed to accomplish subchannel calibration and it was done by moving two wires to different terminals on the amplifier cards. The need for this change was discovered during power ascension (at 20% power) while testing and calibration were in progress.

I This change is not an unreviewed safety question because:

1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eval-uated in the Final Safety Analysis Report has not been increased.

This change takes advantage of the designed flexibility of the system and should be considered a normal calibration adjustment.

2) The possibility for an accident of malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This change does not change the function of any part of the system but simply changes the amplifier gain much as does the installed fine adjustment/calibration potentiometer.

3) The margin of safety as defined in the basis for technical spec-ifications. has not'een decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 89 Plant Change/Modification 116-76 PSL Unit Pl "REACTOR PROTECTION. SYSTEM AT PONEP. NOISE RH)UCTION" This change moved the cable shield ground from the instrument ground to the loop transmitter common (-). 'his gave a significant reduction in 60 cycle electronic noise in the circuits.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

I This change did not alter any functions; it gust rerouted the ground connection.

2. The probability of occurrence or the consequences of an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Instrument failure is already addressed.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 90 Plant Change/Modification 117-76 Unit g1 "INSTALL SNUBBERS ON AP INSTRUCTS FOR INTAKE COOLING WATER STRAINERS" This change installed snubbers on the sensing lines to the tZ instruments.

Previously pressure surges in the lines caused spurious flow alarms even though flow was normal.

This change is not an unreviewed safety question because:

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The snubbers (pulsation dampers) were purchased to the same as or better specifications than the original components.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No design intents or functions were altered. Addition of these pass-ive components will improve overall system reliability by eliminating spurious alarms and reducing effects of pressure pulsations on the in-struments.

3. The ma gin of safety as defined in the basis for Technical Specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 91 Plant Change/Modification No. 118-76 PSL Unit 81

'MODIFICATION TO SEIScfIC RESTRAINTS FOR 1"SAFETY INJECTION LINES" In post core load hot functional testing, it was discovered that restraints for 2 lines (1" SI-120 and 1" SI-237) did not allow for the full thermal gzowth ex-perienced by the lines. This change modified 3 restraints to allow full therm-al growth while still providing Seismic Class I support.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety I

Analysis Report has not been increased.

This change reduces the probability of an (already analyzed) piping failure and does not increase the consequences.

2. The possibility for an accident or malfunction of a diffezent type than any evaluated previously in the Final Safety Analysis Report has not been czeat-ed.

Components used weze designed and built to the same as or better specifica" tions than the original restzaints.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

See comments under (1) and (2) above.

This change does not represent a change to the facility as described in the Final Safety Analysis Report. 'I

Page 92 PLANT CHANGHfHODIPICAIIOH HO. II9-76 PSL UNIT 81 "MODIFICATION TO SEISMIC RESTRAINT FOR BLOtKOWN VALVE FCV-23-4" In post core load hot functional testing it was discovered that the restraint for steam generator blowdown valve FCV-23-4 did not allow for the full thermal growth experienced by the valve/line. 'his change modified the restraint to allo~ full thermal growth while still provid-ing Seismic Class I support.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evalu-ated in the Final Safety Analysis Report has not been increased.

This change reduces the probability of a piping/valve failure already analyzed in the Final Safety Analysis Report and does not increase the consequences.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Components used were designed and built to the same as or better standards than the original restraint.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

See comments under (1) and (2) above.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 93 Plant Change/Modification No. 120-76 PSL Unit 81 DIGITAL DATA PROCESSOR-'OVEABLE IHCORE DETECTOR SYSTEMS IÃZERFACE'"

This change makes wiring modifications and adds a depth encoder driver (amplifier) so the incore detector system can properly feed signals to the DDPS. The DDPS controls the incore system and provides the informa-tion from that system to the operators. The two systems are from different vendors and were not entirely compatible.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This does not alter any functions of either system; it simply corrects an interface problem so these two non safety related systems will function per original design intent.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not zepresent a change to the facility as described in the Final Safety Analysis Report.

Page 94 Plant Change/Modification No. 121-76 7SL Unit 81 "EMERGENCY CORE COOLING SYSTEM AREA LOW VACUUM AL~"

The low vacuum alarm was wf.red so that it was armed at all times. Since the full. which maintain the required vacuum are required to run only under accident fans conditions (auto-start by Safety Injection Actuation Signal) this resulted in many spurious alarms. The alarm was rewired to be armed only when there is a Safety Injection Actuation Signal which is the only time this alarm is meaning-This change is not an unreviewed safety question because:

,1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Reduction in spurious alarms will improve operator response to an actual alarm thus improving overall'ystem operation. This alarm circuit does not affect fan operation or actual vacuum in the ECCS area.

2. The possibility "for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The new relay used meets at least the same qualifications as the original components.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This alarm is not discussed in the Technical Specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 95 Plant Change/Modification No. 122-76 PSL Unit Pl "LIMITORQUE CONTROL CIRCUIT MODIFICATION" This change removed limit switch contacts in the. control circuits which were intended to prevent valve chatter on closure. However the contacts also prevented the valves from closing if they were open 5X or less (without first opening them to >57). The vendor stated that the operators involved would not chatter. The valves involved were the High Pressure Safety Infection pump discharge valves, High Pressure Safety Injection and Low Pressure Safety In]ection header isolation valves and auxiliary feed pump discharge valves.

This change is not an unreviewed safety question because:

1. The'probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change corrects the system to meet original design intent and improves the reliability of the involved valves.

'2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No functions were changed and valve reliability was improved.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page '96 Plant Change/Modification No. 123-76 PSL Unit Pl "COOLANT CHARGING PUMP PACKING Ai%) SEAL MODIFICATION" This change reduces the number of primary packing rings, installs a spring loaded bushing to retain packing instead of only a spring, replaces a metal packing adapter with a closer tolerance non-metallic adaptor and changes the secondary packing design (cross-section). These changes (already done at another plant with similar CCP's) will solve the problem of extremely short packing life presently being experienced.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This will improve operation/reliabil'ity of the CCP's. The new design is equal to or better than the original design and has been approved by both the pump vendor and the Nuclear Steam Supply System vendor.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No pump functions are affected by this change.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 97 Plant Change/Modification No. 124-76 PSL Unit 81 "REVERSE POLARITY - POWER DEPENDENT INSERTION LIMIT" This change reversed two leads in the Reactor Protective System Channel "D" so the PDIL would increase as power was increased. As previously reported in Licensee Event Report 335-76-20, the PDIL was decreasing as power was increased.

This change is not an unreviewed safety question because:

1. The'probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change does not affect any functions of the circuit; it gust corrects the "as-built" input polarity to meet original design intent.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comment under 1 above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

See comment under 1 above.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 98 Plant Change/Modification 126-76 Unit 81 "CHANGE CABLE FOR VIDE RANGE NI CHANNEL C" This change corrects an erroneous cable pulling card and installs- the specified type of cable from the electrical penetration room to the Control Room.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an a'ccident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been in'creased.

Installation of the specified cable changes the channel to meet original design intent and reduces the probability of (already ana'lyzed) cable/channel failure.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

See comments under (1) above.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 99 PLANZ CHANGE/MODIPICAIION NO. 129-76 PSL UNIT 81 "CHARGING PUMP HIGH LEVEL CUTOUT BYPASS" This change installs a key-operated bypass switch to allow running more than one charging pump with high pressurizer level. This will allow better control of pressurizer level during cooldown and more expeditious filling of pressurizer when taking the plant solid and filling the drained pressurizer after maintenance.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evalu-

~ ated in the Final Safety Analysis Report has not been increased.

Strict administrative controls will prevent use other than when specified above and when pressurizer level goes above 100/ (top of indicating range).

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Per original design we could run at least one charging pump at all times.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 100 Plant Change/Modification No. 130-76 PSL Unit 81 "EXCORE NUCLEAR INSTRUMENTATION AUDIO CIRCUIT NOISE" The audio count rate circuit was the source of electronic noise in the Reactor Protective System circuits when switching ranges. This change installed toggle switches to eliminate the noise and prevent more spurious plant trips.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change provides additional separation between Nuclear Instrumentation and Reactor Protective System channels.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The functions of the affected systems remain exactly the same and, no new functions have been added.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 101 Plant Change/Modification No. 134-76 .

PSL Unit 81 INTAKE COOLING MATER PUB@ INSTRUMKZ CABLE REROUTING This change reroutes one cable to another conduit so the original conduit is available to support electrical services to the Steam Generator Blow-

'down Treatment Building.

This change is not an unreviewed safety question because:

1. The probability of occrrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The rerouted cable is a pump pressure indication instrument cable and is not evaluated in the safety analysis. It is rerouted in a proper fashion through safety class conduit and does not reduce redundancy.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under 1 above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facil'ity as described in the Final Safety Analysis Report.

Page 102 Plant Change/Modification 135-76 Unit 81 "RESLOPE STEAM GENERATOR LEVEL REFERENCE LEG PIPING" This change reslopes the reference leg piping for the level transmitters so they will slope downward toward the steam generators when the plant is at hot operating conditions. This removes a low point which could retard circulation and prevent proper operation of the condensate pots and there-by affect indicated steam generator level.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the/ Final Safety, Analysis Report has not been increased.

This change increases overall reliability of the Steam Generator level indication system.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No design functions or intents are changed; this change improves the implementation of the original design intent.

3. The margin of safety as defined in the basis for Technical Specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 103 Plant Change/Modification No. 137-76 PSL Unit 81 S

"EKCORE NUCLEAR INSTRUMENTATION LINEAR POWER RANGE DRAWER MOD IP ICATION" This change replaces a resistor with one of different rating and replaces a potentiometer with a more sensitive vernier drive potentiometer to obtain greater resolution and control of voltage input to the amplifiers of the drawers. It also adds a switch as part of the potentiometer which completely removes the trip test pot from the circuit when the test pot is,"off". This removes a residual signal which formerly was applied to the circuit.

'This change is not an unreviewed safety question because:

I

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Pinal Safety Analysis Report has not been increased.

No functions have been added or deleted and the design intent is not altered. This change improves the implementation of the original design intent by separating the test signal from the actual signal and improving the resolution of the amplifier gain adjust.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under 1 above.

3. The margin of safetyas defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 104 PLANT CNANGEIMODIPICATION NO. 146-76 PSL UNIT 81 "EXTRA REACTOR HEAD CABLE TRAYS" This change adds four (4) temporary (refueling use only) cable trays to be mounted above the existing trays during refueling outages. The reactor head cables (instrument and CEDE power and position indication cables) can be folded up into the new trays after being disconnected from the head. This storage area will prevent tangling of cables (formerly folded back into the original trays) and damage to the cable connectors.

This change is not an unreviewed safety question because:

1. ~ The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been in-creased.

The cables are not in use during refueling when the new trays are in use. The new trays will be removed when the head cables are reconnected.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The added trays simply hold the cables as did the original trays during refueling when the cables are not in use. Same trays and supports are used as in the original design.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 105 PLANT CNANGN/NODTPTCATZON NO. 147-76 PSL UNIT 81 "MODIFICATION OF FUEL POOL PURIFICATION LOOP SIPHON BREAKER" This change temporarily removed the plug installed in the siphon breaker per PC/M 43-76 to allow storage of irradiated fuel in the fuel pool for replacement of poison pins in th'e assemblies (see PCM 176-76).

After the irradiated fuel was removed from the fuel pool, the plug was reinstalled; This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident

, or malfunction of equipment important to safety previously evalu-ated in the Final Safety Analysis Report has not been increased.

Removing the plug returns the fuel pool to original design con-ditions for storage of irradiated fuel. When the fuel is removed, the siphon breaker or the fuel pool are not required.

The plug was reinstalled after fuel was removed (See 3. below).

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The removal of the siphon breaker is administratively controlled per the plant backfit list. Also see PC/M 43-76.'.

The margin of safety as defined in the basis for technical speci-fications has not been decreased. The siphon breaker is required by the Technical Specifications when spent fuel is stored in the fuel pool. The plug is required during plant operation to provide an interim source of tornado protected makeup water per the PSL Unit 1 operating license. The siphon breaker should prevent pumping the water out if it were not plugged.

Page 106 Plant Change/Modification 157-76 PSL Unit 81 "FUEL TRANSFER TUBE SHIELDING"

'I" This change added concrete radiation sh'elding to the north side of the fuel transfer tube between the Reactor Containment and Fuel Handling Buildings. The need for this shielding was identified before licensing and added to the backfit list. It was advanced in schedule to allow defueling of the reactor due to the power distribution anomaly. (See PC/M's 176 and 192.)

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equpment important to safety previously evaluated Xn the Final Safety Analysis Report has not been increased.

This Seismic Class I shielding (passive component) will reduce radiation exposure<< during spent fuel transfers and does not alter the design or configuration of the related structures.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comment under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 107 PIANT CNANGN/MODI."ICATION NO. 158-76 PSL UNIT 81 "PIPE HANGER MODIFICATION" This change'odified five (5) hangers on two (2) non-seismic but safety related (Category 2 & 3) lines. This was done to provide clearance for installation of steam generator blowdown lines to the Steam Generator Blowdown Treatment Facility. The affected lines were Reactor Coolant Pump Seal Bleedoff to the Volume Control Tank and a Fuel Pool Purification line.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident

, or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been in-creased.

The modified pipe support hangers are made of the same materials, to the same specifications as were the originals. The routing of the lines is not changed and no functional changes are in-volved. The original hanger attachment to the building is used.

The hangers were not and are not seismic.

2. The possibility for an accident or malfunction of a different-type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under 1, above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 108 Plant Change/Modification No. 159"76 PSL Unit 81 ADD CHECK VALVE IN BORIC ACID HOLDING SYSTEti The change added a check valve in a line from the Boric Acid holding tank downstream of a relief valve which discharges to the holding tank. During certain evo'utions, thy line was pressurized and leakage past the relief diluted the holdup tank. The check valve will prevent the line and the re-lief from being pressurized.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The line is non-safety related and complete failure of the line would result only in partial loss of feed flow to the non-safety related boric acid concentrators.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See Comments under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

The line/related system is not discussed in the technical specifications.

Page 109 PLAHT CHAHGE/MODZEXCATTOH HO. 160-76 PSL UNIT f/1 "HIGH PRESSURE AND LOW PRESSURE SAFETY INJECTION HEADER ISOLATION VALVES POSITION INDICATION" This change moves a resistor from one leg to another in the position indication circuitry. This was done as the original circuit would not calibrate properly to meet the desired accuracy for the % open indicator. This indicator is one of three indications in the control room for valve position and valve position can be determined at the valve itself.

This change is not an unreviewed safety question because:

/

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluat'ed in the Final Safety Analysis Report has not been in-creased.

No functions are changed. The % open meters are for operator information only. The design intent and function remain the same as the original design.

2. The possibility for an accident or malfunction of a different .

type than any evaluated previously in the Final Safety Analysis Report has not been created.

No new components or functions are added. This change simply ensures the indicators will calibrate to the desired accuracy.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

A

(

Page 110 PLANT CHANGE/MODIFICATION NO. 161-76, PSL UNIT /$ 1 "ADD FUEL OIL SOlPiiDING TAPES AT EMERGENCY DIESEL FUEL OIL TANKS" 0

This change adds sounding tapes to the diesel fuel oil tanks. This is primarily.a company policy regarding inventory control of fuel oil but also provides a back-up level indication to help ensure proper reserves of diesel fuel oil are maintained per the technical speci-fications.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident

,or malfunction of equipment inportant to safety previously eval-uated in the Final Safety Analysis Report has not been increased.

The tapes are mounted on the manhole cover plates on the top of the tanks and do not affect tank integrity.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This provides a back-up indication to ensure technical specifications on diesel fuel oil reserves are met.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.'

Page 111 Plant Change/Modification 163-76 Unit Pl "ADDITION OF BARRIER I'N TWO CABLE PULL BOXES" This change added a barrier in 2 cable pull boxes to mai'ntain minimum separation requirements between safety related and non-safety related cable.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change alters no design functions or intents; it corrects 2 cable pull boxes to meet the original design intent.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under l. above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 112 Plant Change/Modification No. 164-76 PSL Unit 81 "INSTALLATION OF DRIP PANS OVER TSP BASKETS" This change adds drip pans over the trisodium phosphate dissolving baskets located in containment for post-Loca coolant pH control.- The;pans will pre-vent condensing moisture from process lines from reaching and partially dis-solving the TSP in the baskets. This will aid in ensuring proper amounts of TSP are maintained.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The implementation of this PC/M deals with an item which in itself is not required for safe shutdown of the plant, thus it will not increase the probability of occurrence of any accident. The drip pans and all supports are Seismic Class I and will not come loose in the event of an earthquake or a LOCA and thus cannot cause blockage of the containment pump screens.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change improves the margin of safety by ensuring the proper inventory of TSP can be maintained.

This change does not represent a change in the facility as described in the Final Safety Analysis Report.

Page ll3 Plant Change/Modification No. 166-76 PSL Unit 81 INSTRUMENT AIR TIE-IN FOR STEAM GENERATOR BLOVDOWN TREATMENT FACILITY'his change adds the line and valve to allow supplying instrument air from the turbine building to the blowdown treatment facility.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Instrument air is non-safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

Instrument air is non-safety related.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change in the facility as described in the Final Safety Analysis Report.

Page 114 Plant Change/Modification 168-76 Unit dl "PRESSURIZER SPRAY VALVE SEAT MODIFICATION" This change provides'new, interference fit seats and seat retainers for fit the spray valves. This will be aligned by the interference and seal welded in place. This is recommended by the valve vendor as the original screwed-in retainer can loosen and allow leakage past the seat.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

/

No functional or design changes are made and this modification will improve valve performance by eliminating the possibility of leak-age bypassing the seat.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The new seat and retainer ring are provided by the original vendor to the same as or better specifications as the original parts.

3. The margin of safety as defined in the basis for technical specifi-cations has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 115 Plant Change/Modification 167-76 Unit 81 "MODIFY REACTOR HEAD CABLE JUNCTION BOXES" (REFUELING DISCONNECT BOXES)

This change enlarged the opening into the head cable )unction boxes.

The original openings were undesirably small. They were nearly filled, thus restricting access to hook up the connectors and their size/location required some cables to make two sharp curves to mate the connectors. This resulted in excessive time for connecting/dis-connecting the cables and subjected the cables to undesirable risk of damage.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eval-uated in the Final Safety Analysis Report has not been increased.

This change does not alter any function or design intent of these, cables.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

This change makes the systems supplied by these cables more reliable in that it improves access and makes it possible to better insure proper connection without damage or undue stress to the cables or their connectors.

3. 'The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 116 Plant Change/Modification No. 172-76 PSL Unit 81 "ADD NEW GAIN RANGE TO LINEAR POWER RANGE NUCLEAR LfSTRlPiiENTATION DRAWERS" As reported for PC/M 114-76, 3 drawers required gain increases (option was designed into the amplifiers) for proper output. This change installs a new gain range between the two options supplied by the manufacturer.

This allows all drawers to be wired the same and provides proper gain span for all the drawers.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in Final Safety Analysis Report has not been increased. 'he There are no new functions or design intents. This change adds and uses another gain span between (and partially overlapping) the two options (ranges) provided by the manufacturer. This will still provide proper gain for all drawers (as now exists) and allows all drawers to be wired the same which eliminates a possible source of confusion to the personnel calibrating and maintaining the drawers. The new resistors and capacitors used were purchased to the same as or better specifications as the original ones.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 117 Plant Change/Modification No. 173-76 PSL Unit 81 "CORRECTION OF HYDRAULIC SNUBBERS OVERFILL HOLE LOCATION" This change returned 7 snubbers to the vendor for relocation of the overfill (weep) holes. These 7 (of many at PSL 81) had the holes mislocated so that hydraulic fluid inventory was reduced which would require frequent inspection and/or refill thus affecting plant availability.

This change is not an unreviewed safety qui stion because:

1. 'he probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been iztcreased.

This change affects no function or design intent; it just corrects the snubbers to meet their original design intent. The mislocation deprived the snubbers of fluid margin which could affect plant availability due to the Technical Specification requirements.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under 1 above.

3. The margin of safety as defined in the basis for-technical specifications has not been decreased.

This change improves plant ability to meet the Technical Specifications without affecting plant availability.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 118 Plant Change/Modification No. 176-76 PSL Unit 81 "REPLACE PSL CYCLE 1 FUEL POISON PINS" This change replaced (reconstituted) the boron carbide poison pins in 108 fuel assemblies for the St. Lucie Unit 1, cycle 1 core. The original pins had excessive internal moisture content, resulting in hydride corrosion and perforation of the cladding. The flow plate was cut to allow replacement with new poison pins essentially identical to the originals. A retention grid was placed over the holes and mechanically fastened (crimped) in place. The work was done by vendor technicians, in the PSL spent fuel pool, under vendor and Florida Power & Light supervision, with final QC inspection by both vendor and Florida Power & Light personnel.

I This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The temporary equipment used was designed/selected and load tested to ensure that the probability of a fuel handling accident due to equipment failure was not increased. Although the number of fuel transfers over the life of the plant has not been quantified, it should be noted that the fuel transfers required for repair are the same as could be required by fuel inspections which are implicit in the original design of the spent fuel storage facilities.

Also, analysis was performed to verify that transfers/rework could not result in inadvertent criticality even if juxtaposition of up to 4 assemblies in the borated spent fuel pool should occur. For this evolution surveillance was established to verify boron concentra-tion was maintained. Analysis of the results of severe fuel assembly damage was performed and proven to be within the bounds of the FSAR analysis of a fuel handling accident.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The FSAR analysis of damage to fuel assembly fuel pin cladding does not address the mechanism of damage but only addresses the results of such damage. Analysis of the results of severe fuel assembly damage was performed and proven to be within the bounds of the FSAR analysis of a fuel handling accident.

Page 119 Plant Change/Modification No. 176-76 (cont.)

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

A review of the basis for technical specification indicated that no reduction in margin of safety would result due to the minimal exposure (irradiation) time; the maximum power level of 80%

obtained prior to shutdown; and the fuel decay time of 76 days prior to commencement of reconstitution.

NOTE: This PC/M covered the actual work of replacement. See PC/M 192-76 for analysis of the return to critical operation using the modified fuel. Also, see CEf-38 Rev. 0 and Rev. 1 previously submitted to the NRC,

Page 120 Plant Change/Modification No. 177-76 PSL Unit 81 "ANNUNCIATE 15Z PER HOUR POWER CHANCE" This change installed connections and modified the Digital Data Processor System to annunciate on the RTGB any power change of 15/ per hour or, more.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

Change adds an indication (alarm) function only and does not affect any DDPS functions or accidents.

2. The possibility for an accident or malfunction of a different type than any evaluated previously 9n the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change provides indication to help alert the operators to take tech-nical specification required action on a power change of 15% or greater per hour, thus increasing the margin of safety.

This change does not represent a change in the facility as described in the Final Safety Analysis Report.

Page 121 Plant Change/Modification No. 181-76 PSL Unit 81 "REVISE CONTROL SCHICK FOR VOLUME CONTROL TAiK LEVEL CONTROL VALVES" This change provides a new controller with adjustable deadband for controlling valves V-2501 and V-2504. The original controller functioned but caused the valves to cycle open and shut continuously, resulting in eventual damage to the valve operator motors.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

There is no change in design intent or function. System reliability is improved as the valves will not stroke continuously and be damaged. This portion of the Chemical and Volume Control System is non-safety related.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The new controller was purchased to the same as or better specifications than the original.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 122 Plant Change/Modification No. 182-76 PSL Unit 81 "SHUTDOWN COOLING RELIEF VALVE MODIFICATIONS" Due to problems experienced (lifting below setpoint and excess blowdown) these two valves were modified as follows:

1. Repositioned so the valve stem/disc assemblies are in the vertical plane

2. Revamp the inlet piping to provide a direct flowpath of larger pipe size I

3. Install expansion loops in the outlet piping to reduce stresses if one header is operating (hot) and the other is shutdown (cold).

4. Reset the blowdown from 25X to 10Z.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

No design functions or intents are changed and the valves/system will be more reliable (Refer to LER 335-76-40 dated August 18, 1976). The revised piping was designed and fabricated to the same specifications as the original.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under 1 above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 123 Plant Change/Modification No. 183<<76 PSL Unit 81 "LETDOMN BACKPRESSURE CONTROL ifODIFICATION" This change adds a lead/lag unit to the Chemical and Volume Control System backpressure control valves and an adder/subtractor unit to the letdown control valves. The adder/subtractor unit will slightly delay opening of the letdown valves to allow more time for the backpressure valves to respond. The lead/lag unit will cause the backpxessure control valves to anticipate changes in flow from the letdown valves based on the pressurizer level error signal. This will prevent the pressure surges which have lifted the relief valve downstream of the backpressure control valves.

'll This change is not an unreviewed safety question because:

1.. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This does not change any .design intent or functions but merely coordinates the contxols of the two valves to provide smoother, more integrated functioning of the system and impx'ove reliability.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report, has not been created.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 124 Plant Change/Modification No. 185-76 PSL Unit 81 "EXTEND LEAK TEST CONNECTIONS FOR FUEL TRANSFER TUBE PAST THE SHIELDING" This change adds tubing to extend the local leak rate test connections

'to the outside of the radiation shielding. This allows testing the fuel transfer tube nozzle seals without removing the radiation shielding ~

for access.

This change is not an unreviewed safety question because:

l. The probability of occurrence or the consequences of an accident, or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

No design functions or intents are changed and the two extensions are designed and,fabricated to the same specifications as the original connections. The connections remain within containment and the shield building annulus which are protected, filtered release areas.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.,

See comments under 1 above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change significantly reduces the work and radiation exposure necessaiy for the leak rate testing required by the Technical Specifications.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

0 Page 125 PLANT CHANGE/MODIFICATION NO. 190-76 PSL UNIT 81 "LIQUID WASTE DEMINERALIZER SYSTEM'his change installed additional liquid waste ion exchangers to supplement the installed waste ion exchanger. A small demineralizer was added in parallel to the original one to allow resin replacement or maintenance without stopping waste processing. Two "polishing" demineralizers (in parallel) were added to the system to increase the decontamination fac-tor. This allows use of the system at higher inlet activities than was previously possible.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

The new components are equivalent to the original equipment. This system is non-safety related. .Liquid waste release to the environ-ment is a manually control1ed and monitored function independent of the means used for in-plant processing and this change will not affect the FSAR or Technical Specification Environmental Sections.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See comments under 1 above.

3. The margin of safety as defined in the basis for Technical Specifications has not been decreased, See comments under 1 above.

Page 126 Plant Change/Modification 192-76 Unit 81 "RETURN TO POWER USING RECONSTITUTED FUEL" This change presents the safety analysis for using, at power, the fuel modified by PC/M 176-76. Briefly, that modification consisted of: cutting webs/dzilling holes in the fuel assembly flow plate; re-moving 12 poison pins and replacing with new ones; and installing a retention grid over the holes.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an acci-dent or malfunction of equipment impoztant to safety previously evaluated in the Final Safety Analysis Report has not been in-creased.

, The only changes were to the poison pins and the upper flow plate. The fuel pins were not affected and the fuel as-sembly alignment was checked (gaged) to verify it was not affected.

The poison pins were changed only as follows:

A. Internal moisture content was reduced-the cause of the perfozations in the "cladding was hydride corrosion due to excessive moisture content.

Reduction in moisture will improve reliability by removing the prerequisite for this corrosion mechanism, thus minimizing probabilities oz recurrence of the boron 1oss/redistribution and resulting flux anomalies.

B. The nominal OD of the pins was increased .0045 inches - this has been evaluated as having no significant effect on flow characteristics or temperatures within the pins.

C. To facilitate handling the upper end cap of the new poison pins,,was modified this has no effect on in-core performance.

D. The lower end caps of the new poison pins were modified to ensure pzoper retention in the re-tention grid - this was done to ensure down against up>>lift forces such as would proper'old be experienced in a postulated LOCA.

E. The upper end plenum spring has bien changed to the vendor's cuzrent design for similar (14 x 14) fuel assemblies - this has been evaluated as hav-ing no adverse effects on in-coz'e performance.

Page 127 Plant Change/Modification 192-76 (Cont'd) Unit 81

1. (Cont'd)

The same materials were used as in the original pins and poison loading was the same as the original "as built" poison pins. Due to the low depletion of the fuel this loading will have no significant effects on core per-formance.

The upper flow plates were changed only as follows:

A. Holes (.56 in dia.) were drilled in each corner.

These were partially blocked by the hold down pins of the new retention grid and the combi-nation will have no significant adverse effects on flow or strength characteristics of the flow plates.

B. One web was removed from each side of the flow plate and the opening partially filled by the

'new retention grid.

The hold down pins (A above) and retention grid provide poison pin holdown, against postulated LOCA and normal conditions, equivalent to the original design. The material removed from the flow plates does not significantly affect their rigidity or strength. The modified flow plate/

new retention grid does not significantly affect total fuel assembly/core flow characteristics.

To summarize, the modifications to the flow plate and the new poison pins have no signifi-"

cant effects on any performance characteristics of the fuel assemblies and the modification should prevent recurrence of the flux anomaly previously found.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

See Comments under (1) above.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

This change restores the fuel assemblies to essentially "as built" conditions regarding reactor physics parameters while having no significant effects on any mechanical properties.

Restoration of the physics performance will eliminate the flux anomaly which had the potential of exceeding Technical

'pecification li'mits on power distribution.

NOTE:See PC/M 176-76'nd attached summary also. For further details, see CEN-38 (F)-P Rev. 1, submitted with letter L-76-368 dated 25 October 1976.

Page 128 Power Distribution Anomalv and Fuel Reconstitution

,On June 30, 1976, with the reactor a't 807 power and all control rods out, a routine power distribution map gave the first indi-cation of a small azi~thal po~er tilt. This was attributed at that time co detector errors or failure. It should be noted that, at this time Technical Specif cations for til and tot 1 radial peaking factor (FT)'ere suspenaed for physics testing in accord-ance with the special test, ezceptions of the Technical Specifica-tions.~

Within the next week a few incore alarms were received. During evaluation of these, it was found that the calculated alarm set-

'points were in error (LFR 335-76-34, August 6, 1976) and it was also determined that the previously indicated tilt was still pre-sent. The alarm were corrected. On July 6, 1976, plant power was reduced to 50% fo routine cleaning of a condensate pump

@trainer. While 'at 50% power, it was determined conclus'vely tilt of approxi-from the power distribution that an azimuthal mately,4% was present along with an -cial peaking value of 1.5, as compared to an expected value of < 1.35. This tilt was veri-fied 'using the moveable incore detector system. Technical Speci-"

fications for tilt and total radial peaking factor (FT) were reinstated. It should be noted here that at no t'me was the plant in violation of any Technical Specification regarding azimuthal tilt or peaking. (L R 335-76-35, July 23, 1976)

On July 13'eactor po~er was reduced to ~bout 10 7.'nd a low power physics test program commenced. Th"s program was a repeat of selected protions oz the LPPT performed afcer initial startup.

At this time two theories were offered as possible e.planations:

1) a selective depos.'tion oz crud on the fuel leading to local flow maldistributions; and 2) early burnout of the burnable poison pins in the fuel assemblies. The results of these tests (avail-able July 18) verizied that the tilt was present, and that the core was more reactive (about .45%) than predicted. This second finding tended to support the early poison pin buznup theory. It was decided co open tne reactor'vessel foz'nspections and a shut-downlcoaldown was co enced. Over che next week, many discuss'ons were held, data was reduced and eva'uated and theories postulaced.

'None of this inzormation could conclusively explain the existing phenomenon; therefore, on July 27, actual disassembly of the reactor began.

empoison Representative fuel assemblies were removed from various areas of the vessel and inspected. The crud buildup theory was quickly dispensed wich, as blisters and perforations were found on the poison pin cladding. More fuel assemblies were removed and in-spected. Sufficient flaws were found to statistically demonstrate that there was a core wide problem with the cladding of the burnable pins. It should be noted here chat no evidence was noted of any fuel pin anormlies. Due co the core-wide poison pin problem, the plant was defueled.

Page 129 After the discovery of these poison pin cladding failures, a new theory was postulated. This was that 'the failure allowed the .

boron vithin the rods to wash out and be lost or to migrate and redistribute within the poison pins. This boron loss/redistx'=

bution theory.corx'elated much better than any other thaox'ies previously considered.

It vas then necessary to xesolve tvo major concexns: 1) what caused the cladding failure and 2) what must be done to return the plant to power operation. To aid in resolving the first concern, poison pins vere ze oved from selected fuel assemblies.

These vere submitted to on-site visual and eddy current testing.

Then they were sent to zesearch laboratories to detexMne.the cause of the cladding failures, tne mechanisms of boron loss and radistzibution, and verification that loss of boron had occured in soma pins and that it could cause the obsezved results.

As a result of these laboratory/test reactor inspections, the causa

'of the failure vas confirmed to be hydriding of the zircalloy cladding of the pins. This was caused by excessive moisture content within the pins. Under incore conditions ox high temp-erature and neutron flux the moisture produced free hydrogen wnich attacked the cladding. Zt was pzoven that the perforations did result in loss/redistribution of boron from the affected poison pins under incore conditions. And, it was confir ed that this loss/redistribution of boron could c cate the conditions observed at the St. Lucia Plant.

Then, regards~ g resolution of tha second concern, t was detexMned tnat on site replacement of the poison p ns w'th nev ones of much lower moisture content was the appropriate solution. At. the time this decision was made, some of the pin re oval equipment had already been proven in removal of the pins for testing. So, chere was reasonable assu."ance the job could, be dona even though it had to be done under water in the spent fuel pool. The vendor's specifications and controls on moisture content vere sing. 'cantly tightened to avoid repetition of the ozig'nal problem. Replacement of the poison pins resulted in fuel assemblies virtually iden-tical to tha original ones except for minor fuel depletion (ournup).

Actual reconstitution (removal of old pins and installation of new ones) commenced on October 5, 1976. The basic procedure consisted of: dzill'ng or cutting the flow plate vebs above the poi.son pins; cleaning and deburring the ne~ly machined surraces; removing old pins using a template to ensure fuel pins were not removed; in-stalling new pins; installing a retention assembly over tha flow plate; and final QC and PP&L acceptance inspection. This process is described in greater detail in the FP&L submittals leading up to Ammendment 10 to St. Lucia License DPR-67, dated 3 December 1976, vhich allows resumption of power operati,ons.

Page 130 By November 3, 1976 this process was close to complete and core reload was commenced. By November 7, all but 2 assemblies were completed and on November 10,'he last of the 108 assemblies had been reconstituted and core reload was continuing (supplementary LER 335-76-35, December 17, 1976).

We have now resumed power ascention testing and thus far have seen no evidence of any anormlies except those directly related to the uneven fuel depletion (burnup) whicn resulted from the power tilt/

peaking. These have been minor in magnitude and should be self-correctin as plant operation (and fuel depletion) continue. The activities after fuel reconsititution (fuel reload, initial criticality etc) will be descrioed in our supplementary Startup Report(s).

Page 131 Plant Change/Modification No. 194-76 PSL Unit 81 "STEAM GENEHATOR BLOMDOWN TIE-IN FROM UNIT 81 TO TREATMENT FACIL'ITY" This change tied in the blowdown lines. to the Unit 81 interface in the penetration room. The blowdown treatment facility is a requirement of our license.

This change is not an unreviewed safety question because.'.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Reyort has not been increased.

This system is part of the original design per the F.S.A.R. The tip-in and piping added in this PC/M is only to implement the system as described.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

The piping involved is designed as Seismic Class I where applicable and whip restraints are installed to prevent damage to nearby safety related equipment.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

I See comments under 1 and 2 above.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

'C 0

Page 132 Plant Change/Modification No. 196-76 PSL Unit 81 "STEAM GENERATOR PEED RING MODIFICATION" In order to prevent draining the feed ring upon low S/G water level each feed ring, had 74 nozzles penetrating the bottom. From these nozzles, standpipes (1" diameter) extended up into the feedring to near the top. An inspection revealed that the nozzles and standpipes were susceptableto vibration and fatigue failure. The standpipes were removed and the nozzles plugged. To allow feed flow and prevent draining of the feed ring 36 four inch 90 elbows were welded to the top of each feed ring o This change is not an unreviewed safety question because:

/

1. The probability of occurrence or the consequences of an accident

. or malfunction of equipment important to safety previously evaluated in the Final Safety, Analysis Report has not been increased.

The elbows will have lower velocity thus resulting in less vibration and potential for erosion so they are less likely to fail than the previous design. Peedwater instability (water hammer) is already evaluated in the PSAR and this change will not increase the probability of occurrence.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No design function or intent is changed. This modification provides a better method of implementing the original design intent.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 133 Plant Change/Modification No. 197-76 PSL Unit 81 "RELOCATE CRAFT ACCESS GATE du%) GUARD STATION" When the license was issued, contractor craft access was controlled by a guard station at the south end of the turbine building adjacent to the contractor's support facilities. As work started under the Limited Work Authorization for Unit 82, it was found desirable to separate the Unit 81 backfit craft personnel and their access entirely from the Unit 82 area and provide separate (limited) support facilities. The access gate and guard station were relocated to the north side of the site. This location is further away from the safety related systems area than the original gate and is in view of the main guard station which was not true for the old location.

I This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report, has not been increased.

Failure of the site security system is not analyzed in the FSAR.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

No change in design function or intent of the security system or its equipment was made. The location of a gate and its guard station was changed and the old location sealed in the same manner as the rest of the site security perimeter.

3. The margin of safety as defined in the basis for technical specifications has not been decreased.

Page 134 Plant Change/Modification No. 200-76 Unit 81 "FUEL HANDLQiG EQUIPMRIT MODIFICATIONS DUE TO FUEL RECONSTITUTION"

• This change added an alignment plate to the Spent Fuel Handling Machine grapple to ensuxe only the center guide-tube cou'd be grappled and that, the new retention grid

• would not be damaged by the grapple. Also, it removed part of the grapple shoe lugs on the Refueling Machine to ensure the new retention grid would not be damaged by that grapple.

This change is not an unreviewed safety question because:

1. The pxobability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increasdd.

The change will not adversely affect handling of unmodified assemblies and will ensure proper handling of the modified assemblies. No de-sign intents or functions are changed.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical specifica-tions has not, been decreased.

• See PC/M 176"76

Page 135 Plant Change/Modification No. 202-76 Unit 81 "GOVERNOR MODIFICATION FOR STEAM DRIVEN AUXILIARYFEED WATER PUMP" This change added a small oil reservoir directly to the governor system.

This provides a source of oil very close to the governor oil pump and gives faster governor response upon a quick start. This slows the ini-tial rate of acceleration and prevents overspeed trips during turbine startup.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

This change was recommended by the turbine vendor and is now includ-ed on their new units. The parts are supplied by the original vendor to the orig+al specifications. This change will improve system re-liability by preventing overspeed trips and possible equipment damage.

Auxiliary Feedwater pump failure is analyzed. This passive compon-ent will not change/increase probability of that failure.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

'I See Comments under 1. above.

3. The margin of safety as defined in the basis for technical specifications has not. been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 136 PLANT CHANGE/MODIFICATION NO. 203-76 PSL UNIT 81 "DUAL CEA EXTENSION SHAFT REPLACEMENT" This change replaced the original dual CEA extension shafts with ones modified to prevent the uncoupling problems experienced during core deload for fuel reconstitution. The major changes were replacing the tubular operating shaft with a solid one to minimize the possibility of stretching and the addition of extensions on the coupling expanders (plungers) to provide an alternate uncoupling technique if needed.

This change is not an unreviewed safety question because:

1. ,The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased.

No design intents or functions were changed and the new shafts were supplied by the original vendor to specifications as good as, or better than, original.

2. The possibility for an accident or malfunction of a'ifferent type than any evaluated previously in the,Final Safety Analysis Report has not been created.

See comments under 1. above.

3. The margin of safety as defined in the basis for technical specifica-tions has not been decreased.

This change does not represent a change to the facility as described in the Final Safety Analysis Report.

Page 137 PLANT CHANGE/MODIFICATION NO. 204-76 PSL UNIT 81 "REVISE RESTRAINTS ON LINE 2" SI-141" A flanged spoolpiece was added to this line to allow installation of a hose for use of the acoustic emission technique during plant hydrostatic testing. This change adjusts the original restraints and adds 2 new ones.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously eval-uated in the Final Safety Analysis Report has not been increased.

/

The new restraints are designed and fabricated to the same specifications as the original ones.

2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.

3. The margin of safety as defined in the basis for technical speci-fications has not been decreased.

This change do'es not represent a change to the facility as described

, in the Final Safety Analysis Report.

Page 138 PROCEDURE CHANGES'ARCH 1 1976 DECEMBER 31 1975 The following list summarizes those procedure changes which are changes to procedures as listed in the FSAR in accordance with the provisions of Title 10, Code of Federal Regulations, Section 50.59. A summary of the safety evaluation is provided for each change.

Procedure ~Chan e Operating Pro- This change allowed operation of the Spent Fuel cedure No. Handling Machine for handling new fuel for the 1600021 initial core loading without the overload interLock Unit 81 Initial in operation and allowing only one fuel assembly Core Loading to be handled at a time with no water in the spent I fuel pool. This was to comply with Amendment !$ 1 to the St. Lucie Operating License allowing such opera-tion for the period March 5 through March 19, 1976.

This was not an unreviewed safety question as the interlock is for protection against a (spent) fuel handling accident and the fuel was new and unir-radiated. The prohibition on water in the spent fuel pool was an additional precaution against in-advertent criticality.

Emergency and This procedure was revised June 15, 1976 to delete Off-Normal reference to use of the containment hydrogen sampling Procedure No. system due to problems with the environmental quali-0120042, Loss fication of the system valves which are located inside of Reactor containment. (See LER 335-76-28 dated June 18, 1976).

Coolant At the time when sampling was to commence, the hydrogen recombiners will be placed into operation to control containment hydrogen concentration. This is an interim solution; the permanent solution (as described in Followup LER 335-76-28) is to replace the valves. This does not involve an unreviewed safety question as the recombiners will be placed in operation at least as soon as originally specified.

Operating Pro- This procedure was approved December 28 to replace cedure No. the original procedure for RCS flow determination 0120051, RCS which used Reactor Cooling System AP readings to Flow Determina- determine flow. This method of determining flow is tion by Calori- independent of any geometric variations in the Reactor metric Procedure Cooling Pumps or dP instrument taps and an error analysis has shown it to be more accurate than the AP technique. A request for an amendment to our license has been submitted (letters L-76-424 of December 14, 1976 and L-77-4 of January 5, 1977) and contains full details bf this technique and its safety analysis.

Page 139 TESTS The following list summarizes those tests, other than startup tests, performed under the provisions of Title 10, Code of Pederal Regula-tions, Section 50.59. A summary of the safety analysis is included.

Back Pressure Re ulating/Letdown Valve Pressure Testin This test was conducted to determine the as installed transfer functions of the letdown and back pressure regulating valves and to determine the system response to known pressurizer level disturbances. The test was conducted by instrumenting the system response to ramp and step signal changes to the letdown valves. The purpose of the test was to accumulate data in order to improve the letdown flow operation in eon)unction with -the back pressure regulating valve to achieve smoother system. operation. The test was not an unreviewed safety question be-cause the system was operated within its design limits at all times.

CEA Guide Tube Material Xrradiation Test Pro ram - This test installed 3 zircaloy material test specimens in the St. Lucie Unit 81 core.

See PC/M-1 for details and the safety evaluation summary.

Page 140 FAILED FUEL INDICATIONS On June 30, 1976, after operation at 78% power for nearly 5 days, Iodine levels had increased by a factor of about 25 over previous 50X power lev-els. The increase was from 2.2 x 10-4 uci/ml to 5.8 x 10 uci/ml. Back-up samples and calculation of the I-131/I-133 ratio confirmed that a small fuel failure had occurred. Iodine 131 peaked at 1.35 x'10 1 uci/M, which is .17 uci/ml Dose Equivalent Iodine. Our limit is 1.0 uci/ml Dose Equivalent Iodine. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Iodine 131 levels had started to stabilize at about 2 x 10 uci/ml which is equivalent to about 0.002X failed fuel.

On July 6, power was reduced to 50% for cleaning condensate pump strainers.

Due to the power distribution anomaly

• power was not increased after cleaning the strainers. On July 9, the unit was taken off the line for low power physics testing. After this testing, the fuel reconstitution

• shutdown commenced and no further data could be obtained until startup after core reload.

After power operation commenced on December 10, Iodine levels were monitor-ed closely and as of December 31, 1976, at 50% po~er, Iodine levels were only slightly higher than those at the previous 50% plateau in June. This follows the previous tr'end in that the failure was not apparent then until 80X power was reached. It is felt that the failure may be power depend-ent and may return when higher power levels (80% and above) are reached.

• See PC/M's 176-76 and 192-76. The discussion attached to PC/M 192-76 also covers inspection of the fuel, including eddy current inspections of poison pins. It should be noted that no fuel pin failures were de-tected during these inspections.

Page 141 CORE BARREL MOMENT Section 4.4.11.3 of the PSL 81 Technical Specifications requires the results of all periodic Amplitude Probability Distribution (APD) and Spectral Analysis (SA) monitoring to be included in this report.

However, Section 4.4.11.1 requires baseline measurements at various power levels up through (nominal) 100/ power operations and a special report on the results to the NRC within 31 days after reaching 100K

/

power. ,PSL Unit 1 has not reached 100% power and has not completed this baseline study. Therefore, we have not yet performed any meaningful periodic APD and SA monitoring and have not completed the baseline monitoring which will provide the data for evaluation of later results.

The report on the baseline monitoring will be submitted as required by Section 4.4.11.1 and results of periodic APD and SA monitoring will be included in the Annual Operating Report for 1977.

Page 142 STEAM GENERATOR TUBE INSPECTIONS Section 4.4.5.5.b of the PSL 81 Technical Specifications requires reporting all Steam Generator Tube Inspections in this report.

For the period March 1, 1976 through December 31, 1976 no tube inspections were performed.

It is expected that tube inspections, as specified by Section 4.4.5.3.a of the Technical Specifications, will be. performed

'I during oux first refueling in 1978 and reported in the Annual Operating Report for that year.

Number of Personnel'100 mrem) Total Man-Rem Contract Contract Station Utility Workers Station Utility Workers Work & Job Function Em lo ees Em lo ees & Others Em lo ees Em lo ees & Others Reactor 0 erations & Surveillance:

Maintenance Personnel 0 0 0 Operating Personnel Oi 0 0 Health Physics Personnel 0 1. 20 0 Supervisory Personnel 4 0 .46 Engineering Personnel 0 0 0 Routine Maintenance:

Maintenance Personnel 18 3.63 0 .23 Operating Personnel '1 l. 31 0 0 Health Physics Personnel 3 1.60 .61 0 Supervisory Personnel 0 0 0 0 Engineering Personnel 0 0 0 Inservice Ins ection: Not Applfcable for 1976' ecial Hainrenance: (Puel Reconstitution) 0'.66 Maintenance Personnel 0 25 0 Operating Personnel 0 0 0 0 Health Physics Personnel 0 0 .36 0 Supervisory Personnel 0 0 0 0 Engineering Personnel 0 0 0 0 Waste Proces~sin Haintenance Personnel 0 0 0 0 0 Operating Personnel 0 0 0 0 0 Health Physics Personnel 0 0 0 0 0 Supervisory Personnel 0 0 0 0 0 Engineering Personnel 0 0 0 0- 0 Refueling:

Haintenance Personnel 44 10. 00 l. 08 Operating Personnel 0 0 0 Health Physics Personnel 6 .82 0 Supervisory Personnel 0 0 0 Engineering Personnel 0 0 0 TOTAL:

fhlntenance Personnel 62 27 13. 63 1.08 4. 89 Operating Personnel 1 0 1.31 0 0 Health Physics Personnel 15 0 3. 98 .61 0 Supervisory Personnel 0 4 0 En ineerin Personnel 0 0 0 0 GRAND TOTAL 78 31 18. 92 1. 69 5.35

Page 144 CONTAIN PENETRATION LEAK RATE TESTS The following routine local leak tests were performed during the reporting period to comply with Technical Specification 4.6.1.3.

Penetration Tested Test Date

1. Personnel Air Lock 11/29/76

2. Emergency Escape Lock 11/4/76 The above tests were conducted in accordance with Operating Procedure No.

1300052, Rev. 3, "Airlock Periodic. Leak Testing".

All detected leaks were within their acceptance criteria. A summary analysis of the tests will be provided in accordance with 10CFR50 Appendix J, following the next integrated leak rate test on St. Lucie Unit Pl.

Due to the length of the fuel anomaly outage, the first refueling of St'. Lucie Unit 01 will extend beyond the time interval allowed for local leak rate tests. Therefore these tests were performed during this outage from 10-5-76 to 12-2-76 to avoid a shutdown prior to the next refueling outage to comply with Technical Specification 4.6.1.2d.

All tests were performed in accordance with Operating Procedure No.

1300051, Rev. 0.

The total as-found bypass leak rate was 7.0% of the total allowable. The total as found leak rate (bypass + all other) was 18.0Z of the total allowable.

Repairs were made to five penetration boundary valves as seemed appropriate with regards to: probability of further degradation causing failure of next test, leak magnitude versus valve size, ease of repair, and scheduling.

The total as-left leakage valves were as follows:

By-Pass Leakage As-Left 4.5Z of total allowable Total As-Left 1.4X of total allowable The following table lists the valves and penetrations tested with the as-found and as-left leakage valves.

O.

Page 145 As As Found Left Penetration *SCCM *SCCM Number Valves Tested Main Steam Expansion Bellows

~

Main Steam Expansion Bellows Main Feedwater Expansion Bellows Main Feedwater Expansion Bellows Fuel Transfer Tube Expansion Bellows V-,15328, I-MV-15-1 1634.2 14.2 "I-V-18796, I-V-18794 209.0 209.0 I-V-18195, I-MV-18-1 42.7 42.7 10 I-FCV-25-4, I-FCV-25-5 19.8 19. 8 I-FCV-25-3, I-FCV-25-2 98,724.0 820. 0 14 V-6779, V-6741 12. 8 12. 8 23 I-HCV-14-7, I-HCV-14-1 3.4 3.4

, 24 I-HCV-14-6, I-HCV-14-2 8.9 8.9 26 V-2515, V-2516 1.5 1.5 28 V-5200i V-5203 29 V-5201, V5204 7.2 7.2 29 V-5202$ V-5205 79.6 79.6 31 V-65547 V-6555 41 I-V-03-1307, V-3463 4'. 8 4.8 42 I-LCV-07-11B, I-LCV-07-llA 204.2 204.2 43 V-6301, V-6302 185.0 185.0 44 I-SE-Ol-l, V-2505 54. 6 54. 6

• SCCM ~ Standard Cubic Centimeters per minute

Page 146 As As Found Left Penetration Number Valves Tested *SCCM +SCCM 46 I-V07189, I-V07206 50. 0 50. 0 47 I-V07188, I-V-07-170 4.0 4.0 48 I PSE 27 01~ 02~ 03~ 04~ 08 48 I-PSE-27-1341, I-PSE-27-10 0.2 80. 2 51 I-FSE-27-05, 06, 07, 09 51 I-FSE-27-1342, I-FSE-27-11 118.0 42.2

/

52A I-FCV-26, 01, 02 1591.2 747.2 52B I-FCV-26-03, 04 572.1 354.1 52C I-FCV-26-05, 06 ~ 1209. 8 1209.8 52D I-V00140, I-V-.00143 6.2 6.2 52E I-U00139, I-V00144 3.2 3.2 54 Blind Flange each end 0 56 I-V-25-11, I-V-25-12 250 250 57 I-V-25-13, I-V-25-14 301 301 I-V-25-15, I>>V-25-16 1105 1105 67 I-FCV-25-8, I-V-25-20 1583 1583 I-PCV-25-7, I-V-25-21 30.0 30.0 Fuel Transfer Tube Flange 1.7 1.7 Maintenance Hatch 420 420 Personnel Airlock 4.2 4.2 Emergency Escape Lock Electrical Penetrations

• SCW Standard Cubic Centimeters per minute

Page 147 ABBREVIATIONS USED A/C Air Conditioner AOV Air Operated Valve B.A. Boric Acid Coolant Charging Pump Component Cooling Water (for Rx plant components)

Ch Channel (i.e. one of four channels of the RPS) r CVCS Coolant and Volume Control System (Charging and letdown) h Control Wiring Diagram Disch Discharge Plow Control Valve Feedwater Feedwater Pump Hdr Header HPSI High Pressure Safety Injection Heat exchanger ICW Intake Cooling Water (sea water cooling for CCW, Turbine Cooling Mater)

ISO or ISOL Isolation (valve)

Ion exchanger (demineralizer)

LCV Level Control Valve LPSI Low Pressure Safety Injection MOV or MV Motor Operated Valve MSIV Main Steam Isolation Valve NI Nuclear Instrumentation PCV Pr essure Control. Valve

Page 148 ABBREVIATIONS (cont)

PRZR or PZR Pressurizer RCP Reactor Cooling Pump RV Relief Valve Rx Reactor SriC Shutdown Cooling (decay heat removal system)

S/G or S.G. Steam Generator SIT or SI'ank Safety In5ection Tank (Accumulator)

M/LP Thermal Margin-Low Pressure Transmitter VCT Volume Control Tank V/I Voltage to Current (signal) converter

It I

4

I