ML17209A861

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Annual Operating Rept for 1980.
ML17209A861
Person / Time
Site: Saint Lucie NextEra Energy icon.png
Issue date: 02/27/1981
From:
FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17209A860 List:
References
NUDOCS 8103250276
Download: ML17209A861 (109)
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Text

1980 ANNUAL OPERATING REPORT FLORIDA POWER & LIGHT COMPANY ST LUCIE UNIT 81 FEBRUARY 1981 Abstract: This report is submitted in compliance with Technical Specifications 6.9.1.5, 4.4.11.3, 4.7.6.1.2 and 10 CFR50.59.

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SUBJECT PAGE NUMBER Summary of Design Changes Per 10CFR50.59 Summary of Procedure Changes and 72 Special Tests Per 10CFR50.59 Core Barrel Movement Summary 73 Steam Generator Tube Inservice Inspection 74 Radiation .Exposure Summary 77 Challenges to Safety Valves and 78 Power Operated Relief Valves Mangrove Survey 79

I DESIGN CHANGES On the following pages are descriptions, including a summary of the safety analyses, of the design changes implemented at St Lucie Unit

/fl during the period January 1, 1980 through December 31, 1980 in accordance with 10CFR50.59.

PLANT CHANGE/MODIFICATION NO. VARIOUS (SEE LISTING BELOW) PSL UNIT 81 I& E BULLETIN 79-14 RESTRAINT MODIFICATIONS Inspections and evaluations resulting from NRC I & E Bulletin 79-14 revealed the need for modifications to various seismic restraints. The PC/M's listed below implemented these modifications.

PC/M 619-79 23-80, 54-80, 59-80, 72-80, 73-80, 80-80, 85-80, 87-80 This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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. These modifications were made to ensure the applicability of the seismic analysis. The system functions remain the same.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created for the reasons given above.
3. The margin of safety as defined in the .bases of the. Technical Specifica-tions has not been decreased. No Tech Specs are affected.

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

PLANT CHANGE/MODIFICATION NO. 138-76 PSL UNIT f31 INSTALLATION OF CHARGING PUMPS SUCTION STABILIZERS Greer Hydraulic stabilizers were installed in the suction lines of the charging pumps to minimize pressure pulsations which were causing piping vibration and instrument damage.

This change does not constitute an, unreviewed safety question as defined by 10 CFR 50.59 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. All modifications conform to original design specifications and the design will result in a more reliable system, and will not result in a change to the bases of any accident analysis.
2. The possibility of an accident or malfunction of equipment important to safety of.a different type than any previously evaluated in the Final Safety Analysis Report has not been created. The function and operation of the charging system has not been changed.
3. The margin of safety as defined in the..bases of the Technical Specifica-tions has not been decreased. N o Tech Specs have been changed.

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

PLANT CHANGE/MODIFICATION NO. 139-76 PSL UNIT P'1 PAGING SYSTEM ADDITIONS AND ALARM VOLUME OVERRIDE Several new speakers and paging stations were added in the Reactor Auxiliary Building and Turbine Building to provide additional coverage for the paging system. Also, an alarm volume override feature was added to enable the Control Room to increase the volume distributed by the paging speakers for emergency messages. This PC/M is not safety related.

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 because no safety related systems or equipment were affected.
2. The possibility for an .accident or malfunction of equipment important to safety of a diffe'rent type than any previously evaluated in the Final Safety Analysis Report has not been created because no safety related systems or equipment were affected.
3. The margin of safety as defined in the basis for Technical Specifications has not been decreased since no Technical Specifications were changed.

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

4

PLANT CHANGE/MODIFICATION NO. 140<<76 PSL UNIT /11 DIESEL GENERATOR CROSSOVER PLATFORM This PC/M provides for installation of a crossover walkway over each diesel generator. This walkway provides for safe personnel access to both sides of the diesel generator without leaving the'Diesel Generator Building. This PC/M increases personnel safety in a potential fire hazard area.

This change is not an unreviewed safety question because:

1. The probability of occurrence or the consequences of an accident mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. The walkways provided by this PC/M'ill not affect the evaluated diesel generator 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. The most severe accident that can occur is total loss of the diesel generator set, regardless of the cause. Therefore, no other type can occur than already analyzed in the Final Safety Analysis Report.

3. The margin of safety as defined in the basis for Technical Specifications has not been decreased since no Technical Specifications have changed.

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

PLANT CHANGE/MODIFICATION NO. 205-76 PSL UNIT 81 INTAKE COOLING WATER TCV LIMIT CONTROL This Plant Change/Modification provides for closure limit control on the Component Cooling Water Heater Exchangers'ntake Cooling Water Discharge Temperature Control Valves TCV 14-4A and-4B to limit closure to vendor specifications. Allowing, valve closure beyond limits specified by the TCV vendor produces turbulence which results in system and component damage.

This PC/M is not safety related.

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 subject temperature control valves are fail open valves. Additional instruments prevent valves from closing too far.'he probability of insufficient intake cooling water to the component cooling water heat exchangers is decreased.
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 probability of equipment malfunctions due to excessive valve throttling are decreased.
3. The margin of safety as defined in the basis for Technical Specifications has not been decreased since no Technical Specifications have been changed.

PLANT CHANGE/MODIFICATION NO. 209-76 PSL UNIT //1 ST. LUCIE PLANT SWITCHYARD EXPANSION This PC/M expands the switchyard to accomodate three:.;,functions:

1) Provide a 240 KV power source to supply a new distribution substation on Hutchinson Island.
2) Installation of start-up transformer breakers 2E and 4E for added reliability.
3) Installation of Transmission Line No. 3 breaker 3W for added flexibility and off-site power reliability for Unit No. 1.

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 probability of equipment malfunction which will result in a unit blackout has been decreased by the addition of the start-up transformer breakers.
2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report, Accident Analysis Section 15.2.9 has not been created.
3. The margin of safety as defined in the basis for Technical Specifications has not been decreased since Technical Specifications are not affected.

PLANT CHANGE/MODIFICATION NO. 255-77 PSL UNIT //1 CHANGE SGFP RECIRCULATION VALVE Cv & RECIRC SPARGERS The SGFP recirc valves were modified to increase flow to enable valves to pass recommended minimum flows. In addition the condenser spargers were replaced to accommodate the increased flows. This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50;59 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 PC/M was designed to decrease the potential for a FWP trip on low flow. Since the function remAins the same and since the installation meets or exceeds original specifications, the reliability has not been decreased.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

I 1 PLANT CHANGE/MODIFICATION NO. 314-77 PSL UNIT /tl RELOCATION OF SAFETY INJECTION TANK SAMPLING POINT This PC/M installed valves, tubing, restraints, control switches, etc. to extend the safety infection tank sample lines from the local sample sink inside the Containment Building to the Primary Sample Room in the Reactor Auxiliary Building. Sampling of the safety injection tanks from the Primary Sample Room is desired to reduce manhours and personnel radiation exposures associated with Containment Building entries for local sampling.

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'o safety previously evaluated in the Final Safety Analysis Report has not'een increased. The function of the safety in)ection tank sampling system remains unchanged. It is designed as Safety Class 2, Seismic Category I and is consistent with design used in existing systems. The solenoid operated isolation valves are normally closed, fail closed and close on a Containment Isolation Signal. A sampling tube break inside the -Containment Building would be detected by low safety injection tank level and/or high flow rate to the reactor cavity sump. Resulting low leak rate (< 1.5 gpm) allows sufficent time for operator action in accordance with the Technical Specifications.
2. The possibility for an accident or malfunction of equipment important to the safety of a different type than any evaluated previously in the Final Safety Analysis Report has not been created because system function remains unchanged.
3. The margin of safety as defined in the basis of the Technical Specifications has not been decreased. Local leak rate. testing will be performed to assure the" new penetration integrity is within over-all containment leak rate acceptance criteria.

PLANT CHANGE/MODIFICATION NO. 334-78 PSL UNIT 81 SE UENCE OF EVENTS ESFAS TEST POINTS 31 Additional Points from the ESFAS were added to the Sequence of Events Recorder to reduce the time and equipment required for the "Periodic Integrated Test of the Engineered Safety Features".

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 because:

1. The probability of occurrence or the consequences of an accidnet or mal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased because the PC/M does not affect any actuation, control or power circuits, but will only monitor equipment status.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created because of the reasons given above.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased since no Technical Specifications have been changed.

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

10

PLANT CHANGE/MODIFICATION NO. 444-78 PSL UNIT //1 REACTOR CAVITY LEAK RATE DETECTION SYSTEM Electronic Control Unit of Reactor Cavity Leak Rate LT-07-12 was converted to "Fail-Safe" operation.

This change does not .constitute an unreviewed safety question as defined by 10CFR 50.59 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 of all systems remains the same, the mode of operation of LT-07-12 has been changed to detect loss of power or detector failure, which will provide "Fail-Safe" operation.
2. The posibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created for the same reasons as above.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased since no Technical Specifications have changed.

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

11

PLANT CHANGE/MODIFICATION NO. 494-78 PSL UNIT }'/1 MODIFY RELAY ACC/SEC CONFIGURATION ON ALL CARD READER DOORS This PC/M rewires, deletes or adds necessary componets on all security card reader doors for the following reasons:

1) Existing wiring was causing false alarms due to relay racing and the security door opening time span cannot be adjusted.
2) Egress doors were not opening due to excess voltage drop across blocking diodes.
3) Security doors failed repeatedly due to underrated blocking diodes.

This PC/M is not safety related. 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 security door card reader system is not safety related and cannot affect any safety related systems or equipment.
2. The possibility for an accident or malfunction of equipment-.important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created because the security door card reader system function has not changed and the modification to the card reader doors cannot affect any safety related systems or equipment.
3. The margin of safety as defined in the basis of the Technical Specifications has not been decreased since no Technical Specifications have been changed.

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

12

PLANT CHANGE/MODIFICATION NO. 505-78 PSL UNIT 81 TRANSPORT OF UNIT II HEAVY E UIPMENT THROUGH UNIT 1 AREA This PC/M documented the reviews and approvals for moving the Unit II stator, rotor and NSSS heavy components across Unit 1.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 move was analysed in the Unit 1 FSAR.
2. .The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created since the move has already been analysed.
3. The margin of safety as defined in the bases of the 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.

13

PLANT CHANGE/MODIFICATION NO. 518-79 PSL UNIT 81 ROOF FOR FUEL CASK WASHDOWN AREA A removable structural steel frame/metal decking roof was added to the cask washdown enclosure to protect it from the weather.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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. The addition was designed to produce minimal affect on existing structures, is considered non nuclear safety re-lated, and will not affect any safety related systems or equipment or the bases of any safety analysis.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased since no tech. specs. have changed.

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

PLANT CHANGE/MODIFICATION NO. 520-79 PSL UNIT f/1 COMPONENT COOLING WATER SEAL CYCLONE SEPARATOR A Borg Warner Model LT-4125-DJ Cyclone Separator was installed in each seal water line in order to reduce the infiltration of corrosion products and resultant wear on the mechanical seals.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 PC/M was designed to reduce wear and increase reliabilty.
2. The possibility of an accident or'alfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created since the function of the CCW pump remains the same and the PC/M will increase the reliability of the seals.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased, and no Technical Specifications have been changed.

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

15

PLANT CHANGE/MODIFICATION NO. 528-79 PSL UNIT 81 HARDEN CRAFT ACCESS (GATE 12)

Craft access (Gate 12) was "hardened" for incxeased security to comply with NRC requirements. This change was non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 because:

1. The probability of occurrence or the consequences of an accident or mal-fun0tion of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This modification did not affect any safety related systems or equipment or the bases of any safety evaluation.
2. The possibility of an accident ox malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions were affected.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased.- No Tech Specs have been changed.

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

16

PLANT CHANGE/MODIFICATION NO. 533-79 PSL UNIT 81 SEAL LEAKOFF TANK OVERFLOW The gland seal leakoff tank overflow was routed to a drain dumping into the storm drain sys-tem instead of into the condenser pit where it caused a personnel safety problem.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased. No tech specs have been changed.

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

17

PLANT CHANGE/MODIFICATION NO. 537-79 PSL UNIT 81 CIRCULATING WATER PUMP BASEPLATE, SUPPORT The original circulating pump baseplate supports did not meet the vendor's requirements, therefore, a new vendor-approved design was implemented.

This change was non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 safety related systems or equipment were affected by this change and there was no affect on the bases of any accident analysis.
2. The possibility 'of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions were affected.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have been changed.

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

18

PLANT CHANGE/MOD IPICATION NO. 542-79 PSL UNIT P1 CONDENSATE PIPE SUPPORTS FOR DRAIN COOLER 1A Two additional hangers were added to the condensate inlet line to drain cooler 1A after investigation of a modification to the drain cooler revealed a deficiency.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 since this PC/M affects only the secondary side, is intended to correct a deficiency and does not affect the bases of any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety AnalysisReoort has not been created. No safetv functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased since no tech. specs. have been changed.

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

19

PLAN1 CHANGE/MODIFICATION NO. 551-79 PSL UNIT 81 INTAKE CANAL EROSION PROTECTION A permanent mat type bank erosion protection system was installed to reduce tur-bulence caused exosion of the intake canal banks.

This change is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 was implemented to reduce erosion problems, does not affect any safety related systems or equipment, or the bases of any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the 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.

20

PLANT CHANGE/MODZFZCATION NO. 562 79 PSL UNIT iI1 SODIUM ION DETECTORS PHASE I An improved salt water detection system is planned. This PC/M installed 1" taps into each hot well section during a unit outage for this proposed improvement. The balance of the system w'ill be completed later.

This change is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 safety related systems or equipment were affected and there was no affect on the bases of any safety evaluation.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety A'nalysis Report has not been created. No safety functions were affected.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased since no Tech. Specs. have been changed.

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

F 21

PLANT CHANGE/HODIPICATION NO. 564-79 PSL UNIT N 1 UNIT EFFICIENCY ENHANCEMENT Various modifications and improvements were made to the MSRs to improve unit efficiency.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 safety related systems or equipment has been affected and there was no affect on any bases used in the safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. The change did not affect any safety functions.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have been changed.

This change does not represent a change to the Final Safety Analysis Report.

facility as described in the 22

PLANT CHANGE/MODIFICATION NO. 567-79 PSL UNIT // 1 AUTO START OF AUXILIARY FEEDWATER PUMPS Automatic start circuitry on low steam generator level was added to all auxiliary feedwater pumps to comply with NUREG 0978. Non Safety Grade components were used for this "short term" fix.

This change does not constitute an unreviewed safety. question as defined by 10CFR 50.59 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. Safety grade isolation devices were installed to ensure that no safety related systems or equipment were affected. An evaluation was made to determine that the bases of original .safety analyses were not changed.
2. The possibility of an accident or malfunction of equipment important to safety of a'ifferent type than any previously evaluated in the Final Safety Analysis Report has not been created.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased since no tech specs have changed.

23:

PLANZ CHANGE/NODPPECATION NO. 373-79 PSL UNIT /$ 1 TEMPORARY INSTALLATION OF VACUUM BREAKERS ON ICW NON-ESSENTIAL HEADERS Temporary vacuum breakers were installed in various locations on non-essential ICW piping to prevent a recurrence of damage which occurred ~during <<est.

These vacuum breakers are an interim measure while engineering completes an analysis of the system.

This change is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 did not affect any safety related sy'stems or equipment or the bases of any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the 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.

24

PLANT CHANGE/MODIFICATION NO. 579-79 PSL UNIT 81 ALARM & LAMP TEST CONTROL CIRCUIT ARMS/PRMS ARMS & PRMS lamp test circuits were modified to eliminate 9 relays that are not used. This PC/M is not nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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. The ARMS & PRMS functions have not been revised, only test circuits that are not used have been modified.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No functional changes have occurred.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased since no Tech Specs have changed.

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

25

PLANT CHANGE/MODIFICATION NO. 586-79 PSL UNIT PI C & D MSR DRAIN CHECK VALVES V-11169 & V-11170 Existing tilt type check, valves were replaced with counter weighted swing check valves to eliminate maintenance problems.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 did not affect any safety related systems or equipment or the bases of any safety evaluations.
z. The:possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created since no safety functions have been changed.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been affected.

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

26

PLANT CHANGE/MODIPICATION NO. 600-79 PSL UNIT /3 1 FIRE PUMP SUCTION CROSS CONNECT A cross connect line and associated valves was installed to enable the fire pumps to take suction from either city water storage tank.

This change was implemented to comply with NRC requirements. The PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 because:

1. The probability of occurrence or the consequences of an accident oraal-function of equipment important to safety previously evaluated in the Final Safety Analysis Report has not been increased. This change did not affect any safety related system or equipment or the bases for any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions were affected by this change.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have been changed.

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

27

PLANT CHANGE/MODIFICATION NO. 602-79 PSL UNIT f/1 DIESEL GENERATOR DAY TANK VENT MODIFICATION The diesel generator day tank vents were rerouted outside of the diesel generator building and the day tank level guage vents were routed to the top of the tank to comply with NRC fire protection requirements.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 reliability of the D.G.S. has not been reduced,. the day tank vents locations has no affect on any safety evaluation.
2. The possibility of an accident or malfunction of equipment important 'to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

28

PLANT CHANGE/MODIFICATION NO. 623-79 PAL UNIT 81 r

REPLACE V6478 WITH A TEE AND DIAPHRAM VALVE V6478 was a 3-way valve from the Waste Management System discharging into either the "A" or "B" laundry drain tank. The 3-way valve was susceptible to blackage.

It was replaced with a tee and a new diaphram valve in conjunction with an exist-ing valve such that each tank now has an inlet stop valve in order to eliminate the blockage problems.

This change is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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. No safety related systems or equipment have been affected.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. This change did not affect any safety functions.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased since no Technical Specifications have been changed.

This change does not represent a change to the facility as described in the Final Safety Analysis Report. c 29

PLANT CHANGE/MODIFICATION NO. 624-79 PSL UNIT //1 NPS OFFICE IN CONTROL ROOM A 12' 7'ffice was constructed in the control room for the on shift NPS in order to meet NRC requirements.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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. This design conforms to current cri-teria for installations within the Control Room and an engineering evaluation has determined that no failure will have any affect on any safety related systems or equipment.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased since no Technical Specifications have been changed.

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

30

PLANT CHANGE/MODIFICATION NO. 627-79 PSL UNIT 8 1 FIRE PUMPS UPGRADE e fire pumps control circuitry was modified such that

) pumps will, auto start on loss of offsite power with coincident low header pressure.

except under ESFAS 'conditions.

2) Pressure switches provided near each pump.
3) Control room stop capability removed.

The above items were incorporated to meet NRC requirements. This change does not con-stitute an unreviewed safety question as. defined by 10 CFR 50.59 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 modifications were designed to meet original design criteria of the affected systems such that no safety systems or equipment have been degraded and the bases used in the safety analysis have not changed.

'I

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created since no safety functions have changed.
3. The:margin of safety as defined in the bases of the Technical Specifications has not been decreased. No tech specs have changed.

31

PLANT CHANGE/MODIFICATION NO. 628-79 PSL UNIT /$ 1 TEMPORARY MODIFICATION OF EFFLUENT MONITORS TO PROVIDE FOR HIGH RANGE MONITORING Temporary sample lines were tied into the plant vent and containment H2 analyzer and routed to the condenser air ejector monitors as an interim measure to meet NRC re-quirements for high range noble gas effluent monitoring.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 Analy-sis Report has not been increased since no safety related systems or equipment were affected and the modification did not change the bases of any safety evaluation.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created as no safety functions have changed.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased since no Technical Specifications have changed.

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

32

PSL UNIT //1 DDPS UPGRADE The DDPS was provided with a back-up system which required additonal power. This PC/M increased the feeder breaker size and,the cable size.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases of any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created as no safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased. No Technical Specifications have been changed.

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

33

PLANT CHANGE/MODIFICATION NO. 644-79 PSL UNIT fi 1 DIRECT POSITION INDICATION SAFETY 6 RELIEF VALVES Position indication using calibrated sonic flow detectors was- installed on each pressurizer porv and code safety valve discharge line to comply with NUREG 0578. This PC/M was non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 change does not affect any safety related systems or equipment or the bases of any safety analysis. This new position indication system will increase the ability of the operators to detect porv and safety valve leakage.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs-have been changed.

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

34

PLANT CHANGE/MODIFICATION NO. 6-80 PSL UNIT I/1 BATTERY ROOM EXHAUST DAMPER The dampers from battery rooms roof ventswere removed as requir'ed by the NRC.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 because:

1. The probability of occurence 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 absence of these dampers will have no affect on any accident analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created by removing these dampers ~
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. This change does not affect. any Tech Spec.

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

35

PLANT CHANGE/MODIFICATION NO. 7-80 PSL UNIT 81 AUTOMATIC SYNCRONIZING MANUAL PUSH BUTTON A manual pushbutton was added to the automatic syncronizing circuit as a permissive to close the generator breakers. This change was implemented to conform to FPGL standardized design. The PC/M is not nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 because this change cannot affect any accident analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created by adding a manual push-button to the generator syncronizing circuit.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

36

PLANT CHANGE/MODIFICATION NO. 8-80 PSL UNIT 81 START UP TRANSFORMERS UNIT foal/f32 INTERTXE This PC/M documents additional design details for the'ntertie of Unit I and Unit II start up transformer. This is not a design change but an ex-pected evolution resulting from the continuing construction and completion of Unit II. This change is non-nuclear safety related.

This change does not constitute an, unreviewed safety question as defined by 10CFR 50.59 because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the Safety Analysis Report has not been increased'ince this is not a 'inal change in the sense that the Unit I FSAR assumed'he construction and intertie of Unit II.
2. The possibility of an accident. or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created'for'the same reasons as given above.
3. The margin of safety as defined in the bases of'the Technical Specifi-cations. has not been decreased. No Tech, Specs- have been'hanged.

This change represents a change to the facility as described in the Final Safety Analysis Report in that Unit XI kerite cable is specified in lieu of the cable described in the FSAR.

37

PLANT CHANGE/NODTPTCATXON NO. 12-80 PSL UNIT /tl REACTOR CAVITY HANDRAIL A permanent handrail was installed on the east side of the reactor head missile shield to provide safety for personnel while working on the movable in-core detectors.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 installation was designed so that it cannot affect any safety related systems or equipment.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been changed.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. This change did not affect any tech. specs.

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

38

PLANT CHANGE/MODIFICATION NO. 18-80 PSL UNIT //1 ADD CARD READER TO ENTER CAS DOOR 109 A card reader was added to the CAS entrance door to enable guards to identify personnel requesting authorization to enter.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 does not affect any safety related systems or equipment or the bases for any safety evaluation.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. This PC/M does not affect any Tech. Specs.

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

39

PLANT CHANGE /MODIFICATION NO. 20-80 PSL UNIT /31 ADD NEW GATE STRIKES AT GATE 113 Original strikes would not allow personnel to exit area if power or circuitry failed.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created., No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

40

PLANT CHANGE/MODIFICATION NO. 24-80 PSL UNIT //1 CONTAINMENT PURGE VALVE MODIFICATION A mechanical stop was installed in the containment purge valves operators to limit opening to 50 . This modification was required by the NRC until an accident/stress analysis was completed. This anaylsis has been com-pleted by the manufacturer and has verified that the valve will operate under accident conditions from the 50 0 positions.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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. Limiting the valve opening is a conservative interim measure until the analysis can be complete.
2. The possibility of an accident of malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created for the same reason as above.
3. The margin of safety as defined in the bases of the Technical Speci-fications has not been decreased since no Tech Specs have been changed.

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

41

PLANT CHANGE/MODIFICATION NO. '6-80 PSL UNIT f/L UPGRADE OF FLOW SYSTEM/CONDENSER A/E MONITOR A .Hastings Mass Flowmeter was installed to replace an existing vacuum gage &

rotameter on the condenser air ejector monitor to provide better accuracy.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 safety related systems or equipment have been affected.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased. No Tech Specs have changed.

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

42

PLANT CHANGE/MODIFICATION NO. 27-80 PSL UNIT 81 E UIPMENT STORAGE AREA UNDER PERSONNEL HATCH Equipment storage racks were installed under the stairs and platform at the personnel hatch and the area was fenced off to provide storage space, This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

43

PLANT CHANGE/MODIFICATION NO. 33-80 PSL UNIT //1 D. G. TANK DIKE PENETRATIONS The diesel generator fuel oil storage tank dike penetrations will be sealed with fire retardent.foam .filler to satisfy NRC requirements.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.
2. The 'possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has.

not been created. No safety functions have been affected 3.The. margin of safety as defined in the bases of the Technical Specifications has not been decreased. No tech specs have been changed.

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

44

PLANT CHANGE/MODIFICATION NO. 35-80 PSL UNIT 81 AUXILIARYFEEDWATER PUMP RECIRC LINES MATERIAL CHANGE Carbon steel piping and fittings in the auxiliary feedwater recirc line were replaced with stainless steel to eliminate erosion and resultant repairs.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 modification was designed to increase the reliability of and service life of the subject lines.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No system or equipment functions have changed.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased since no Tech Specs have changed.

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

45

PLANT CHANGE/MODIFICATION NO. 38-80 PSL UNIT 81 QUENCH TANK COOLING SYSTEM Wrap around heat transfer units were installed on the pressurizer quench tank and cooled by non-essential ccw to remove heat from porv and safety valve leakage. This was required to reduce the amount of waste water generated by bleed and feed cooling evolutions.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 did not affect any safety related systems or equipment or the bases for any safety evaluation.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety func-tions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased since no Tech Specs have changed.

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

PLANT CHANGE/MODIFICATION NO. 39-80 PSL UNIT //1 REACTOR COOLANT PUMP SEAL INJECTION WATER LINES The RCP seal injection water lines connection to the RCP's was changed from flanged to swagelok to allow easier connection and disconnection for maintenance. This change is non nuclear safety related.

This change does not constitute an unreviewed .safety question as defined by 10CFR 50.59 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. The above seal injection lines are not required for any safety function and their failure will not affect any accident analysis.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created since the subject lines perform no safety function, and the function of the system has not changed.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have been changed.

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

PLANT CHANGE/MODIFICATION NO. 46-80 PSL UNIT 81 D.G. AIR START COMPRESSOR EXHAUST The exhaust piping from the engine of the air start compressor for the diesel generators was extended up the. wall to direct the exhaust away from the operator.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 exhaust pipe performs no safety function and has been seismically restrained so that it will not affect any safety system or equipment.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. All system functions remain the same.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased since no Tech Specs have changed.

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

48

PLANT CHANGE/MODIFICATION NO. 47-80 PSL UNIT

//1'URBINE GENERATOR GENERATOR CONDITION MONITOR TAPS Taps were provided on the generator to allow future connection of a generator core monitor which will monitor the hydrogen atmosphere for early detection of overheating.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 becuase:

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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

49

PLANT CHANGE/MODIFICATION NO. 49-80 PSL UNIT 8 1 MSR TUBE BUNDLE REPLACEMENT Unit II tube bundles were used to replace leaking Unit I bundles in the MSRs. This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in'he Final Safety Analysis'Report has not been increased. This change did not affect any safety related systems or equipment or the bases of any safety analysis.
2. The possibility of an accident or malfunction of equipment- important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been. created. No safety functions or equipment were affected.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have been changed.

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

50

PLANT CHANGE/MODIFICATION NO. 57-80 PSL UNIT 81 DIESEL GENERATOR CABLE CHANGE A design error resulted in the Diesel Generator field cables being undersized.

This PC/M documented the replacement with properly sized cables.

This error has been reported to the NRC in accordance with 10 CFR criteria.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 since this PC/M was implemented to correct a deficiency and all safety functions, systems and equipment remain the same.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created for the reasons given above.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

This change represents a change to the facility as described in the Final Safety Analysis Report in that Unit II Kerite Cable was used in this mod.

51

PLANT CHANGE/MODIFICATION NO. 58-80 PSL UNIT 81 BECKMAN GAS ANALYSER POWER SOURCE Excessive voltage drop in the original power cable supplying the gas analyser caused spurious alarms in the system. A new power source. and new cable were installed to correct the situation.

This PC/M is non nuclear safety related.,

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.

The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.

3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased. No Tech Specs have been changed.

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

52

PLANT CHANGE/MODIFICATION NO. 62-80 PSL UNIT 81 USE OF PSL-2 KERITE 600V POWER AND CONTROL CABLE ON PSL-1 This PC/M documents the analysis and approval to use PSL-2 approved kerite HTK ins/FR 3kt 600V power and FR ins./FR 3kt control cable on PSL-1 in lieu of the FSAR approved CLPE cable in Class IE applications.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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. Independent testing and analysis has shown that the PSL-2 cable is as good as, and in some cases performs better than the PSL-1 cable.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created by using an equal or better cable type.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have changed.

This change represents a change to the facility as described in the Final Safety Analysis Report.

53

PLANT CHANGE/MODIFICATION NO. 63-80 PSL UNIT 81 REFUELING MACHINE MODIFICATIONS Numerous modifications were made to the refueling machine equipment and controls under the supervision of the vendor to upgrade the operation and efficiency of the machine.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 modifications made all comply with original design criteria used in the safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. The function of the refueling machine remains the same, the design was implemented to improve operability and reliability and all required interlocks were maintained.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have been changed.

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

54

PLANT CHANGE/MODIFICATION NO. 64-80 PSL UNIT 81 GENERATOR REVERSE POWER INDICATION/GROSS MEGAWATT INDICATOR A digital generator gross megawatt indicator was installed on RTGB 101 with capability to indicate negative values for reverse power conditions.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 since this change di4 not affect any s"fety related systems or equipment or the bases for any safety analysis.

2. The possibility. of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

55

PLANT/CHANGE MODIFICATION NO. 66-.80 PSL UNIT 81 RAB MECHANICAL PENETRATION 'ROOM WALL PENETRAZION A 6" diameter hole was made in the east wall of the RAB mechanical penetration room to accommodate the containment spray nozzle test air supply line. When not in use, it will be sealed.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined 50.59 because:

by'0CFR

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the 7&al Safety Analysis Report has not been increased since this change did not affect any safety related systems or equipment or the bases for any safety analysis.

~ The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.

3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased. No Tech Specs have been changed.

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

'-.'inal 56

foal PLANT CHANGE/MODIFICATION NO. 70-80 PSL UNIT R EPLACEMENT OF ROSEMOUNT DIFFERENTIAL PRESSURE TRANSMITTERS Rosemount Model 1152 differential pressure transmitters were replaced with Model 1153 due to a defect in the 1152 Model identified by the manufacturer. This item was reported under 10 CFR 21.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 was required to replace an instrument model with a know defect with an improved model.

H

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created since all safety functions remain the same.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased since no Tech Specs have been changed.

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

57

PLANT CHANGE/MODIFICATION NO. 79-80 PSL UNIT 81 CONTROL ROOM DATA PROCESSOR FLOOR PENETRATION The Control Room Data Processor was relocated to improve accessability to RTGB's and Control Room habitability. This required new electrical penetrations through the Control Room floor.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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. Original design criteria was met in placing and drilling new penetration and Control Room air tightness requirements will be maintained such that the bases of previous safety analysis have not been changed.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. This PC/M will not affect any safety functions.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased since no Tech. Specs. have been changed.

Tliis change doesnot represent a change to the facility as described in the Final Safety Analysis Report.

58

PLANT CHANGE/MODIFICATION NO. 81-80 PSL UNIT 81 REACTOR VESSEL STUD HOLE 821 REPAIR Reactor Vessel Stud Hole 821 was damaged when the stud was removed for refueling. The stud hole was bored out and rethreaded to 7 3/4 diameter and a threaded sleeve was installed.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 because:

1. The probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously e'valuated in the Final Safety Analysis Report has not been increased. The reactor vessel vendor reviewed the new design and verified that it met or exceeded their original design requirements.

2.'he possibility of an accident or malfunction of equipment important to safety or a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No system or equip-ment reliability or functions have changed.

3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been. decreased. No Tech Specs have been changed.

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

59

PLANT CHANGE/MODIFICATION NO. 83-80 PSL UNIT 81 FLOW MEASUR12KNT OF OBCW-ICW Annubars were, installed to measure flow to OBCW Hz to support flow balance testing of ICW'.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysts Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech. Specs. have been changed.

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

60

PLANT CHANGE/MODIFICATION NO. 84-80 PSL UNIT 81 OCATE CONTROL ROOM E UIPMENT This PC/M provided cable routing for the relocation of the sequence of events recorder to improve operator access to RTGB's.

. This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 of an accident or malfunction of equipment important to safety of a differnet type than any previously evaluated in the Final Safety Analysis Report has not been created.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased.

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

61

PLANT CHANGE/MODIFICATION NO. 89-80 PSL UNIT Pl CABINET VENTILATION FOR SGBTF RADIATION MONITORING SYSTEM Cabinet vent fans were installed on the SGBTF RMS cabinets to reduce operating temperature and increase service life and reliability.

This PC/M is non nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.

The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.

3. The margin of safety as defined in the bases, of the Technical Specifica=

tions has not been decreased. No Tech. Specs. have been changed.

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

62

PLANT CHANGE/MODIFICATION NO. 90-80 PSL UNIT 81 RCP "START PERMISSIVE" INDICATION

'he amber speed indication light was changed to indicate CCH flow and lift oil pressure permissive signals have been received for RCP start.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

63

PLANT CHANGE/MODIFICATION NO. 95-80 PSL UNIT /31 A Tee and Valve were installed in the chemical addition tank to allow draining of chromated water to CCW pit instead of into waste management system.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.'.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased. No Tech Specs have been changed.

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

64

PLANT CHANGE/MODIFICATION NO. 112-80 PSL UNIT /$ 1 120 VAC POWER FOR PHONE E UIPMENT To comply with NRC IGE Bulletin 80-15, telephone equipment was repowered from an uninterruptable source.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 modification does not affect any safety related equipment.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have been changed.

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

65

PLANT CHANGE/MODIFICATION NO. , 120-80 PLS UNIT dl POWER SUPPLY FOR PRESSURIZER RELIEF AND BLOCK VALVES In order to comply with NUREG 0578, PORV 1402 power supply was changed to a class IE source. Although this PC/M ties equipment into a class IE supply, considered safety related since the class IE supply is protected by a class IE it is not circuit breaker and therefore no safety related equipment wi~ll be affected.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 becuase:

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 will not affect the bases of any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have changed.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased. No Tech. Specs. have been changed.

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

66

PLANT CHANGE/MODIFICATION NO. 121-80 PSL UNIT /f1 RCP SEAL WATER PROTECTION SHORT TERM MODIFICATIONS Various modifications were made to systems and components involved in the loss of CCW to RCP's event in order to reduce the possibility of again losing CCW.

This change does not constitute an unreviewed safety question as defined by =

IO CFR 50.59 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 of the system has not been changed; however, provisions have been made to allow the non-essential CCW to containment. isolation valves to be opened manually. Administrative and Operation Procedures will prevent any CIS being bypassed. The above will ensure that no bases for any accident analysis have been affected.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created for the reasons given above.
3. The margin of safety as defined in the bases of the Technical Specifica-tions has not been decreased since no Tech Specs have been changed.

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

67

e PLANT CHANGE/MODIFICATION NO. 130-80 PSL UNIT //1 DELETE ALARM FROM CONDENSATE CONDUCTIVITY ANALYSER The alarm from condensate conductivity analyser was deleted because of numerous nuisance alarms. Conductivity is monitored and alarmed by con-ductivity recorder CR 12-100.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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 equipment

'has no affect on any safety evaluation.

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions were involved.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech Specs have been changed.

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

68

i PLANT CHANGE/MODIFICATION NO. 131-80 PSL UNIT //'1 EMERGENCY POWER SUPPLY FOR TELEPHONE E UIPMENT RACK IN RAB To comply with NRC I&E Bulletin 80-15, telephone equipment in the RAB (including ENS equipment) was repowered from an uninterxuptable supply.

Although a Class IE source was used, this PC/M was considered non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10CFR 50.59 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. Non safety related telephone equipment is supplied from a Class IE electrical source through a Class. IE breaker which serves to isolate the Class IE source so that no safety related equipment can be affected.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions or equipment have been changed.
3. The margin of safety as defined in the bases of the Technical Specifi-cations has not been decreased. No Tech. Specs. have been changed.

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

69

'I PLANT CHANGE/MODIFICATION NO. 147>>80 PSL UNIT f/ 1 ALARM RX CAVITY LEAKAGE

/

rmanent modification was installed to replace temporary wiring which added a redundant alarm point on the RTGB for hi Rx cavity leakage.

This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 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 since this change did not affect any safety related systems or equipment or the bases for any safety analysis.

'I

2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.
3. The margin of safety as defined in the bases of the Technical Specifications has not been decreased. No tech specs have been changed.

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

70

0 0,

PLANT CHANGE/MODIFICATION NO. 157-80 PSL UNIT 8 1 SELF CONTAINED EMERGENCY LIGHTING FOR AUX FEEDWATER PUMP AREA contained Dual Lite D.C. lighting was installed in the IC aux. FWP area to satisfy NRC requirements. This PC/M is non-nuclear safety related.

This change does not constitute an unreviewed safety question as defined by 10 CFR 50.59 because:

1. The probability of occurrence or the consequences of an accident or malfunction of equip-ment important to safety previously evaluated in the Final Safety Analysis Report has not been increased since this change did not affect any safety related systems or equipment or the bases for any safety analysis.
2. The possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. No safety functions have been affected.

he margin of safety as defined in the bases of the Technical Specifications has not been, decreased. No tech specs have . been changed.

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

71

PROCEDURE CHANGES HEALTH PHYSICS PROCEDURE 43 LEAK CHECK AND INVENTORY OF RADIOACTIVE SEALED SOURCES This change to Procedure HP 43 addresses the surveillance requirements of radioactive sealed sources not covered by T.S. 3/4.7.9. It requires radioactive sealed sources not covered by the Tech Specs to be leak tested and inventoried annually. This is more inclusive than the existing Procedure but in variance with the FSAR which states that all sealed sources having activities greater than those as defined as licensable by 10 CFR 30 are to be tested every six (6) months.

This variance is acceptable since it does conform to the Tech Specs and 10 CFR 30 as approved by the NRC as a condition to the operating license and threrfore does not constitute an unreviewed safety ques-tion as defined by 10 CFR 50.59.

72

CORE BARREL MOVEMENT Section 4.4.11.3 of PSL 81 Technical Specifications requires the results of all periodic Amplitude Probability Distribu-tions (APD) and Spectral Analysis (SA) monitoring to be in-cluded in this report.

Routine monitoring in 1980 consisted of weekly APD pro-cessing- and SA processing performed in February, May and September. SA measurements in May were taken at nominal thermal power levels of 20%, 50%, 80% and 100% at the beginning of fuel cycle 4. At no time during the year were the alert or action levels exceeded.

As pmeriously observed and reported during the last three fuel cycles, the broadband RMS levels of all excore neutron detector signals showed a gradual increase through-out 1980 with the exception of an anticipated step decrease following refueling. This pattern has continued consistent with that observed at St. Lucie and as reported at other PWR's of similar design.

The RMS levels at the end of 1980 over the frequency band generally associated with core motion (4-10 Hertz) provide an estimate of approximately 3 mils RMS at the core barrel snubber level, a slight decrease from the 1979 estimate.

The core barrel motion at St. Lucie Unit 1 is not deemed to be significant and the APD analysis program continues to confirm that the core barrel movement is normally dis-tributed (i.e. not restrained).

73

4 "STEAM GENERATOR TUBE INSPECTIONS" An inservice eddy current examination of selected tubes in the No. 1A and 1B St. Lucie Unit No. 1 steam generators was performed during the period of A'pril 26 through .May 2, 1980, by C-E Power Systems, System Integrity Services personnel. The inspection was conducted in accor-dance with C-E Test Procedure Nos. 00000-SIS-005, Revision Ol and 00000-ESS-070, Revision 02, and satisfied the requirements of the St. Lucie Plant Technical Specification 3/4 4-5 and the A'SME Boiler and Pressure Vessel Code, Section ZI, 1974 Edition through the Summer 1976 Addenda.

The inspection program consisted of multi-frequency testing for the detection of tube wall anomalies and the assessment of tube denting with both data being taken simultaneously with one pass of-. the eddy current probe through the tube. 25KHz testing for sludge accumulation on the secondary side of. the hot side tube sheets was carried out. Selection of tubes to be examined was based on an evaluation of where, in the tube bundle, problems had occurred in other steam generators in service.

The data from the inspection was recorded on magnetic tape and also on two channel strip chart recorders with the first. recorder recording flaw data and the second recorder recording dent data. These recordings were evaluated by the data analyst and the results recorded on Eddy Current Examination Report Sheets..

Table I, below, summarizes the results of this inspection. Table I also lists the number of tubes inspected during each of the three test periods, both from the hot and the cold sides. This number does not coincide with the comparison table tube count listed under "Tubes Inspected" becuase not all tubes tested pass through. the Number 9 and 10 drilled support plates. During this inspection, all testing was carried out from the hot side. The probe was advanced up the hot leg side, through the ",U" bend and down the cold side to a point below the Number 9 cold side support plate. Data was thereby collected for both the hot leg and cold leg sides of the support plates.

The eddy current test results indicate that the. existing dents show little change in magnitude (< + 0.5 mils) at the drilled support plate elevations on .both the hot and cold sides of both steam generators in the period between the May 1979 and April 1980 inspections. There'is however an increase in the number and percentage of tubes dented in both the lA and 1B.steam generators. There were no reportable flaws detected during this. examination.

74

V TABLE I

SUMMARY

OP EDDY CURRENT TEST RESULTS INSPECTION CONDUCTED APRIL 1980 Total No. of Tubes (By design) 8519 Notes: 1) No tubes inspected from cold side Tubes thru partial support plate No. 9 2225 (26.1%) 2) Total tube count exceeds that of Tubes thru partial support pla te No. 10 771 (9.1%) support plate region because all tubes tested do not pass thru support plates.

I Tubes Examined S/G No. Tubes Examined  % of Total I

Hotside tubes thru partial support No. 9 lA 628 28.2% No Tube Wall Degradation Indications 1B 606 27.2% No Tube Wall Degradation Indications Coldside tubes thru partial support No. 9 1A 628 28.2% No Tube Wall Degradation Indications 1B 606 27.2% No Tube Wall Degradation Indications I

I Hotside tubes thru partial support No. 10 lA 305 39.6% .No Tube Wall Degradation Indications 1B 288 37.4% No Tube Wall Degradation Indications Coldside tubes thru partial support No. 10 lA 305 39.6% No Tube Wall Degradation Indications 1B 288 37.4% No Tube Wall Degradation Indications Total tubes inspected from hot side lA 735 8.6% No Tube Wall Degradation Indications 1B 827 9.7% No Tube Wall Degradation Indications Hotside sludge measurement 20 .23% Max 6.8 inches 1B 57 .67% Max 7.0 inches

TABLE I (cont)

DENT ASSESSMENT RESULTS (APRIL 1980 S/G Support Side No. Tubes Percent X D Plate No. Inspected Occurrence mila mils Hot 628 37 2.50 1.01 Cold 628 6.5 1. 11 0;.11 10 Hot 305 32 1.44 0.59 Cold 305 8.5 0.95 0.21 1B Hot 606 39 2.37 1.01 Cold 606 8.9 1.01 0.14 10 Hot 288 52 2.15 1.57 Cold 288 30 1.67 0.83 Notes:

1) All tubes inspected from Hot side.
2) X value is the average dent size for observed dents (9 not included)
3) D value includes the 5 dent tubes and add a statistical error corres-ponding to a 95% confidence level to provide a conservative assessment on tube condition at the support plates.'6

ST. LU LANT UNIT //1 80 APPENDIX B STANDARD FORMAT FOR REPORTING NUMBER OF PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION Nun>bcr of I'cfsonncl () l00 n>rcn>) 'I'<Ital hlan Rcm Contract iVo(L'ers Contract W<.rl;ers (yorl' job I'unction Station Lmployecs Utility Employees and Others Station En ployccs Utility Employccs and 0tlicrs Ikcactor Operations E.. Surveillance

. Maintenance Personnel 0 0 0 0 0 0 Opera(lug I crsoAncl 16 0 0 38.2 0 0 I ical th Physics Personnel 8

-0 0 10 19.1 17.6 Supervisory Personnel 3 0 0.8 0 En" inccring Pcrsonncl 0 0 4 0 0 1.0 Routine hlaintcnancc hlalillcnaAcc Pe(so>>Acl 101 32 '18 31. 8 8.2

10. 1 0

72.3 0

Operating I'crsonncl 26 0 0 18.9' 0 60 5.4 0 I lcalth Physics Personnel Supervisory Pc(son>>cl 1 0 2.2 0.3 Engineering Pe(sonncl 4 0 0 0.6 0 0 Insc(vice Inspection 6.8 hlaintcnancc Personnel 0 20 0 0 Opcrallllg I crsonncl 0 0 0 0 0 l leal(h I'hysics Personnel 8 0 22 2.'7 0 7.6 Supervisory Personnel 4 ,2 '4 1.8 0.9 1.6 Engineering I'cfsunnil 3 1 1 1.0 0.3 ~ 5 Special hlaintcnancc htainlcl>'i>ice I cfsslilnct 56 29 ~ 34 39.8 20.6 18.6-Opcraling Personnel 12 0 0 5.'2 0 0 llcalth I'tiysics Personnel 15 0 32 6.5 0 13.9 Supcrv'sory Pcrs>ii>llci '4 5 4 1.9 2.2 '1;6 I'A>'.inccflil' vfso>IACI 1 3 1 0.6 1.3 0.4 (Paste Processing 0 hlaintcnancc I'crsonncl 36 5 0 9.'1 1.5 3.1 0 0 Okra(>i'lg PcfsOAAC( 12 0 0 2.5 I lcalth Physi; Personnel 6 10 1.6 0 Supcfvishry I'crsonnct 1 0 2.5 0 0

. En u>ccring Personnel 0 0 0 0 0 0 ltctuelu>g hlaintcnancc I'crsi>noel 71 27 15 64.1 24.4 13.5 Opera(in Pcrs lore( 31

'0 0 3.0 0 0 lleallh I'hysi's I'crsonnel 15 0 48 1.4 0 4.6 Supervisory Personnel 4 1 0 0.4 0.1 0 Engineering Personnel 5 3 0 0.5 0.3 0 TOThL 111.2 htaintcllancc Pcrsonncl 107 88 3>) 144.8 56.6 Operating Personnel 42 0 57.7 0 0 l leal(h I'Iiysics Personnel 15 0 60 36.7 ~ 0 65. 1 Supervisory Pcrsunncl 5 5 9.6 3.5 3'.2 8 1.9 Engineering Personnel 5 3 6 2.7 1.9

'Grand Total 177 96 450 251.5 62.0 181.4

CHALLENGES TO PRESSURIZER SAFETY RELIEF VALVES In accordance with NRC letter of May 7, 1980, Five Additional TMI-2 Related Requirements for Operating Reactors, Enclosure 3, II.K.3.3.b, all challenges to the Relief Valves should be documented in the Annual Report. There were no challenges to the Relief Valves during this reporting period.

CHALLENGES TO PRESSURIZER PORV'S In accordance with NRC letter of May 7, 1980, Five Additional TMI-2 Related Requirements for Operating Reactors, Enclosure 3, II.K.3.3.d, all challenges to the PORV'S should be documented in the Annual Report.

There were no challenges to the PORV'S during this reporting period.

78

1 rt

MANGROVE SURVEY This report is in accordance with Technical Specification 4.7.6.1.2 (Flood Protection).

A comparison of the December, 1975, aerial photo FS-8770-310, with the January 15, 1981, aerial false color infrared photo indicates approximately 3 acres of mangroves. have been removed from the area adjacent to and south of the discharge canal at the point where the discharge canal penetrates the beach dunes. The removal is associated with the construction and installation of the additional discharge pipeline. This was previously reviewed during the planning phase of the construction work. This new area will be surrounded by a dike, and it is felt that there is no adverse impact on the flood protection afforded to Unit !i'1.

There continue to be indications of additional mangrove growth outside of the construction work areas. This would tend to provide flood protection to the power plant at or'reater than the design criteria level.

P

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