unit docket number experiments made as 23'994 period

ST. LUCIE UNIT 2DOCKET NUMBER 50-389

CHANGES, TESTS AND EXPERIMENTSMADE AS ALLOWED BY 10 CFR 50.59

FOR THE PERIOD OFAPRIL 23'994 THROUGH JANUARY 5g 1996

9607ii032i. 960705PDR ADOCK 05000389R PDR

0INTRODUCTION

This report is submitted in accordance with 10 CFR 50.59 (b), which requires that:i) changes in the facility as described in the SARii) changes in procedures as described in the SARiii) tests and experiments not described in the SAR

which are conducted without prior Commission approval be reported to the Commission inaccordance with 10 CFR 50.59(b) and 50.71(e) (4). This report is intended to meet thisrequirement for the period of April 23, 1994, through January 5, 1996.

This report is divided into three (3) sections; the first, changes to the facility asdescribed in the Updated Final Safety Analysis Report (FSAR) performed by a PlantChange/Modification (PC/M); the second, changes to the facility or procedures asdescribed in the Updated FSAR not performed by a'C/M and tests and experiments notdescribed in the Updated FSARg the third, a summary of any fuel reload safetyevaluations.

TABLE OF CONTENTS

S ECTION 1

114-985

003-287

025-287

143-289

016-290

178-290

054-293

081-993

132-293

173-293

025-294

068-294

104-294

111-294

125»294

148-294

008-295

023-295

027-295

029-295

PLANT CHANGE MODIFICATIONS

GOULDS CENTRIFUGAL PUMP OZL SEALS REA-SLN-85-130

BECKMAN WASTE GAS SYSTEM OXYGEN ANALYZERREPLACEMENT

DIESEL OIL TRANSFER PUMP MECHANICAL SEALREPLACEMENT

RAB MAINTENANCE WORK'REA GANTRY

CCW HEAT EXCHANGERS SHELL DRAIN ADDITIONS

ZCW PUMPS 2B 6 2C SELF LUBRICATING MODIFICATION

REPLACEMENT OF EXCORE NEUTRON FLUX MONITORING SYS

EMERGENCY COMMUNICATIONS SYSTEM UPGRADE

CV2 RELAY REPLACEMENT

MV-08-12 6 MV-08-13 MODIFICATIONS

WASTE GAS ANALYZERS

WIDE RANGE STEAM GENERATOR LEVEL UPGRADE

DELETION OF HPSI/LPSZ LO FLOW ALARMS

SAFEGUARDS BYPASS DISPLAY INDICATION PANELMODIFICATION (A7tB7)

CONTAINMENT VACUUM HI ALARM SETPOINT CHANGES

MODIFY 3Y RELAYS IN AB BUS FOR ICW/CCW 2C PUMPS

REPLACEMENT OF THE EXCORE NEUTRON FLUX MONITORINGAND PROTECTIVE SYSTEM (NZ DRAWERS) FOR THE RPSSYSTEM

NRC GEN LTR 89-10 MOV CONTROL SWITCH SETTINGS

PZR LI{}UIDSPACE INSTRUMENT NOZZLES REPLACEMENT

LETDOWN FLOW CONTROL LOOP TUNING

PAGE

10

12

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14

as

16

17

18

19

20

21

22

23

24

25

26

27

28

SECTION 1

035-295

036-295

063-295

096-295

132-295

133-295

135-295

136-295

152-295

156-295

161-295

167-295

168-295

194-295

216-295

245-295

PLANT CHANGE MODIFICATIONS (Continued)

COND TUBE 'CLEANING & DEBRIS FILTER SYS PHASE IDEBRIS FZLTER & CONT TUBE CLEANING SYS PHASE IZ

H

HP TURBINE BLADE RING CHANGEOUT

REPLACEMENT OF 2A & 2B UNDERGROUND DIESEL FUEL OZLTRANSFER LINES

V3439 & V3507 RELIEF VALVE REPLACEMENT

LOWER LETDOWN BACKPRESSURE CONTROL SETPOZNT OF V2345

THERMAL RELIEF VALVE BLOWDOWN MODIFICATZON

SAFETY RELIEF VALVE V3417 SETPOINT AND BZOWDOWNMODIFICATION

RCS HOT LEG INSTRUMENT NOZZLE REPLACEMENT

DELETION OF EDG AUTO START ON CIAS & CSAS

RCP VAPOR SEAL LEAKOFF LINE MOD

DRILLING OF VALVE DISC V3481'3545CONTROL LOGIC MOD

ZCW & CCW LOCAL PUSHBUTTON STATION REMOVAL

EMERGENCY DIESEL GENERATOR LOW COOLING PRESSUREALARM SETPOZNT

HVE 21 A&B CEDM COOLING FAN TIME DELAYSETPOINT CHANGE

PAGE

29

30

31

32

33

34

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36

37

38

39

40

41

42

43

44

5

SECTION 2

SENS-95-032

FRG 95-34

SEES-95-034

FRG 95-36

SAFETY EVALUATIONS (Continued)

REFUELING OPERATIONS WITH A STUCK REACTORVESSEL STUD

OPERATION OF NEW FUEL HANDLING CRANE FORMAINTENANCE

EVALUATION FOR THE PROVISIONS TO TRIP EMERGENCYDIESEL GENERATOR OUTPUT BREAKER ON CZAS IN PLANTMODES 5 AND 6

BYPASSING AUTOMATIC ESFAS ACTUATION DURINGOPERATING MODES 5 AND 6

PAGE

64

65

66

67

SENS-95-037 EVALUATION FOR CROSS-CONNECTING 480V LOAD CENTERS 68DURING MODES 5 AND 6

SENS-95-041

SENS-95-043

SEEP-95-045

SENP-95-046

FRG 95-48

SENP-95-094

SENP-95-096

FRG 95-145

SESP-96-054

TEMPORARY REMOVAL OF DIESEL OZL LINE TORNADOMISSILE BARRIER

EVALUATION FOR THE TEMPORARY INSTALLATION OFACOUSTICAL MONITORING EQUIPMENT

POST ACCIDENT MONITORING INSTRUMENTATION COVEREDBY TECHNICAL SPECIFICATION

EVALUATION FOR THE REMOVAL OF THE UNIT 2PRESSURIZER MISSILE SHIELD ROOF

JUMPER/LIFTED LEAD 2-95-008

DELETION OF THE CONDENSATE DEGASIFIER1

CLASSZFZCATZON OF EMERGENCY DIESEL GENERATORAUXILIARIES AND FUEL OIL PIPING ASME DESIGN, QUALITYGROUP

USING COMPUTER SOFTWARE PROGRAM AND ASSOCIATEDSENSORS FOR TESTINGS SETTING AND CALIBRATINGPRIMARY RELIEF VALVES AND OTHER RELIEF VALVES

EVALUATION OF PRESSURIZED THERMAL SHOCK(10 CFR 50.61) OF REACTOR VESSEL BELTLINE MATERIALSFOR ST LUCIE UNITS 1 AND 2

69

70

71

72

73

74

75

76

77

'6

I

F

p

A

SECTION 2

SENS-94-018

SEEP-94-035

SENP-94-037

SENP-94-039

JNO-95-001

SEMS-95-001

SEMS-95-002

SEMP-95-004

SENS-95-005

SENS-95-008

SENS-95-010

SENS-95-013

SENS-95-021

SENP-95-021

SEFJ-95-022

SEMS-95-024

SEIP-95-031

SENS-95-031

SAFETY EVALUATIONS

HYPOCHLORITE SYSTEM MODIFICATIONS

UPDATE FSAR LOOP ACCURACIES/VARIOUS ELECTRICALDISTRIBUTION SYSTEM METERS WITH CURRENT INFORMATION

SIT DISCHARGE/LOOP CHECK VALVE STROKE TEST

JUMPER/LIFTED LEAD FOR PDIS-2216

NUCLEAR PLANT CHEMISTRY PARAMETERS MANUALSREV 18

LETDOWN PRESSURE CONTROLLER PZC-2201 SETPRESSURE REDUCTION

OPERATION OF THE UNIT 2 REFUELING WATER TANKDURING INSPECTIONS

REDUCED PRESSURIZER HEATER CAPACITY

EVALUATION FOR AN ALTERNATE REACTOR COOLANTGAS VENT SYSTEM ALIGNMENT

DEENERGZZATZON OF RAB VENTILATION DAMPERSD-5B 6 D-7B TO SUPPORT ACTUATOR REMOVALFROM DAMPER D-7A

EVALUATION FOR OPERATION WITH A DIESEL OZLSTORAGE TANK BUILDING MISSILE DOOR REMOVED

EVALUATION FOR OPERATION WITH DIESEL OILTRANSFER PUMP 2B DISCHARGE ISOLATIONVALVE V17216 CLOSED

FUEL HANDLING EQUIPMENT FSAR CHANGES

DELETION OF SR-A1A EMBANKMENT SURVEY ANDANNUAL AERIAL PHOTOGRAPH OF BEACH AREACOMMITMENTS AT ST. LUCIE SITE

TEMPORARY USE OF DUMMY FUEL ASSEMBLY SKELETONREPLACING A CELL BLOCKING DEVICE FOR THESPENT FUEL STORAGE RACK

ECCS HEADER 2A2 PRESSURIZATION DUE TO BACKLEAKAGE THROUGH V3217

TEMPORARY USE OF AN ACOUSTIC FLOW METER TOCORRECT THE DDPS FEEDWATER FLOW COEFFICIENT

TEMPORARY REMOVAL OF A CCW BUILDING MISSILEBARRIER

PAGE

46

47

48

49

50

51

52

53

54

55

56

'7

58

59

60

61

62

63

I

SECTION 3

112-295

RELOAD SAFETY EVALUATIONS

RELOAD CORE DESIGN OF ST. LUCIE UNIT 2 CYCLE 9 79

SECTION 1

PLANT CHANGE / MODIFICATIONS

PLANT CHANGE/MODIFICATION 114-985

SMALL CENTRIFUGAL PUMP OIL SEAL .REPLACEMENT

~Summa rThis modification consisted of replacement of lip type oil seals onsmall centrifugal pumps throughout Units 1 and 2 with labyrinthtype seals. This package is classified as non-nuclear safetyrelated for all pumps listed with the exception of the Boric AcidMakeup (BAM) and Fuel Pool pumps on both Unit 1 and 2 and theDiesel Oil Transfer pumps on Unit 2 which are classified as nuclearsafety related. Revision One of this modification provided theadditional details for replacement of the existing Trico oilerswith standard sight glasses. The sight glasses are required toadequately monitor oil levels.

Safet Evaluate.on:

The oil seals and level gauges are located in or attached to thebearing housings of those pumps and do. not affect any pressureboundary portions of the pumps. The use of the labyrinth sealsprovides a superior sealing system in that the stationary androtating members do not touch as with the existing lip seals.Therefore, the increased likelihood for leakage caused by rubbingof the shaft with subsequent localized shaft wear does not occur.Thus the improved design significantly reduces the likelihood ofoil leakage along the pump shaft and pump reliability is increased,The oil level sight glasses utilize the same materials andconnecting mechanisms as the existing oilers while providing amethod of monitoring'il leaks over the operating band. Thesesight glasses therefore do not increase the likelihood of oilleakage. Therefore with respect to 10 CFR 50.59 the use of thelabyrinth type seals and oil level gauge; does not increase theprobability of an accident or malfunction of equipment important tosafety, does not create possible accident scenarios not previouslyaddressed by the Safety Analysis Report and does not affect orrequire changes to the Technical Specifications. Therefore, priorNRC approval was not required for implementation of thismodification.


BECKMAN WASTE GAS SYSTEM OXYGEN ANALYZER REPLACEMENT

~summar

The Waste Decay Tank oxygen analyzers continuously monitor theoxygen levels in the Waste Decay Tanks. This modification replacesthe existing Waste Decay Tank oxygen analyzers with newer modeloxygen analyzers. The replacement oxygen analyzers are morereliable and provide an additional advantage in that the sensing,element is suitable for sampling oxygen in either a liquid orgaseous sample environment.

Safet Evaluation:

The replacement oxygen analyzers perform the same function and havethe essentially the same characteristics as the previous oxygenanalyzers. The probability of occurrence or the consequences of anaccident or malfunction of equipment important to safety previouslyevaluated in the Safety analysis Report is not increased as aresult of this modification. This modification did not constitutean unreviewed safety question or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for implementation of this modification.

10

PLANT CHANGE/MODIFICATION 02S-287

DIESEL FUEL OIL TRANSFER PUMPMECHANICAL SEAL REPLACEMENT

~summa rThis modification replaces the existing, conventional (non-cartridge) mechanical seals in the Diesel Fuel Oil (DFO) TransferPumps with self-aligning, balanced cartridge type seals. Cartridgeseals of this design have been proven to be superior in performanceand reliability and decrease overall maintenance requirements.

Safet Evaluation:

This modification provides for replacement of the mechanical sealwhich utilizes a pressure retaining gland plate. The specifiedseal complies with ASME code Section III Class 3 requirements andtherefore, has been designed to the'ame criteria as the pumpassemblies. In addition, since the size and mass of the new sealgland plate is the same as the existing seal gland plate and themechanical seal is of similar mass to the existing seal, the effectof the new seal on the seismic response of the pump isinfinitesimal. Thus the Seismic Class I qualification of the pumpis retained. This modification did not, constitute an unreviewedsafety question or require changes to the plant TechnicalSpecifications. Therefore, prior NRC approval was not required forimplementation of this modification.

11


RAB MAINTENANCE WORK AREA GANTRY

~summar

The modification installs a removable 5 ton capacity floor gantryin the Unit 2 Reactor Auxiliary Building (RAB) Drumming StorageArea. The gantry has an electric motorized drive system and canaccommodate hoists up to 5 tons rated capacity. The gantry is usedto facilitate maintenance activities in the Radiation Control Area(RCA).

Saf et Evaluation:

The gantry is used to facilitate maintenance activities in theRadiation Control Area (RCA) and does not perform any safetyrelated function. However, this modification required theinstallation of concrete expansion anchors in the Seismic Class IRAB structure. Therefore, this modification has been classifiedQuality Related. This modification did 'not constitute anunreviewed safety question or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for implementation of this modification'.

12


COMPONENT COOLING RATER HEAT EXCHANGERSSHELL DRAIN ADDITIONS

~summar

This modification adds two six inch flanged pipe stub drains to theunderside of each component cooling water heat exchanger shell toreduce drainage time and to improve flushing effectiveness.

Safet Evaluation:

The new drains perform no active safety related function, only thepassive function of retaining the pressure boundary integrity ofthe component cooling water system. This modification is NuclearSafety Related since it affects the pressure boundary of the CCW

heat exchangers which are Quality Group C, Seismic Components.This modification did not constitute an unreviewed safety questionor require changes to the plant Technical Specifications.Therefore, prior NRC approval was not required for implementationof this modification.

13


ICED PUMP 2B AND 2C SELF LUBRICATION MODIFICATION

~aummar

This modification replaces the external lube water system on ICW

pumps 2B and 2C with'a self lubricating system and removes the lowlube water flow alarm functions which are no longer required. Thismodification was previously installed on ICW pump 2A and has provensatisfactory. The modification removes the ICW pump shaftenveloping tube thus exposing 'the bearings to the process streamwhich serves as their lubricant. Additionally, the existingbearings are replaced with bearings of a more suitable design andmaterial for this service. A portion of ~the external lube watersystem piping to the supply side of the upper and lower bearings ofthe ICW Pump is removed and the remaining piping is blanked off bystainless steel blind flanges. I

Safet Evaluation:

The Design Analysis for ICW pumps 2B and 2C self lubricationmodification was verified by reference to design documents, as.compared with and substantiated by design inputs. The plantTechnical Specifications were reviewed to ensure that no change toplant Technical-Specifications were involved. This modification didnot constitute an unreviewed safety question or require changes tothe plant Technical Specifications. Therefore, prior NRC approvalwas not required for implementation of this modification.

14


REPLACEMENT OF EXCORE NEUTRON FLUX MONITORING SYSTEM

~summa'his

modification replaces the existing excore monitoring systemdue to obsolescence and availability of spare parts. Thereplacement involves the following: 1) replacing the two excoredetectors and in-containment cables with Gamma-Metric detectors andcabling, 2) replacing existing amplifiers and signal processors,new equipment is installed at new locations, and 3) replacing.existing meters with new Versatile meters. The new excoremonitoring system meets the design requirements of Appendix R andRG 1.97. The new excore monitoring system maintains its originaltwo redundant class 1E channel design and function.

Safet Evaluation:

The design bases and original function for the excore neutron fluxmonitoring system as described in the FSAR remain unaffected by thereplacement system. The design of the new system meets Appendix Rand RG 1.97, Rev. 3 Category 1, Type B variable. New failure modesare not created by the replacement excore monitoring system. Thismodification did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for implementation of

this'odification.

15

PLANT CHANGE/MODIFICATION 083.-993

EMERGENCY COMMUNICATION SYSTEMS UPGRADE

~Summa

This modification upgrades the plant emergency communicationsystems to incorporate lessons learned from Turkey Point duringhurricane Andrew. Its primary puxpose is to ensure that offsitecommunications to county, state, and federal agencies and FPLdepartments is maintained during and after a hurricane. The designwind speed is set at 194 mph.

The exposed potion of the existing Unit 1 Local Government Radio(LGR) system is upgraded to withstand the required wind loads. Anew cable raceway, antenna, and mast are also added to the system.The upgraded system is accessible from the both control rooms, theTechnical Support Center (TSC), and the Emergency Offsite Facility(EOF) .

A new High Frequency/Automatic Link Establishment (HF/ALE) radiosystem is installed to provide a direct long-range communicationlink with the State of Florida in Tallahassee and the NRC inAtlanta, GA. The HF/ALE radio system also allows communicationwith other HF radios within FPL, including Turkey Point Plant, theTurkey Point EOF, the Lejunne-Flagler Office, the Juno Office, andthe Miami Radio Shop.

Permanent cellular telephone systems are also added. Each basestation is parallel connected to a new telephone in its respectivecontrol room and the TSC. The antennas for these systems areinstalled on the auxiliary building roofs and rated for 125 mphwinds.

The existing low band channel C radio system is removed from theplant in concert with" the communications upgrade.

Safet Evaluation:

The new hardware and LGR modifications pxovide a diverse offsitecommunication and notification capability designed to functionunder hurricane foxce wind conditions. The modifications did notconstitute an unreviewed safety question or xequire changes to theplant Technical Specifications. Prior NRC approval was not,required for implementation.

The newly installed transceivers, antennas, masts, conduit andraceways do not adversely affect the integrity of any safetyrelated structures or interact with any safety related systems orcomponents that provide accident mitigation functions.

16


CV-2 RELAY REPLACEMENT

~slam ar

This Engineering Package (EP) provides the design necessary toreplace the existing Westinghouse type CV-2 undervoltage relays inthe 480 Volt PSB-1 cabinets with the solid state relaysmanufactured by ABB, type 27N. These relays are a part of theundervoltage/degraded voltage (PSB-1) protective scheme. Thismodification was requested by the System protection and ElectricalMaintenance personnel to enhance the outage calibration activities.In addition, this EP adds to each of the existing 4.16 kVundervoltage protective relaying schemes a timer relay to provideadequate time for the test circuit to change state before re-armingthe trip circuit.

Safet Evaluation:

The new hardware and the modifications to the undervoltageprotection scheme result in more sensitive protection and detectionof the degraded voltage conditions of the electrical distributionsystem. New relays are easier to calibrate and set and theirtolerances are much smaller than those of the relays beingreplaced. All the undervoltage protective functions areindividually testable and the addition of timers to the loadshedding circuitry eliminates possible relay race potentiallyresulting in inadvertent start of EDGs. These performance featuresprovide for an enhanced fulfillment of the original design basis.This modification did not constitute an unreviewed safety questionor require changes to the plant Technical . Specifications.Therefore, prior NRC approval was not required for implementationof this modification.

17


MV-08 12 & MV-08-13 MODIFICATIONS

~summa

This modification replaces the existing Unit 2 main steam systemvalves MV-08-12 & MV-08-13 (auxiliary feedwater pump 2C steamadmission valves), including the valve actuators. These valveswere replaced to improve component reliability. The existingvalves were a wedge gate design. The new valves are double discgate valves with enhanced Limitorque actuators. The newreplacement valves and actuators meet the design requirements,quality classifications, and code requirements of the originaldesign. The design bases and function of these valves are notchanged by the replacement valves.

Safet Evaluation:

The design bases and original function for MV-08-12 & MV-08-13 asdescribed in the FSAR remain unaffected by the replacement valves.This modification did not constitute an unreviewed safety questionor require changes to the plant Technical Specifications.Therefore, prior NRC approval was not required for implementationof this modification.

18


HASTE GAS ANALYZERS

~8ummar

This modification installs a condensate removal system off therefrigerant cooler at the suction of the Automatic Gas Analyzersample pump, a conversion of a temporary installation into apermanent installation. This modification provides increasedreliability of the waste gas management system's automatic gasanalyzer by removing condensate from the sample stream which couldotherwise lead to failure of the sample pump.

Safet Evaluation

The waste gas analyzer is not required to perform any safetyfunction related to mitigating the consequences of an accident.However, due to the potential for the release of gaseousradioactive effluent into the confines of the Reactor AuxiliaryBuilding (RAB), this modification is considered Quality Related;The installation of the condensate recovery system off therefrigerant cooler inside the automatic gas analyzer room in theRAB has no impact on any safety related system or equipment. Thismodification did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for implementation of thismodification.


WIDE RANGE STEAM GENERATO LEVEL UPGRADE

~summa rThis modification provides additional Wide Range Steam GeneratorLevel (WRSGL) instrumentation (control room indication) for PostAccident Monitoring. This modification provides additionalinstrumentation to the original plant configuration and is beingadded to the FSAR Table 7.5-1. This modification fulfills acommitment to the NRC to add redundant wide range steam generatorlevel indication in lieu of meeting all the requirements for RGla97 Category 1, Type D,variable.

Safet Evaluation:

The instrumentation added by this modification provides itsfunction, thus there is no direct or indirect impact on theanalysis of any design basis accident, nor is there an increasedpotential of any radiological hazards. In addition, the potentialfor an un-analyzed accident 'as 'not been increased by thismodification. This modification did not constitute an unreviewedsafety question or require changes to the plant TechnicalSpecifications. Therefore, prior NRC approval was not required forimplementation of this modification.

20


DELETION OF HPSZ LPSZ LO FLOW ALARMS

~Bummar

This modification removed obsolete non-safety related ultrasonicflow monitors from HPSI and LFSI piping. These monitors wereinstalled early in the life of the plant to provide additionalannunciation in the control room of potential low flov conditionsin the HPSI and LPSI pumps during shutdown cooling and post-LOCA.

Safet Evaluation:

The monitors did not perform any safety related functions and werenot required to operate the plant. The monitoxs vere not requiredto mitigate the consequences of an accident, were not requixed tosafely shutdown the plant, and did not provide input to anyautomatic function. The monitors were redundant to safety relatedindication and plant procedures for identifying and correcting lowflow conditions. This modification did not constitute anunreviewed safety question or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for implementation of this modification.

PLANT CHANGE/MODIFICATION 111 294

SAFEGUARDS BYPASS DISPLAY INDICATZONPANEL MODIFICATION

~Summar

The Safeguards Bypass Indication System is a part of the plantannunciator system. The bypass panel provides the operator withinformation about the important safety related systems which areremoved from service, tested or being repaired, or disabled. Thismodification corrected the anomaly that windows A7 and B7 wereincorrectly labeled for the existing inputs by deleting the

inputs'o,

blanking and sparing windows A7 and B7.

Safet Evaluation:

This modification had no functional interaction with any equipmentor structures important to safety. The modified section is notpart of the safety related portion of the annunciator system. Thismodification did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for implementation of thismodification.

22

PLANT CHANGE/MODXFXCATXON 125-294

CONTAINMENT VACUUM HX ALARM SETPOXNT CHANGES

~summa rThis modification revised the setpoint for Containment to AnnulusDifferential Pressure Indicating Switches PDIS-25-11A and PDIS-25-11B. The PDIS-25-11A and PDIS-25-11B setpoints were changed from-11.5" Wg to -9.0" Wg in ordex to provide the containment highvacuum alarm before the technical specification limit isencountered. The setpoints for opening (-9.85" Wg) and closing(7.75" Wg) of containment vacuum relief valves by transmitters PDT-25-1A, 1B, 13A, and 13B were not changed. In addition, thismodification provides an early tripping of containment purge fans.

Safet Evaluation:

The Containment Purge System is designed to reduce the level ofradioactive contamination in the containment atmosphere below thelimits of 10 CFR 20 so as to permit personnel access to thecontainment during shutdown and refueling. The containment purgefans are designed to trip on high differential pressure between thecontainment and the annulus in order to prevent 'ncrease incontainment vacuum. An early warning of containment high vacuumcondition and the early tripping of the containment purge fans doesnot affect the opexation of Containment Vacuum Relief System andContainment Purge System. This modification did not constitute anunreviewed ~ safety question or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for implementation of, this modification.

23'


MODIFY 3Y RELAYS IN AB BUS FOR XCW CCW 2C PUMPS

~Summa rPrior to this modification the 2C CCW and 2C ICW pumpsautomatically started immediately upon receipt of a SIAS if offsitepower was available and started using the EDG load sequence timersif offsite power was not available. This modification changed theautomatic start circuits of the 2C CCW and 2C ICW pumps such thatthe pumps automatically start using the EDG load seguence timersregardless of offsite power availability to the 4.16 kV 2AB bus.This change was made by eliminating the function of relays 3YA1 and3YB1 for the 2C CCW pump and relays 3YA2 and 3YB2 for the 2C ICWpump. Elimination of the function of these four relays results ina pump loading time delay for an automatic start signal, even underconditions with offsite power available. The manual startin'g ofthe pumps has not been affected (no time delay).

Safet Eva1uation:

The changes simplify the pump start circuits by effectivelybypassing relays 3YA1, 3YA2, 3YB1, and 3BY2, which results inelimination of a premature start failure mode. This modificationdid not constitute an unreviewed safety question or require changesto the plant Technical Specifications. Therefore, prior NRCapproval was not required for implementation of this modification.

24

PLANT CHANGE/MODIFICATION 008-295REVISION

REPLACEMENT OF THE EXCORE NEUTRON FLUX MONITORINGAND PROTECTIVE NI DRAWERS FOR THE RPS SYSTEM

~summa

This modification replaced the RPS Nuclear Instrumentation (NI)System as follows:

The four (4) existing Westinghouse wide range excore detectorsRE-001-A2, B2, C2 & D2, along with the detector cable wasreplaced with Gamma-Metrics wide range detectors and detectorcable.

2) The four (4) existing Amplifiers (RT-001A, B, C & D) werereplaced with Gamma-Metrics amplifier assemblies. TheAmplifier assemblies were moved to a new location in the samegeneral area.

3) The four (4) filter assemblies were removed (RX-110A, B. C &

D). The new amplifier assemblies did not require externalfilters.

4) The four (4) existing RPS NI drawers (RPS CAB A, B, C &

D(Assoc. W)), were replaced with new Gamma-Metrics NuclearInstrument drawers (Analog display).

Safet Evaluation:

A safety evaluation for this replacement was performed inaccordance with 10 CFR 50.59. This evaluation concluded that theimplementation of this modification did not involve an unreviewedsafety question nor a change to plant Technical Specifications andhad no detrimental effect on plant safety or operation. Therefore,prior NRC approval was 'not required for implementation of thismodification.

25


NRC GENERIC LETTER 89 10 MOV CONTROL SNITCH SETTINGS

~8llEIIa

This engineering package (EP) provided enhancements to motoroperated butterfly valves for selected safety related valves toensure that the valves will function as intended during maximumdesign basis conditions. These enhancements provide for: 1) fieldtesting to verify that the MOV torque switches are properly set toallow the MOVs to generate sufficient torque to perform the mostlimiting stroke function without over stressing any valve oractuator component, and 2) modify certain MOVs to better facilitatediagnostic testing of the MOVs. This'P provides the specificdesign information necessary to set both the open and close torqueswitches for twenty-two (22) St. Lucie unit 2 motor operatedbutterfly, valves within the scope of NRC Generic Letter 89-10.Also within the scope of this EP is the replacement of actuatorgearing for ten (10) motor operated butterfly valves and thereplacement of one (1) thermal overload.

Safet Evaluation:

The valve enhancements accomplished by this engineering package donot adversely affect the ability for these valves to perform theirdesign functions. The MOVs operate in a manner identical to thatprior to the modification. This modification did not const'itute anunrpviewed safety question or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for implementation of this modification.

26


PRESSURIZER LI UID SPACE INSTRUMENTNORRLES REPLACEMENT

~Summa

This Engineering Package (EP) modified the Pressurizer liquid spaceinstrument nozzles E and F on the bottom of the Pressurizer and atemperature nozzle on the side. The replacement nozzles are lesssusceptible to primary water stress corrosion cracking (PWSCC).The modification moved the partial penetration weld joint to thePressurizer outside surface whereas the former joint was on theinside.

Safet Evaluation:

The replacement Pressurizer instrument nozzles are equivalent tothe former nozzles and fabricated from acceptable material.Welding of the weld buildup pads and partial penetration J weldswere controlled by weld procedures which were reviewed and approvedby FPL. The new nozzle design does not change, degrade, or preventactions described in, or assumed to occur in the mitigation of anyaccident described in the SAR. This modification did notconstitute an unreviewed safety question or require changes to theplant Technical Specifications. Therefore, prior NRC approval wasnot required for implementation of this modification.

27


LETDOWN FLOW CONTROL LOOP TUNING

~Summa

Letdown flow is a function of the Pressurizer Level Control System.As Pressurizer level changes in response to Reactor Coolant Systemfluctuations, control signals are generated to start/stop chargingpumps and to modulate the Letdown Flow Control Valves (LCV-2110P &

Q), accordingly. A drop in pressurizer level will requireadditional charging and reduced letdown flow. The followingmodifications were made to the Pressurizer Level Control System toimprove the balance and smooth operation of the system:

a

~ Changing LIC-1110X & X's gain from 6.02 (16.6% proportional) to5.14 (19.444 proportional).

~ Providing a' —20 second operating range for the lag componentwithin the control loop to accommodate changes in the va'lvecharacteristics in the future.

~ Changing the high pressure alarm setting from 500 to 535 psig.~ Changing the low pressure alarm setting from 420 to 415 psig.

Safet Evaluation:

This modification did not change the design function of the letdownsystem as described in the FSAR. The mechanical functions for theletdown system remain unchanged by this design. The controlchanges enhance the operation of the system. This modification didnot constitute an unreviewed safety question or require changes tothe plant Technical Specifications. Therefore, prior NRC approvalwas not required for implementation of this modification.

28


CONDENSER TUBE CLEANING & DEBRIS FILTER SYSTEM PHASE I

~summa

This modification installed a Debris Filter System (DFS) whichmechanically removes entrained solid particles and a ContinuousCondenser Tube Cleaning System (CTCS) that mechanically cleans thecondenser tubes. The DFS operates continuously with a periodicautomatic backwash to the circulating water discharge. The CTCSoperates continuously to inject sponge-type balls into thecondenser where they pass through (and clean) the tubes, arecollected on the discharge, and returned for re-use. The neteffect for both the DFS and the CTCS is to reduce condenserdowntime for cleaning and prevent, a reduction in generatingcapability due to condenser fouling limitations. This PCM installscomponents (auxiliary piping/valves, instrumentation, andelectrical components) of the DFS and CTCS. Taprogge AmericaCorporation (TAC) is providing the modification and installation ofthe DFS and CTCS spool pieces in the circulating water inlet andoutlet lines under PCM 036-295. Two PCMs were developed in orderto differentiate the implementation scopes of the contract withTAC.

Safet Evaluation:

The main condenser and circulating water system are both non-safetyrelated as neither the system or any of its components perform asafety function. As such, the implementation of this modificationdoes not constitute an unreviewed safety question and theimplementation of this modification does not require a change tothe plant Technical Specifications. Therefore, prior NRC approvalwas not required for implementation of this modification.

29

t I


CONDENSER TUBE CLEANING & DEBRIS FILTER SYSTEM PHASE XX

~summaII

This modification, installed a Debris Filter System (DFS) whichmechanically removes entrained solid particles and a ContinuousCondenser Tube Cleaning System (CTCS) that mechanically cleans thecondenser tubes. The DFS operates continuously with a periodicautomatic backwash to the circulating water discharge. The CTCSoperates continuously to inject sponge-type balls into thecondenser where they pass through (and clean) the tubes, arecollected on the discharge, and returned for re-use. The neteffect for both .the DFS and the CTCS is to reduce condenserdowntime for cleaning and prevent a reduction in generatingcapability due to condenser fouling limitations. This PCMinstalled the DFS and CTCS spool pieces in the circulation waterinlet and outlet lines by Taprogge America Corporation (TAC). PCM035-295 installed components (auxiliary piping/valves,instrumentation, and electrical components) of the DFS and CTCS.Two PCMs were developed in order to differentiate theimplementation scopes of the contract with TAC.

Safet Evaluation:

The main condenser and circulating water system are both non-safetyrelated as neither the system or any of its components perform asafety function. As such, the implementation of this modificationdoes not constitute an unreviewed safety question and theimplementation of this modification does not require a change tothe plant Technical Specifications. Therefore, prior NRC approvalwas not required for implementation of this modification.

30


HP TURBINE BLADE RING CHANGEOUT

~mumm ar

This modification replace HP turbine blade ring, blade ring carrierbolting, blade ring pins, outer gland case and associated hardware.These changes were made per a vendor recommendation to precludepremature failure of the components due to erosion or erosion-corrosion. The modified components are all non-rotating parts thatdirect steam flow over the rotor blade, provide internal support orseal the shaft area. The blade rings 'and their bolting aid incontaining postulated turbine missiles.

Safet Evaluation:

The replacement components resulted in no degradation, eitherdirectly or indirectly, to any safety functions required foranalyzed accidents, and do not increase any radiological hazards.This modification did not constitute an unreviewed safety questionor require changes to the plant Technical Specifications.Therefore, prior NRC approval was not required for implementationof this modification.

31


REPLACEMENT OF 2A & 2B UNDERGROUNDDIESEL FUEL OIL TRANSFER LINES

~summa

This modification rerouted and replaced the underground fuel oiltransfer lines between the 2A & 2B diesel oil transfer pump and theexpansion joint upstream of the 2A1, 2A2, 2B1, & 2B2 diesel oil daytanks. This modification .was essentially a like-for-likereplacement of the existing piping, with a few enhancements added.The new underground piping utilizes a cathodically protected guardpipe which provides"'improved corrosion protection and isolates anypotential future fuel oil leakage from the environment. The newpiping is provided with design, basis tornado missile protection inaccordance with FSAR requirements. The old diesel fuel oiltransfer lines have been abandoned in place.

Safet Evaluation:

This modification was essentially a like-for-like replacement. TheEDG fuel oil system and the EDGs have not been effected. Thedesign differences between the original design and the design ofthe new piping have been addressed and determined to be acceptable.This modification did not constitute an unreviewed safety questionor require changes to the plant Technical Specifications.Therefore, prior NRC approval was not required for implementationof this modification.

32

PLANT CHANGE/MODZFZCATEON 132-295

V3439 & V3507 RELIEF VALVE REPLACEMENT

~Summar

This modification replaced relief valves V3439 and V3507, increasedthe lift pressure from 500 to 535 psig, reduced the blowdownpercentage from 104 to 6% — 84, and increased the discharge pipingsize from 1/2" to 2" diameter. This modification was performed toreduce the potential for V3439 or V3507 to lift during normalshutdown cooling system (SDC) operation and reduce the time for thevalve to reseat.

Safet Evaluation:

Valves V3439 and V3507 provide, protection against pressuredeveloped due to fluid thermal expansion in an isolable section ofthe low pressure safety injection system (LPSI) headers 2A and 2B,respectively. The function of V3439 and V3507 was not changed bythe installation of the new replacement valves, the increase in the-set pressure, the reduction in blowdown setting and themodification of the discharge piping. This modification did notconstitute an unreviewed safety question or require changes to theplant Technical Specifications. Therefore, prior NRC approval wasnot required for implementation of this modification.

33

PLANT CHANGE/MODZFZCATZON 133-2 95

LOWER LETDOWN BACKPRESSURE SETPOZNT OF V2345

~Summar'

This modification lowered the operating set pressure for theLetdown intermediate leg pressure controller PIC-2201 from 460 to430 psig and associated alarm PA-2201 setpoints revised to 510 and395 psig, and reduced the design blowdown value for the existingrelief valve V2345 from 254 to 15 +0/-24, but maintained V2345'spressure setpoint at 600 psig. The modification was made toprovide positive margin between the system normal operatingpressure and the pressure at which V2345 will reseat followingactuation.

Safet Evaluation:

Valve V2345 is installed downstream of the LCV-2110P&Q to provideoverpressure protection for the .intermediate pressure letdownpiping ,and letdown heat exchanger should PCV-2201P&Q closeunexpectedly. The reduction of the Letdown Backpressure Controlset pressure to 430 psig did not adversely affect the operation ofthe Letdown system. The bases of the backpressure control setpressure is to ensure that the fluid downstream of the letdowncontrol valves does not flash to steam following a reduction inpressure to the intermediate pressure piping system pressure.Relief valve V2345 reduction in the blowdown from 25% to 15 +0/-2%increases the reseat margin of the valve. The function of V2345has not been changed. Adjustments of. the alarm setpoints ensuresthat adequate margin was provided to eliminate spurious alarmindications, while also ensuring that abnormal operating conditionsare identified to plant operators through annunciation. Thismodification did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for implementation, of thismodification.

34


THERMAL RELIEF VALVE BLOWDOWN MODIFICATION

~summar

This modification reset thermal relief valve V3412's blowdown from254 to 10-154 to ensure that the valve's minimum reseat pressureexceeds the maximum operating pressure'n HPSI header 2B. The liftpressure is unchanged at 1585 psig.

Safet Evaluation:

Reducing the blowdown from 25% to 10-154 ensures that valve V3412will reseat even if there is a large unanticipated pressuretransient following the valve lift which could prevent the valvefrom reseating. The function and the setpoint of the valve has notbeen changed. This modification did not constitute an unreviewedsafety question or require changes to the plant TechnicalSpecifications. Therefore, prior NRC approval was not required forimplementation of this modification.

35


SAFETY RELIEF VALVE V3417 SETPOINT ANDBLOWDOWN MODIFICATION

~Summa'

This engineering package (EP) modified the St. Lucie Unit 2 reliefvalve V3417. Valve V3417 provides protection against pressuredeveloped due to charging pump discharge into an isolable sectionof the 2A high pressure header. The pressure setpoint of V3417 wasincreased from 2400 to 2485 psig, the blowdown was reduced from 254to 10-12%'. This modification was performed to reduce the potentialfor V3417 to lift during system testing and off-normal operationand to reduce the potential of the valve failing to reseat atsystem operating

pressures.'afet

Evaluation:

The function of V3417 was not changed by this increase in thesetpoint and the reduction in blowdown settings. This modificationdid not constitute an.unreviewed safety question or require changesto the plant Technical Specifications. Therefore, prior NRCapproval was not required for implementation of this modification.

36

PLANT CHANGE/MODXFECATZON 152-295

RCS HOT LEG XNSTRVMENT NOZZLE REPLACEMENT

~Summa

This engineering package(EP) modified the St. Lucie Unit 2 ReactorCoolant System (RCS) hot leg instrument and sample nozzles. Thenozzles are designated as flow measurement nozzle J (quantity ofeight) and sampling nozzle K (quantity of one) on drawing 2998-3793. These nozzles were made of a specific heat of material thatis known to be susceptible to primary water stress corrosioncracking (PWSCC) and the flow measurement nozzle J for PDT-1121Bwas found to be leaking on 10/10/95. The purpose of thesereplacements was to remove and replace the sensing line nozzle forPDT-1121B and to minimize the possibility of a future nozzlefailure. The modification moved the partial penetration weld jointto the hot leg outside surface where previously the joint was onthe inside surface. The material of the nozzles was changed fromZnconel 600 to Inconel 690.

Safet Evaluation:

This engineering package did not involve an unreviewed safetyquestion, not did it recpxire a revision to the plant TechnicalSpecifications. This modification did not effect plant safety andoperation. Therefore, prior NRC approval was not required forimplementation of this modification.

37


DELETION OF EDG AUTO START ON CXAS 6 CSAS

~summa

This modification deleted the Containment Isolation ActuationSignal (CIAS) and Containment Spray Actuation Signal (CSAS)Emergency Diesel Generator (EDG) start signals. The EDG start forSafety Injection Actuation Signal (SIAS) and Loss of Offsite power(LOOP) start signals remain. This modification was necessary toprevent operation of the EDGs with protective trips blocked whileparalleled with offsite power. CIAS and CSAS are redundant tripsto SIAS and the SIAS opens the EDG output breaker preventingoperation of the EDG in parallel with offsite power and withdisabled protective trips.

Safet Evaluation:

This modification did not affect the operation of the ContainmentIsolation System or the Containment Spray System. All functions ofthe CIAS and CSAS remain'ith the exception of the EDG startsignals. The SIAS signal starts the EDG providing emergencyelectrical power. The purpose of this modification was toeliminate any possibility of damage to the EDGs due to spuriousactuation of CIAS or CSAS relays, which may result from operatingin parallel to offsite power with the'protective relay functionsblocked. This modification did not constitute an unreviewed safetyquestion or require changes to the plant Technical Specifications.Therefore, prior NRC approval was not required for implementationof this modification.

38


RCP VAPOR SEAL LEAEOFF LINE MODIFICATION

~Summar

This modification replaced and rerouted the Not Nuclear Safety(NNS) reactor coolant pump (RCP) vapor seal lines. The RCP vaporseal drain system was reconfigured from a closed system thatdrained to the reactor drain tank to an open system that drains tothe containment sump via open floor drains. This modification wasaccomplished to prevent accumulation of boric acid crystals on thetops of the RCP vapor seal assembly due to lack of proper drainagefrom the vapor seal assembly.

Safet Evaluation:

Operation of plant systems, including the RCPs and the liquid wastemanagement system, was unaffected by this change; however,operators have been informed that RCP vapor seal leakage is nowdirected to a floor drain where it may slightly affect containmentatmosphere radiation levels and reactor cavity sump levelindication. This modification did not constitute an unreviewedsafety question or require changes to the plant TechnicalSpecifications. Therefore, prior NRC approval was not require forimplementation of this modification.

39


DRILLING OP VALVE DISK V3481

~Summa'his

modification consisted of drilling a hole on the upstxeam orReactor Coolant System (RCS) side of the valve disk of ShutdownCooling (SDC) isolation valve V3481 in order to vent the bonnet ofhigh pressure fluid, and thereby prevent potential pxessurelocking. This modification was performed to satisfy therequixements of NRC Generic Letter 95-07 (endorsed under NUREG-1275Vol. 9), and has been performed within the nuclear industry toaddress the pressure locking issue.

Safet Evaluation:

The functional safety related requirement of V3481 to open orclose, or to maintain the integrity of the LPSI/SDC were notchanged by drilling a hole on the upstream or'CS side of the disk.This modification enhances the reliability of the valve and the SDCsystem since it will vent the bonnet of high pressure fluid andthereby eliminate the potential for a pressure locking condition.This modification will not affect the structural integrity or theseismic qualification of the valve. This modification did notconstitute an unreviewed safety question or require changes to theplant Technical Specifications. Therefore, prior NRC approval wasnot require for implementation of this modification.

40

PLANT CHANGE/MODIFICATION 168 295

3545 CONTROL LOGIC MODIFICATION

~summar

This Engineering Package changed the position of valve V3545 fromnormally locked closed to normally locked open to eliminate thepotential for pressure locking the valve in its isolated position.The "valve open« annunciator was rewired to annunciate a "valveclosed" position and the RTGB hand switch was reconfigured topermit the valve to be administratively locked open.

,Safet Evaluation:

The shutdown cooling suction cross-tie line was added to the Unit2 system design to satisfy single failure criteria. A cross-tieisolation valve was included in the design to preserve the

ability'o

remove one train from service for maintenance or repairs.Changing the normal valve position from normally shut to normallyopen does not affect the valves functional ability to isolate thecross-tie. This modification did not constitute an unreviewedsafety question or require changes to the plant TechnicalSpecifications. Therefore, prior NRC approval was not. required forimplementation of this modification.

41


CW & CCW LOCAL PUSH BUTTON STATION REMOVAL

This engineering package (EP) disconnected the St. Lucie Unit 2

local push-button control station from the 2A, 2B, and 2C IntakeCooling Water (ICW) pump and the 2A, 2B, and 2C Component CoolingWater (CCW) pump control circuits. This modification wasimplemented to delete the operational requirement of manuallyresetting the RTGB control switch to "stop" and then back to "auto"which may have been required to preserve the automatic SIAS startfeature of the subject pumps. The resetting of the RTGB controlswitches was only required after a local push-button stop of arunning pump which was started via the RTGB control switch.Disconnection of the local push-button stations from the associatedcontrol circuit prevents the stopping of a running pump via thelocal push-button station and thus preserve the automatic SIASstart feature of the pumps without requiring a manual reset of theRTGB control switches.

Safet Evaluation:

This engineering package did not involve an unreviewed safetyquestion nor did it require a change to the .TechnicalSpecifications. It was also determined to have no adverse effecton plant operations or safety. Therefore, prior NRC approval wasnot required for implementation of this modification.

42


EMERGENCY DIESEL GENERATOR LO'R COOLINGPRE88URE ALARM SETPOINT

~summa@

This engineering package lowered the Emergency Diesel Generator(EDG) 2A and 2B Low Cooling Pressure Alarm setpoint from 30 psigdecreasing to 23 psig decreasing. This setpoint reduction was doneto eliminate a nuisance alarm that occurred when the EDG speed wasreduced from full speed to idle speed.

Safet Evaluation:

This engineering package did not have an adverse effect on plantsafety, security or operation, did not constitute an unreviewedsafety question and did not require a change to plant TechnicalSpecifications. Therefore, prior NRC approval was not required forimplementation of this modification.

43


EVE 21 A&B CEDM COOLING FAN SETPOINT CHANGE

~811DBIK

This modification changed the time delay setpoint in the startinglogic for the'"Control Element Drive Mechanism (CEDM) cooling fans(HVE-21 A&B) from 10 seconds to 30 seconds. When a fan switch istaken to the start position, one fan starts immediately and then,after the start delay, the standby fan starts if a low flowcondition exists. Before this modification the standby fan wouldfrequently start. This modification prevents the inadvertent startof 'the standby fan when starting a fan from stop but does notaffect the time the standby fan will automatically start afterfailure of a running primary fan.

Safet Evaluation:

This modification is to increase the setpoint of time delay relays62/507 and 62/508 associated with the CEDM Cooling System (HVE-21A&B). The CEDMs are referenced in the FSAR as Not Nuclear Safety.In addition, the change has no effect on CEAs. This modificationdid not constitute an unreviewed safety question or require changesto the plant Technical Specifications. Therefore, prior NRCapproval was not required for implementation of this modification.

SECTION 2

SAFETY EVALUATIONS

45

SAFETY EVALUATION ZPN-PBL-SENS-94-018REVISZON 0

OCHLORITE SYSTEM MODIFICATIONS

~summa

This safety evaluation demonstrated that Hypochlorite (CL) systemmodifications performed in accordance with the guidance of thisevaluation did not, adversely affect plant safety, security oroperation. The CL system is a non-safety system common to St.Lucie Units 1 & 2 that produces a sodium hypochlorite solution viaelectrolytic decomposition of filtered seawater. The hypochloritesolution is periodically injected into the suction side of theintake cooling water (ICW) and circulating water (CW) pumps for thecontrol of biological fouling.

Safet Evaluation:

This safety evaluation addressed the technical and licensingrequirements for the Hypochlorite (CL) system and concluded thatthe proposed plant modifications were bounded by the TechnicalSpecifications and did not change the analysis of accidentsaddressed in the FSAR or the results and conclusions of anyprevious safety evaluations. The actions or changes identified andevaluated in this safety evaluation did not have any adverse effecton plant safety or plant operations. The actions or plant changes(procedures and/or hardware), identified in this safety evaluationdid not constitute an unreviewed safety question or require changesto plant Technical Specifications. Therefore, prior NRC approvalwas not required for implementation of the actions or changesidentified in this safety evaluation.

46

I

SAFETY EVALUATION JPN PSL-SEEP-94-035REVISION 0

UPDATE FSAR LOOP ACCURACIES VARIOUSELECTRICAL DISTRIBUTION SYSTEM METERS

WITH CURRENT INFORMATION

~8llEEIK

This safety evaluation demonstrated the acceptability of updatedLOOP accuracy data for various meters which monitor electricalparameters (voltage, current, frequency, etc.) and are included inTable 7.5-1 ("Safety Related Display Instrumentation" ) of the St.Lucie Unit 2 FSAR. PC/M 428-291M performed a like for likereplacement of GE meters with meters manufactured by Yokogawa.This safety evaluation considered the accuracy values for these newmeters.

Safet Evaluation:

The accuracies for the electrical parameter meters are not used asan assumption in any analyzed accident of the accident analyses.This update for the metering instrumentation neither involves anunreviewed safety question not requires changes to TechnicalSpecifications. ~ Therefore, this change and revision to the FSARwere performed without prior NRC approval.

47

SAFETY EVALUATZON iTPN-PSL-SENP-94-037REVISION

SZT DISCHARGE LOOP CHECK VALVE STROKE TEST

~summar

The purpose of this safety evaluation was to demonstrate theacceptability of performing a partial flow, full stroke test of theSafety Injection Tank (SIT) discharge/loop check valves.Successful performance of the test was used to satisfy NRCrequirements for SIT check valves as delineated in Generic Letter89-04. There are four SITs with two check valves associated witheach tank. The check valves tested were V-3215, V-3225, V-3245,V-3217, V-3227, V-3237, and V-3247. Partial flow, full stroketesting supersedes disassembly and inspection as a method todemonstrate acceptability. Plant restrictions applicable to thetest were identified and listed.

Safet Evaluation:

This safety evaluation addressed the effect of the test on safetyrelated components, including the fuel, steam generator nozzle dams,refueling water cavity clarity, and induced cyclical thermalstresses in the SIT. The actions or plant changes in proceduresidentified in this safety evaluation did not constitute anunreviewed safety question or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for implementation of the actions or changes identifiedwithin this evaluation.

48

SAFETY EVALUATION ZPN-PSL-SENP-94-039REVISION 1

PUMPER LIFTED TEAD FOR PDIS-2216

~Summa rThis safety evaluation demonstrated the acceptability"of applyinga .jumper to the pressure differential switch (PDIS-2216) in thechemical & volume control system (CVCS) letdown line. This switchmeasures the differential pressure across the regenerative heatexchanger as a means of detecting high fluid flow that would occurfrom a downstream line break. The jumper of this switch defeatsclosure on high differential pressure. Without the differentialpressure switch, letdown isolation still occurs from a temperatureelement (TE-2221) located immediately downstream of theregeperative heat exchanger. TE-2221 senses high temperature(470 F) downstream of the regenerative heat exchanger fromexcessive letdown flow resulting from a line break.

Safet Evaluation:

This safety evaluation addressed the technical and licensinga

~requirements for the jumpering of pressure switch PDIS-2216 andconcluded that the proposed plant configuration and mode ofoperation was bounded by the Technical Specifications and did notchange the analysis of accidents addressed in the FSAR or theresults and conclusions of any previous safety evaluations. Theactions or changes identified and evaluated in this safetyevaluation did not have any adverse effect on plant safety or plantoperations. The actions or plant changes in procedures, identifiedin this safety evaluation did not constitute an unreviewed safetyquestion or require changes to plant Technical Specifications.Therefore, prior NRC approval was not required for implementationof the actions or changes identified in this safety evaluation.

49

SAPETY EVALUATION iTPN-PSL-JNO-95-001REVISION 0

NUCLEAR PLANT CHEMISTRY PARAMETERS MANUAL REV. 18

~8UENIThis safety evaluation determined that Revision 18 to the NuclearPlant Chemistry Parameters Manual had no adverse affect on plantsafety.

Safet Evaluation:

This safety evaluation determined that the changes did notadversely affect the safety related equipment, plant operations, orsafety functions. The actions or plant conditions identified inthis safety evaluation did not constitute an unreviewed safetyquestion or require changes to'he plant Technical Specifications.Therefore, prior NRC approval was not required for implementationof the actions or conditions identified within this evaluation.

50

SAFETY EVALUATION i7PN-PSL-SEMS-95-001REVISION 0

LETDOWN PRESSURE CONTROLLER PIC-2201 SET PRESSURE REDUCTION

~Summa rThis safety evaluation demonstrated the acceptability of operatingwith a reduced set pressure as low as 360 psig for St. Lucie Unit2 Letdown Backpressure Controller PIC™2201 under controlledconditions. The normal operating set pressure is 450 psig withcontroller in the automatic mode. The reduced operating setpressure was necessary'o allow diagnostic testing of the LetdownControl Valves LCV-2110 P&Q control modules and valve response withdynamic flow conditions.

The conditions for operating at the reduced pressure were tosupport diagnostic testing of LCV-2110 P&Q and/or the starting orstopping of a second charging pump. An additional restriction wasthat reactor power would be held at steady state conditions. PIC-2201 would have been restored to the normal operating set pressure(450 psig) should these conditions have not been met or if an RCStransient was initiated.

Safet Evaluation:

This safety evaluation considered the effect on the operation ofthe Chemical Volume and Control System with a reduced set pressureon PIC-2201 under controlled conditions. It was determined thatthere were no adverse effects on plant safety under the conditionsdescribed in the safety evaluation. Therefore, the conditions orchanges in procedures and design documents addressed under thisevaluation did not involve an unreviewed safety question or requirea change in plant Technical Specifications. Therefore, prior NRC

approval was not required for implementation of the actions orchanges identified in this safety evaluation.

51

SAPETY EVALUATION i7PN-PSL-SEMS-95-002REVISION 0

OPERATION OP THE UNIT 2 REPUELZNGWATER TANK DURZNG ZNBPECTZONB

~8UBQSEX'his

safety evaluation documented the acceptability of performingan inspection of the inside of the Refueling Water Tank (RWT)during normal plant operations. The inspection by divers of theinside of the RWT was necessitated by failure of a retentionelement in the Unit 2 Spent Fuel Pool Ion Exchanger which resultedin a release of approximately 25 ft of resin which may havetraveled to the RWT and settled to the bottom.

Safet Evaluation:

The evaluation determined that due to the RWT arrangement, pipingsizes, and function an inspection by divers and/or submersiblemechanical devices did not have an adverse effect on plant safety.The plant conditions identified in this safety evaluation did notconstitute an unreviewed safety question or require changes to theplant Technical Specifications. Therefore, prior NRC approval wasnot required for the conditions that were identified within thisevaluation.

52

SAFETY EVALUATION iTPN-PSL-SEMP-95-004REVISION 1

REDUCED PRESSURIZER HEATER CAPACITY

~Summa

This safety evaluation documented the acceptability operating theunit with a reduced pressurizer heater capacity of 1200 Kw. Thispermitted up to 6 pressurizer heaters, a total of 300 Kw, to beremoved from service. The safety evaluation specified, the issuanceof a PC/M to cover the physical modifications and that changes bemade on the affected documents (control wiring diagram, TEDB, andpower distribution data sheets). Table 1 of the safety evaluationi'ncludes the revisions to account for the heaters that have beenretired and to trend heater performance.

Safet Evaluation:

The pressurizer heaters have a total installed capacity of 1500 Kw,300 kw of which are on two proportional heater banks (P-1 and P-2),and 1200 Kw are on six backup heater banks (B-1 to B-6). Becauseof the design margin of the pressurizer heaters, safe plantshutdown and the results of postulated events in the FSAR safetyanalyses are not adversely affected by operation with reduced

- heater capacity. he plant conditions identified in this safetyevaluation did not constitute an unreviewed, safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for the conditions that wereidentified. within this evaluation.

53

SAFETY EVALUATION iTPN-PSL-SENS-95-005,REVISION 0

EVALUATION FOR AN ALTERNATE REACTORCOOLANT GAS VENT SYSTEM ALIGNMENT

~summa

This safety evaluation documented an alternate alignment for thereactor coolant gas vent system (RCGVS). The changes to the systemthat were evaluated are: 1) installation of a blank at the flangedconnection downstream of V1465 for the purpose of arresting theleak through V1465, 2) providing 125V DC safety bus SA controlpower to V1464, and 3) depowering the normal SB bus control wiringto V1464. A blank fabricated from ASTM A-240 Type 316 SS (M&S1032753110) was installed at the flanged location on line 1-RC-852,downstream of V1465.

Safet Evaluation:

The evaluation determined there is no impact on the ability of thesystem to meet the Technical Specification " 3/4.4.10 LCOrequirements. Operability of the RCGVS is not affected since twovent paths are available from both the pressurizer and the reactorvessel head. The plant conditions identified„ in this safetyevaluation did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for the conditions that wereidentified within this evaluation.

54

SAFETY EVALUATION O'PN-PSL-SENS-95-008REVISION 0

DEENERGIZATION OF RAB VENTILATION DAMPERS D-5B & D-7BTO SUPPORT ACTUATOR REMOVAL FROM DAMPER D-7A

~summar

This safety evaluation documented acceptability of deenergizing theactuator for HVAC damper D-7B while performing maintenance on HVACdamper D-7A. HVAC damper D-7A required maintenance that requiredremoving the D-7A actuator. This safety evaluation considered thedeenergization of damper D-7B by pulling of a fuse which alsoprovided power for damper D-5B and the position indicators for bothD-5B & D-7B. Both of these dampers close on receipt. of a SIAS toensure proper HVAC flow to the emergency core cooling system (ECCS)pump area. Since these dampers are deenergize to close, pullingthe fuse ensured they remained in their fail-safe positions(closed) .

Safet Evaluation:

The safety functions of the RAB HVAC systems were not affected bythis change since deenergization resulted in the proper post-accident system configuration. Although normal operation of theHVAC system was affected (loss of normal ventilation flow throughD-5B & C-7B) the impact was insignificant. There were no plantoperating restrictions required as a result of this evaluation.The plant conditions identified in this safety evaluation did notconstitute an unreviewed safety question or require changes to the

~ plant Technical Specifications. Therefore, prior NRC approval wasnot required for the conditions that were identified within thisevaluation.

55

SAFETY EVALUATION J'PN-PSL-SENS-95-010REVZSION 0

EVALUATZON FOR OPERATZON WITH A DIESEL OIL STORAGETANK BUILDING MZSSILE DOOR REMOVED

~summa

This safety evaluation documented acceptability of plant operationwith a single Diesel Oil Storage Tank (DOST) building missile doorremoved. The two St. Lucie Unit 2 DOSTs are housed within aconcrete enclosure which provides missile protection for the DOSTs.Each section of the building has its own access door on the westside of the building which also serves as a missile barrier. These ~

doors have become difficult to operate and require maintenance.This safety evaluation was conducted to support the doormaintenance.

Safet Evaluation:

This evaluation concluded that operation of the plant with a singleDOST building missile door removed did not impact plant safety anddid not constitute an unreviewed safety question nor require achange to the technical specifications. Therefore, prior NRCapproval was not required for the conditions that were identifiedwithin this evaluation.

56

SAFETY EVALUATION JPN-PSL-SENS-95-013REVISION 0

EVALUATION FOR OPERATION WITH DIESEL OIL TRANSFER PUMP

2B DISCHARGE ISOLATION VALVE V17216 CLOSED

~Summar

This safety evaluation documented acceptability of plant operationwith Diesel Oil Transfer Pump (DOTP) 2B discharge isolation valveV17216 in the closed position. Compensatory measures wereestablished to open the valve upon operation of the 2B EmergencyDiesel Generator (EDG). V17216 is normally a LOCKED OPEN valve;however, due to a suspected leak in the underground pipingdownstream of the valve it was desired to isolate the piping,identify the leak, and make repairs.

Safet Evaluation:

This evaluation concluded that operation of the plant with valve1I216 in the closed position did not impact plant safety and didnot constitute an unreviewed safety question nor require changes tothe plant Technical Specifications. Therefore, prior NRC approvalwas not required for the conditions that were identified withinthis evaluation.

57

SAFETY EVALUATION iTPN-PSL-SENS-95-021REVISION 1

FUEL HANDLING E UIPMENT FSAR CHANGES

~BMUBRX'his

safety evaluation justified and provided for Unit 2 FSARchanges identified as a result of an FSAR review conducted by theFPL Quality Assurance Department and documented in Audit No.QSL-OPS-94-24.

Saf et Evaluation:

This safety evaluation demonstrated that the FSAR changes providedin the FSAR Change Package did not adversely affect plant safety,security or operation. The FSAR changes considered in this safetyevaluation did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for the conditions that wereidentified within this evaluation.

58

SAFETY EVALUATION a7PN-PSL-SENP-95-021REVISION 0

DELETION OF SR-A1A EMBANKMENT SURVEY AND ANNUAL AERIALPHOTOGRAPH OF BEACH AREA COMMITMENTS AT ST LUCZE SITE

~8ummar

This safety evaluation provided justification for deletion ofregulatory commitments to visually inspect the SR-A1A highwayembankment following passage of any major storm and to perform anannual aerial photograph of the beach area. Additionally,, thissafety evaluation provided the revision to the FSAR that documentedthese commitment changes.

Safet Evaluation:

These commitments were not required to achieve compliance with anyrule, regulation or NRC order, were not made to minimize recurrenceof an adverse condition, and were not relied upon in the originalerosion analysis to assure protection to the plant from designbasis flooding. The FSAR changes identified in this safetyevaluation did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for the conditions that wereidentified within this evaluation.

59

SAFETY EVALUATION JPN-PSL-SEFJ-95-022REVISION 0

TEMPORARY USE OF DUMMY FUEL ASSEMBLY SKELETON REPLACING ACELL BLOCKING DEVICE FOR THE SPENT FUEL STORAGE RACK

~summar

This safety evaluation documented acceptability of continued SpentFuel Pool {SFP) operation with a dummy fuel assembly skeleton usedas a physical barrier in place of a cell blocking device. A cellblocking device of the Region 2 storage cell, T15, wasinadvertently knocked out of position during a recent SFP resincleanup vacuuming. The blocking device prevented placing a fuelassembly in storage cell T15. A dummy fuel assembly skeleton wasplaced in cell T15 to prevent inadvertent use of the cell for fuelstorage.

Safet Evaluation:

This safety evaluation discussed th'e SFP criticality physical modeand the effect of storage of a fuel assembly skeleton on rackreactivity. The evaluation concluded that the use of the dummyfuel assembly skeleton serving as a cell blocking device istechnically justifiable. The plant conditions identified in, thissafety evaluation did not constitute an unreviewed safety questionor require changes to the plant Technical Specifications.Therefore, prior NRC approval was not required for the conditionsthat were identified within this evaluation.

60

SAFETY EVALUATION JPN PSL-SEMS-95-024REVISION 1

ECCS HEADER 2A2 PRESSURIRATION DUE TOBACK LEAKAGE THROUGH V3217

~mmmm ar

This safety evaluation evaluated the safety significance of theback leakage through RCS loop 2A2 ECCS cold leg injection headercheck valve V3217 and to allow the upstream piping segment toremain pressurized to RCS pressure until the next refueling outagein approximately 3 weeks. The configuration analyzed by thisevaluation did not alter the RCS pressure boundary. The segment ofpiping pressurized to RCS pressure simply captured check valveleakage from V3217 which was determined to be within the TechnicalSpecification limit of 1 gpm. The evaluation also included the useof a temporary pressure gage for monitoring purposes.

Safet Evaluation:

The safety evaluation considered the effects on RCS pressureboundary, piping design, ECCS operation, RCS leakage monitoring,containment integrity, and design integration. The plantconditions identified in this safety evaluation did not constitutean unreviewed safety question or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for the conditions that were identified within thisevaluation.

61

SAFETY EVALUATION JPN-PSL-SEZP-95-031REVISION 0

TEMPORARY USE OF AN ACOUSTIC FLON METER TO CORRECTTHE DDPS FEED%PATER FLOP/ COEFFICIENT

~summa rThis safety evaluation documented acceptability of using theDigital Data Processing System {DDPS) flow coefficients used in thecalorimetric calculation to be corrected by the Leading EdgeFlowmeter {LEFM) measurements. The feedwater flow measurement hasbeen determined to be in error due to fouling within the venturiused to measure differential pressure. The LEFM was installedexternally to the feedwater line where it is not subject to errorscaused by pipe fouling. The more accurate flow readings provide abetter calorimetric calculation.

Safet Evaluation:

The procedure covered by this safety evaluation did not change theintent of the feedwater flow measurement for either operator reviewor use in the calorimetric equation. The transmitters providingdifferential pressure across the feedwater venturi continue to bethe primary instrument. providing indication. This evaluation doesnot change any mechanical components within the Feedwater .System.The thermohydraulic parameters were changed on the secondary side,but remained within bounds of those established for Stretch Poweroperation. The plant conditions identified in this safetyevaluation did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for the conditions that wereidentified within this evaluation.

62

SAFETY EVALUATION ZPN-PSL-SENS-95-031REVISION 0

TEMPORARY REMOVAL OF A CCW BUILDING MISSILE BARRIER

~smamar

This safety evaluation documented acceptability of plant operationin MODE 5 or MODE 6 with one of the missile shield doors in theComponent Cooling Water (CCW) building removed to performmaintenance activities. The function of the missile shield doorsis to protect the CCW components during a hurricane/tornado fromdesign basis missiles.

Safet Evaluation:

This safety evaluation demonstrated that temporarily removing a CCWbuilding missile shield door while in MODE 5 or 6 did notconstitute an unreviewed safety question or require changes to theplant Technical Specifications. Therefore, prior NRC approval wasnot. required for the conditions that were identified within thisevaluation.

63


REFUELING OPERATIONS NITH A STUCK REACTOR VESSEL STUD

~mamma

This safety evaluation documented acceptability of leaving stuckreactor vessel stud f44 in place during refueling and until it canbe replaced during the next scheduled refueling outage. Reactorvessel stud f44 stuck during refueling. This safety evaluationallowed the stuck stud to be exposed to a 'borated water environment(refueling concentration) for up to three weeks.

Safet Evaluation:

The FSAR description regarding the removal of the reactor vesselstuds is with respect to corrosion concerns. A technicalevaluation of the stuck stud concluded that the expected corrosioneffects were minimal over a three week period. The plantconditions identified in this safety evaluation did not constitutean unreviewed safety question or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for the conditions tha't were identified within thisevaluation.

64

, SAFETY EVALUATION FRG f95-34

OPERATION OF NEW FUEL HANDLING CRANE FOR MAINTENANCE

~summa rThis safety evaluation documented acceptability of operating theNew Fuel Handling Crane for maintenance activities when no new fuelis stored in the Fuel Handling Building. The FSAR states that thiscrane is used exclusively for new fuel handling and is locked inposition when fuel handling is not in progress. Several of thelights in the Fuel Handling Building had burned out bulbs whichcould only be accessible through the use of the installed new fuelhandling crane.

Safet Evaluation:

The safety evaluation concluded that this change in operation ofthe New Fuel Handling Crane while there is no fuel in the new fuelstorage area of the Fuel Handling Building does not adverselyaffect plant safety, security or operation, does not constitute anunreviewed safety question, nor does it require changes to theTechnical Specifications. Therefore, prior NRC approval was notrequired for the conditions that were identified within thisevaluation.

65

SAFETY EVALUATION JPN-PSL-SEES-95-034REVISION 0

EVALUATION FOR THE PROVISIONS TO TRIP EMERGENCY DIESELGENERATOR OUTPUT BREAKER ON CIAS IN PLANT MODES 5 AND 6

~Summa rThis safety evaluation provided the basis for a plant temporarymodification (plant MODES 5 and 6 only), involving tripping of anemergency diesel generator (either 2A -or '2B) output breaker on aCIAS, which will eliminate potential for equipment damage if anactual or a spurious CIAS occurs while the EDG being tested isconnected to the grid. In October, 1995 an event occurred where anEDG was being test run in parallel with the grid when an unexpectedCIAS signal (without SIAS)'caused the EDG to be motorized by off-site power. This change prevented possible EDG damage undersimilar circumstance.

Safet Evaluation:

This change did not affect plant operating practices and enhancedEDG protection in MODES 5 and 6. This evaluation concluded thatthis modification did not constitute an unreviewed safety questionor require changes to the plant Technical Specifications.Therefore, prior NRC approval was not required for the conditionsthat were identified within this evaluation.

66

SAFETY EVALUATION FRG-95-36

BYPASSING AUTOMATIC ESFAS ACTUATION DURINGOPERATING MODES 5 AND 6

~Summar

This safety evaluation documented acceptability of operating theplant in cold shutdown (MODE 5) and refueling (MODE 6) with SIAS,MSIS, CSAS, RAS and the high containment pressure initiated CIASdisabled as long as the administrative controls .placed on. theadditional sets of ESFAS bypass keys are adequate to ensure that nomore than one channel can be bypassed in MODES 1 through 4.Additional sets of bypass keys were used to bypass the desiredportions of the Engineered Safety Features Actuation System. Theadditional keys are maintained under strict administrativecontrols, providing assurance that they were removed prior toentering MODE 4 and were not available to the operators during thetime the unit is not in MODES 5 OR 6.

Safet Evaluation:

This change in the operation of ESFAS does not adversely affectplant safety, security or operation, does not constitute anunreviewed safety question, nor does it require changes to theTechnical Specifications. Therefore, prior NRC approval was notrequired for the conditions that were identified within thisevaluation.

67


EVALUATION FOR CROSS-CONNECTING 480VLOAD CENTERS DURING MODES .5 AND 6

~summa rThis safety evaluation documented acceptability of operating the Aand B trains of the safety related 480V electrical system crossconnected with one of the safety 4.16 kV busses (2A3 or 2B3) out-of-service for required maintenance during MODES 5 or 6.

Safet Evaluation:

This. safety evaluation demonstrated that operating the' and B

trains of the safety related 480V electrical system cross connectedwith one of the safety 4.16kV busses (2A3 or 2B3) out-of-servicefor required maintenance while in MODES 5 or 6 neither involves anunreviewed safety question nor requires a change to plant TechnicalSpecifications. Therefore, prior NRC approval was not required forthe conditions that were identified within this evaluation.

68

SAFETY EVALUATION ZPN-PSL-SENS-95-04l.REVISION 0

TEMPORARY REMOVAL OF DIESEL OIL LINETORNADO MISSILE BARRIER

~Summa@

This safety evaluation documented acceptability of plant operationin MODE 6 with one section of the Diesel Oil transfer line tornadomissile barrier removed to facilitate inspection and/or repair ofthe 2A/2B diesel oil transfer line. The discharge lines from thefuel oil transfer pumps exit. the Diesel Oil Storage Tank (DOST)above grade then run underground to the Diesel Generator buildingwhere they rise above grade and enter the Diesel GeneratorBuilding. Both the above grade and underground portions of thelines are protected from design basis tornado missiles. Thisevaluation concluded that the missile barriers are designed tosupport maintenance, inspection, and repair on the fuel oil linesand that temporary removal of that section during MODE 6 isacceptable.

Safet Evaluation:

This safety evaluation demonstrated that temporarily removing thediesel oil transfer line missile lid while in MODE 6 neitherinvolves an unreviewed safety question nor requires a change toplant Technical Specifications. Therefore, prior NRC approval wasnot required for the conditions that were identified within thisevaluation.

69


EVALUATION FOR THE TEMPORARY INSTALLATION OFCOUBTICAL MONITORING E UIPMENT

~mumm ar

This safety evaluation documented acceptability of mountingacoustic sensors to the pressurizer code safety valves (SRVs),adjacent piping and additional locations as approved by Engineeringin order to monitor for potential valve seat leakage. The sensorsand/or their associated mounting studs may remain installed asnecessary to support data acquisition through the current operatingcycle, at which time they will be removed.

Safet Evaluation:

This evaluation concluded that the ,operation of the plant asdescribed does not represent and unreviewed safety question, doesnot require a change to plant Technical Specifications and does notadversely affect plant operation or safety. Therefore, prior NRCapproval was not required for the conditions that. were identifiedwithin this evaluation.

70

SAFETY EVALUATION ZPN-PSL-SEEP-95-045REVISION 0

POST ACCIDENT MONITORING INSTRUMENTATIONCOVERED BY TECHNICAL SPECIFICATIONS

~summar

This safety evaluation documented acceptability of updatedinformation for Table 7.5-1 ("Safety Related DisplayInstrumentation" ) of the St. Lucie Unit 2 FSAR, as to whichinstruments are used for post accident monitoring. The new FSARinformation is consistent with that provided to the NRC. No plantconfiguration changes were made or analyzed by this safetyevaluation.

Safet Evaluation:

This evaluation determined that the updated information, specifyingthe post accident monitoring instrumentation, neither involved anunreviewed safety question nor required a change to the TechnicalSpecifications. Therefore, this update to the FSAR may beperformed without prior NRC approval.

71

SAFETY EVALUATION JPN-PSL-SENP-95-046REVISION 0

EVALUATION FOR THE REMOVAL OF THE UNIT 2PRESSURIZER MISSILE SHIELD ROOF

operating pressure

Safet Evaluation:

~sum'mar 'h

This safety evaluation documented acceptability of permanentlyremoving the St. Lucie Unit 2 pressurizer missile shield roof. Thepurpose of the pressurizer missile shield roof is to preventcredible pressurizer missiles from penetrating and puncturing thecontainment vessel. The containment vessel is a 2" thick steelright hand, circular cylinder with a 1" thick steel hemisphericaldome. The evaluation concluded that there does not exist anycredible missile which require the pressurizer missile shield roofto protect the containment vessel. The plant benefits identifiedwith removal of the pressurizer missile shield roof include thereduction in 'risk from dropping a heavy load, reduction in thermalaging effects on components located in the pressurizer cubicle andit facilitates, inspection activities while the unit is at normal

and temperature.

The safety evaluation concluded that the removal of the pressurizermissile shield roof does not adversely affect safe operation of theplant and does not constitute an unreviewed safety question anddoes not require a change 'o the Technical Specifications.Therefore, prior NRC approval was not required for the conditionsthat were identified within this evaluation.

72

SAFETY EVALUATION FRG 95-48

JUMPER LIFTED LEAD 2-95-003

~Summar

This safety evaluation documented acceptability of jumpering andlifting lead 2-95-008 in order to eliminate a ground on thepressurizer heater back-up B-1 power panel 226.

Safet Evaluation:

The safety functions of the pressurizer heaters were not affectedby this change. There were no plant operating restrictionsrecgxired as a result of this evaluation. The plant conditionsidentified in this safety evaluation did not constitute anunreviewed safety questi*on or require changes to the plantTechnical Specifications. Therefore, prior NRC approval was notrequired for the conditions that were identified within thisevaluation.

73

SAFETY EVALUATION JPN PSL-SENP-9S-094REVISION 0

DELETION OF THE CONDENSATE DEGASIFIER

~Summa rThis safety evaluation documented acceptability of removing thecondensate degasifier from the Unit 2 turbine building. Thedegasifier was ineffective in removing gases from demineralizedwater and was abandoned in place since startup of Unit 2.

Safet Evaluation:

This safety evaluation concluded that the removal of the condensatedegasifier did not represent an unreviewed safety question, did notrequire a change to plant Technical Specifications, and did notadversely affect plant operation or safety. Therefore, prior NRCapproval was not required for the conditions that were identifiedwithin this evaluation.

74

SAFETY EVALUATION ZPN-PBL-SENP-95-096REVISION 0

CLASSIFICATION OF EMERGENCY DIESEL GENERATOR AUXILIARIESAND FUEL OIL PIPING ASME DESIGN UALZTY GROUP

~summa

This safety evaluation documented acceptability of reclassifyingthe emergency diesel generator (EDG) auxiliaries and fuel oilsystem as owner optional upgrade Quality Group C. These componentsare the EDG air start system, the EDG fuel oil system, the EDG lubeoil system, the EDG cooling water system, and the intake andexhaust system. These systems were designed and constructed to ahigher ASME Quality Classification to increase reliability,however, there were no provisions made to allow for individualcomponent ZST testing which was optional since they were ownerupgrades. These sub-components of the EDG sets are tested when theEDGs pass their Technical Specifications required tests.

Safet Evaluation:

This evaluation concluded that operation of the plant with the ASME

Quality Group Reclassification of the EDG auxiliaries and fuel oilsystems did not represent an unreviewed safety question and did notrequire a change to plant Technical Specifications and did notadversely affect plant operation or safety. Therefore, prior NRC

approval was not required for the conditions that were identifiedwithin this evaluation.

75

SAFETY EVALUATION JPN FRG 95-145

USING COMPUTER SOFTWARE PROGRAM AND ASSOCIATED SENSORSFOR TESTING SETTING AND CALIBRATING PRIMARY

RELIEF VALVES AND OTHER RELIEF VALVES

~Summar

This safety evaluation documented acceptability of using a computersoftware program and associated sensors and electronics for thepurpose of testing, setting and calibrating primary safety reliefCorporation was verified by Dunn's Valve Tester's Inc. under theoversight of representatives of the St. Lucie I&C Department andthe Mechanical Maintenance Engineering Group.

Safet Evaluation:

This safety evaluation concluded that the use of this computersoftware did not constitute an unreviewed safety question orrequire changes to the plant Technical Specifications. Therefore,prior NRC approval was not required for the conditions that wereidentified within this evaluation.

76

SAFETY EVALUATION JPN-PSL-SESP-96-054REVISION 0

EVALUATION OF PRESSURIZED THERMAL SHOCK OF REACTORVESSEL BELTLINE MATERIALS FOR ST LUCIE UNITS 1 & 2

~Summar:

This safety evaluation and its attachments determined the EOL RTpys

values for the St. Lucie Unit 1 and 2 reactor vessel beltlinematerials. This information is required for 10 CFR 50.61submittal. The EOL RT~s values have been calculated for both St.Lucie Units 1 and 2 and found to be acceptably below the 10 CFR'50.61 screening limit. Specifically the limiting material at St.Lucie Units 1 and 2 have an EOL PTS values of 213 F and 160 F

respectively compared to a 270 F limit.

Safet Evaluation:

This evaluation does not involve an unreviewed safety questionbecause it does not involve a change: to the plant TechnicalSpecifications; to plant equipment or procedures; and does notchange the plant's licenses. Furthermore, the plants beltlinematerials meet the 10 CFR 50.61 requirements. Therefore, prior NRC

approval was not required for the conditions that were identifiedwithin this evaluation.

77

SECTION 3

RELOAD SAFETY EVALUATIONS

78


LUCIE UNIT 2 CYCLE 9 RELOAD

~summa rThis engineering package provided the reload core design of the St.Lucie Unit 2 Cycle 9. The Cycle 9 core is designed for cyclelengths up to 13,009 and 12,627 EFPH, depending upon variation inthe cycle 8 length of between 12,284 and 11,284 EFPH, respectively.The cycle lengths for Cycle 9 included an end of cycle inlettemperature coastdown to 535 F followed by a coastdown in power toapproximately 854 power. Cycle 8 is expected to reach an EOC

exposure of approximately 12,100 EFPH.

The primary design change to the core for Cycle 9 is thereplacement of 84 irradiated fuel assemblies with fresh Region Lfuel assemblies. The fuel is arranged in a low leakage patternwith no significant differences between the Cycle 9 loading patternand the Cycle 8 design. The mechanical design of Region L fuel isthe same as that of Region K (Cycle -8) and Region J (Cycle 7)reload fuel, except for the following primary changes:

a) Gadolinia Burnable Absorber (with increased cutback lengthon poison rods) in lieu of A1~03-B<C Burnable Absorber.

b) Changeover from TIG to laser welded Zircaloy intermediatespacer grids.

Safet Evaluate.on:

The safety analysis of this design was performed by Asea BrownBoveri Combustion Engineering Nuclear Operations (ABB CENO) andindependently reviewed by Florida Power and Light Co. It has beendetermined that the operation of the Cycle 9 reload core does notpose an unreviewed safety question and can be implemented with nochanges to the St. Lucie Unit 2 Technical Specifications.Therefore, prior NRC approval was not required for implementationof this modification.

79

unit docket number experiments made as 23'994 period

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