ML17228B550

Changes,Tests & Experiments Made as Allowed by 10CFR50.59 for Period 940423-960105.

ML17228B550
Person / Time
Site:	Saint Lucie
Issue date:	01/05/1996
From:	FLORIDA POWER & LIGHT CO.
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ML17228B549	List: ML17228B548 ML17228B550
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NUDOCS 9607110321
Download: ML17228B550 (110)

Text

ST. LUCIE UNIT 2 DOCKET NUMBER 50-389 CHANGES, TESTS AND EXPERIMENTS MADE AS ALLOWED BY 10 CFR 50.59 FOR THE PERIOD OF APRIL 23'994 THROUGH JANUARY 5g 1996 9607ii032i. 960705 05000389 PDR ADOCK R PDR

0 INTRODUCTION This report is submitted in accordance with 10 CFR 50.59 (b), which requires that:

i) changes in the facility as described in the SAR ii) changes in procedures iii) tests and experiments as described in the SAR not described in the SAR which are conducted without prior Commission approval be reported to the Commission in accordance with 10 CFR 50.59(b) and 50.71(e) (4). This report is intended to meet this requirement 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 as described in the Updated Final Safety Analysis Report (FSAR) performed by a Plant Change/Modification (PC/M); the second, changes to the facility or procedures as described in the Updated FSAR not performed by a'C/M and tests and experiments not described in the Updated FSARg the third, a summary of any fuel reload safety evaluations.

TABLE OF CONTENTS S ECTION 1 PLANT CHANGE MODIFICATIONS PAGE 114-985 GOULDS CENTRIFUGAL PUMP OZL SEALS REA-SLN-85-130 003-287 BECKMAN WASTE GAS SYSTEM OXYGEN ANALYZER 10 REPLACEMENT 025-287 DIESEL OIL TRANSFER PUMP MECHANICAL SEAL REPLACEMENT 143-289 RAB MAINTENANCE WORK'REA GANTRY 12 016-290 CCW HEAT EXCHANGERS SHELL DRAIN ADDITIONS 13 178-290 ZCW PUMPS 2B 6 2C SELF LUBRICATING MODIFICATION 14 054-293 REPLACEMENT OF EXCORE NEUTRON FLUX MONITORING SYS as 081-993 EMERGENCY COMMUNICATIONS SYSTEM UPGRADE 16 132-293 CV2 RELAY REPLACEMENT 17 173-293 MV-08-12 6 MV-08-13 MODIFICATIONS 18 025-294 WASTE GAS ANALYZERS 19 068-294 WIDE RANGE STEAM GENERATOR LEVEL UPGRADE 20 104-294 DELETION OF HPSI/LPSZ LO FLOW ALARMS 21 111-294 SAFEGUARDS BYPASS DISPLAY INDICATION PANEL 22 MODIFICATION (A7tB7) 125>>294 CONTAINMENT VACUUM HI ALARM SETPOINT CHANGES 23 148-294 MODIFY 3Y RELAYS IN AB BUS FOR ICW/CCW 2C PUMPS 24 008-295 REPLACEMENT OF THE EXCORE NEUTRON FLUX MONITORING 25 AND PROTECTIVE SYSTEM (NZ DRAWERS) FOR THE RPS SYSTEM 023-295 NRC GEN LTR 89-10 MOV CONTROL SWITCH SETTINGS 26 027-295 PZR LI{}UIDSPACE INSTRUMENT NOZZLES REPLACEMENT 27 029-295 LETDOWN FLOW CONTROL LOOP TUNING 28

SECTION 1 PLANT CHANGE MODIFICATIONS (Continued) PAGE 0 35-295 COND TUBE 'CLEANING & DEBRIS FILTER SYS PHASE I 29 036-295 DEBRIS FZLTER & CONT TUBE CLEANING SYS PHASE H

IZ 30 063-295 HP TURBINE BLADE RING CHANGEOUT 31 096-295 REPLACEMENT OF 2A & 2B UNDERGROUND DIESEL FUEL OZL 32 TRANSFER LINES 132-295 V3439 & V3507 RELIEF VALVE REPLACEMENT 33 133-295 LOWER LETDOWN BACKPRESSURE CONTROL SETPOZNT OF V2345 34 135-295 THERMAL RELIEF VALVE BLOWDOWN MODIFICATZON 35 136-295 SAFETY RELIEF VALVE V3417 SETPOINT AND BZOWDOWN 36 MODIFICATION 152-295 RCS HOT LEG INSTRUMENT NOZZLE REPLACEMENT 37 156-295 DELETION OF EDG AUTO START ON CIAS & CSAS 38 161-295 RCP VAPOR SEAL LEAKOFF LINE MOD 39 167-295 DRILLING OF VALVE DISC V3481 40 168-295 CONTROL LOGIC MOD

'3545 41 194-295 ZCW & CCW LOCAL PUSHBUTTON STATION REMOVAL 42 216-295 EMERGENCY DIESEL GENERATOR LOW COOLING PRESSURE 43 ALARM SETPOZNT 245-295 HVE 21 A&B CEDM COOLING FAN TIME DELAY 44 SETPOINT CHANGE

5 SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SENS-95-032 REFUELING OPERATIONS WITH A STUCK REACTOR 64 VESSEL STUD FRG 95-34 OPERATION OF NEW FUEL HANDLING CRANE FOR 65 MAINTENANCE SEES-95-034 EVALUATION FOR THE PROVISIONS TO TRIP EMERGENCY 66 DIESEL GENERATOR OUTPUT BREAKER ON CZAS IN PLANT MODES 5 AND 6 FRG 95-36 BYPASSING AUTOMATIC ESFAS ACTUATION DURING 67 OPERATING MODES 5 AND 6 SENS-95-037 EVALUATION FOR CROSS-CONNECTING 480V LOAD CENTERS 68 DURING MODES 5 AND 6 SENS-95-041 TEMPORARY REMOVAL OF DIESEL OZL LINE TORNADO 69 MISSILE BARRIER SENS-95-043 EVALUATION FOR THE TEMPORARY INSTALLATION OF 70 ACOUSTICAL MONITORING EQUIPMENT SEEP-95-045 POST ACCIDENT MONITORING INSTRUMENTATION COVERED 71 BY TECHNICAL SPECIFICATION SENP-95-046 EVALUATION FOR THE REMOVAL OF THE UNIT 2 72 PRESSURIZER MISSILE SHIELD ROOF FRG 95-48 JUMPER/LIFTED LEAD 2-95-008 73 SENP-95-094 DELETION OF THE CONDENSATE DEGASIFIER 74 1

SENP-95-096 CLASSZFZCATZON OF EMERGENCY DIESEL GENERATOR 75 AUXILIARIES AND FUEL OIL PIPING ASME DESIGN, QUALITY GROUP FRG 95-145 USING COMPUTER SOFTWARE PROGRAM AND ASSOCIATED 76 SENSORS FOR TESTINGS SETTING AND CALIBRATING PRIMARY RELIEF VALVES AND OTHER RELIEF VALVES SESP-96-054 EVALUATION OF PRESSURIZED THERMAL SHOCK 77 (10 CFR 50.61) OF REACTOR VESSEL BELTLINE MATERIALS FOR ST LUCIE UNITS 1 AND 2

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SECTION 2 SAFETY EVALUATIONS PAGE SENS-94-018 HYPOCHLORITE SYSTEM MODIFICATIONS 46 SEEP-94-035 UPDATE FSAR LOOP ACCURACIES/VARIOUS ELECTRICAL 47 DISTRIBUTION SYSTEM METERS WITH CURRENT INFORMATION SENP-94-037 SIT DISCHARGE/LOOP CHECK VALVE STROKE TEST 48 SENP-94-039 JUMPER/LIFTED LEAD FOR PDIS-2216 49 JNO-95-001 NUCLEAR PLANT CHEMISTRY PARAMETERS MANUALS 50 REV 18 SEMS-95-001 LETDOWN PRESSURE CONTROLLER PZC-2201 SET 51 PRESSURE REDUCTION SEMS-95-002 OPERATION OF THE UNIT 2 REFUELING WATER TANK 52 DURING INSPECTIONS SEMP-95-004 REDUCED PRESSURIZER HEATER CAPACITY 53 SENS-95-005 EVALUATION FOR AN ALTERNATE REACTOR COOLANT 54 GAS VENT SYSTEM ALIGNMENT SENS-95-008 DEENERGZZATZON OF RAB VENTILATION DAMPERS 55 D-5B 6 D-7B TO SUPPORT ACTUATOR REMOVAL FROM DAMPER D-7A SENS-95-010 EVALUATION FOR OPERATION WITH A DIESEL OZL 56 STORAGE TANK BUILDING MISSILE DOOR REMOVED SENS-95-013 EVALUATION FOR OPERATION WITH DIESEL OIL '7 TRANSFER PUMP 2B DISCHARGE ISOLATION VALVE V17216 CLOSED SENS-95-021 FUEL HANDLING EQUIPMENT FSAR CHANGES 58 SENP-95-021 DELETION OF SR-A1A EMBANKMENT SURVEY AND 59 ANNUAL AERIAL PHOTOGRAPH OF BEACH AREA COMMITMENTS AT ST. LUCIE SITE SEFJ-95-022 TEMPORARY USE OF DUMMY FUEL ASSEMBLY SKELETON 60 REPLACING A CELL BLOCKING DEVICE FOR THE SPENT FUEL STORAGE RACK SEMS-95-024 ECCS HEADER 2A2 PRESSURIZATION DUE TO BACK 61 LEAKAGE THROUGH V3217 SEIP-95-031 TEMPORARY USE OF AN ACOUSTIC FLOW METER TO 62 CORRECT THE DDPS FEEDWATER FLOW COEFFICIENT SENS-95-031 TEMPORARY REMOVAL OF A CCW BUILDING MISSILE 63 BARRIER

I SECTION 3 RELOAD SAFETY EVALUATIONS 112-295 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 r This modification consisted of replacement of lip type oil seals on small centrifugal pumps throughout Units 1 and 2 with labyrinth type seals. This package is classified as non-nuclear safety related for all pumps listed with the exception of the Boric Acid Makeup (BAM) and Fuel Pool pumps on both Unit 1 and 2 and the Diesel Oil Transfer pumps on Unit 2 which are classified as nuclear safety related. Revision One of this modification provided the additional details for replacement of the existing Trico oilers with standard sight glasses. The sight glasses are required to adequately monitor oil levels.

Safet Evaluate.on:

The oil seals and level gauges are located in or attached to the bearing housings of those pumps and do. not affect any pressure boundary portions of the pumps. The use of the labyrinth seals provides a superior sealing system in that the stationary and rotating members do not touch as with the existing lip seals.

Therefore, the increased likelihood for leakage caused by rubbing of the shaft with subsequent localized shaft wear does not occur.

Thus the improved design significantly reduces the likelihood of oil leakage along the pump shaft and pump reliability is increased, The oil level sight glasses utilize the same materials anda connecting mechanisms as the existing oilers while providing method of monitoring'il leaks over the operating band. These sight glasses therefore do not increase the likelihood of oil leakage. Therefore with respect to 10 CFR 50.59 the use of the labyrinth type seals and oil level gauge; does not increase the to probability of an accident or malfunction of equipmentnotimportant previously safety, does not create possible accident scenarios addressed by the Safety Analysis Report and does not affect or require changes to the Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

PLANT CHANGE/MODIFICATION 003-287 BECKMAN WASTE GAS SYSTEM OXYGEN ANALYZER REPLACEMENT

~summar The Waste Decay Tank oxygen analyzers continuously monitor the oxygen levels in the Waste Decay Tanks. This modification replaces the existing Waste Decay Tank oxygen analyzers with newer model oxygen analyzers. The replacement oxygen analyzers are more reliable and provide an additional advantage in that the sensing

,element is suitable for sampling oxygen in either a liquid or gaseous sample environment.

Safet Evaluation:

The replacement oxygen analyzers perform the same function and have the essentially the same characteristics as the previous oxygen analyzers. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety analysis Report is not increased as a result of this modification. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

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PLANT CHANGE/MODIFICATION 02S-287 DIESEL FUEL OIL TRANSFER PUMP MECHANICAL SEAL REPLACEMENT

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

Safet Evaluation:

This modification provides for replacement of the mechanical seal which utilizes a pressure retaining gland plate. The specified seal complies with ASME code Section III Class 3 requirements and therefore, has been designed to the'ame criteria as the pump assemblies. In addition, since the size and mass of the new seal gland plate is the same as the existing seal gland plate and the mechanical seal is of similar mass to the existing seal, the effect of the new seal on the seismic response of the pump is infinitesimal. Thus the Seismic Class I qualification of the pump is retained. This modification did not, constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

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I PLANT CHANGE/MODIFICATION 143-289 RAB MAINTENANCE WORK AREA GANTRY

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

Saf et Evaluation:

The gantry is used to facilitate maintenance activities in the Radiation Control Area (RCA) and does not perform any safety related function. However, this modification required the installation of concrete expansion anchors in the Seismic Class I RAB structure. Therefore, this modification has been classified Quality Related. This modification did 'not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification'.

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I J

PLANT CHANGE/MODIFICATION 016-290 COMPONENT COOLING RATER HEAT EXCHANGERS SHELL DRAIN ADDITIONS

~summar This modification adds two six inch flanged pipe stub drains to the underside of each component cooling water heat exchanger shell to reduce drainage time and to improve flushing effectiveness.

Safet Evaluation:

The new drains perform no active safety related function, only the passive function of retaining the pressure boundary integrity of the component cooling water system. This modification is Nuclear Safety Related since heat exchangers which it affects the pressure boundary of the CCW are Quality Group C, Seismic Components.

This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.

Therefore, prior NRC approval was not required for implementation of this modification.

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PLANT CHANGE/MODIFICATION 178-290 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 low lube water flow alarm functions which are no longer required. This modification was previously installed on ICW pump 2A and has proven satisfactory. The modification removes the ICW pump shaft enveloping tube thus exposing 'the bearings to the process stream which serves as their lubricant. Additionally, the existing bearings are replaced with bearings of a more suitable design and material for this service. A portion of ~the external lube water system piping to the supply side of the upper and lower bearings of the ICW Pump is removed and the remaining piping is blanked off by stainless steel blind flanges. I Safet Evaluation:

The Design Analysis for ICW pumps 2B and 2C self lubrication modification was verified by reference to design documents, as

.compared with and substantiated by design inputs. The plant Technical Specifications were reviewed to ensure that no change to plant Technical-Specifications were involved. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

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0 ,~

PLANT CHANGE/MODIFICATION 054-293 REPLACEMENT OF EXCORE NEUTRON FLUX MONITORING SYSTEM

~summa'his modification replaces the existing excore monitoring system due to obsolescence and availability of spare parts. The replacement involves the following: 1) replacing the two excore detectors and in-containment cables with Gamma-Metric detectors and cabling, 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 excore monitoring system meets the design requirements of Appendix R and RG 1.97. The new excore monitoring system maintains its original two redundant class 1E channel design and function.

Safet Evaluation:

The design bases and original function for the excore neutron flux monitoring system as described in the FSAR remain unaffected by the replacement system. The design of the new system meets Appendix R and RG 1.97, Rev. 3 Category 1, Type B variable. New failure modes are not created by the replacement excore monitoring system. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this'odification.

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J PLANT CHANGE/MODIFICATION 083.-993 EMERGENCY COMMUNICATION SYSTEMS UPGRADE

~Summa This modification upgrades the plant emergency communication systems to incorporate lessons learned from Turkey Point during hurricane Andrew. Its primary puxpose is to ensure that offsite communications to county, state, and federal agencies and FPL departments is maintained during and after a hurricane. The design wind 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. A new cable raceway, antenna, and mast are also added to the system.

The upgraded system is accessible from the both control rooms, the Technical Support Center (TSC), and the Emergency Offsite Facility (EOF) .

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

Permanent cellular telephone systems are also added. Each base station is parallel connected to a new telephone in its respective control room and the TSC. The antennas for these systems are installed on the auxiliary building roofs and rated for 125 mph winds.

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

Safet Evaluation:

The new hardware and LGR modifications pxovide a diverse offsite communication and notification capability designed to function under hurricane foxce wind conditions. The modifications did not constitute an unreviewed safety question or xequire changes to the plant Technical Specifications. Prior NRC approval was not, required for implementation.

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

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PLANT CHANGE/MODIFICATION 132-293 CV-2 RELAY REPLACEMENT

~slam ar This Engineering Package (EP) provides the design necessary to replace the existing Westinghouse type CV-2 undervoltage relays in the 480 Volt PSB-1 cabinets with the solid state relays manufactured by ABB, type 27N. These relays are a part of the undervoltage/degraded voltage (PSB-1) protective scheme. This modification was requested by the System protection and Electrical Maintenance personnel to enhance the outage calibration activities.

In addition, this EP adds to each of the existing 4.16 kV undervoltage protective relaying schemes a timer relay to provide adequate time for the test circuit to change state before re-arming the trip circuit.

Safet Evaluation:

The new hardware and the modifications to the undervoltage protection scheme result in more sensitive protection and detection of the degraded voltage conditions of the electrical distribution system. New relays are easier to calibrate and set and their tolerances are much smaller than those of the relays being replaced. All the undervoltage protective functions are individually testable and the addition of timers to the load shedding circuitry eliminates possible relay race potentially resulting in inadvertent start of EDGs. These performance features provide for an enhanced fulfillment of the original design basis.

This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.

Therefore, prior NRC approval was not required for implementation of this modification.

17

PLANT CHANGE/MODIFICATION 173-293 MV-08 12 & MV-08-13 MODIFICATIONS

~summa This modification replaces the existing Unit 2 main steam system valves MV-08-12 & MV-08-13 (auxiliary feedwater pump 2C steam admission valves), including the valve actuators. These valves were replaced to improve component reliability. The existing valves were a wedge gate design. The new valves are double disc gate valves with enhanced Limitorque actuators. The new replacement valves and actuators meet the design requirements, quality classifications, and code requirements of the original design. The design bases and function of these valves are not changed by the replacement valves.

Safet Evaluation:

The design bases and original function for MV-08-12 & MV-08-13 as described in the FSAR remain unaffected by the replacement valves.

This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.

Therefore, prior NRC approval was not required for implementation of this modification.

18

PLANT CHANGE/MODIFICATION 025-294 HASTE GAS ANALYZERS

~8ummar This modification installs a condensate removal system off the refrigerant cooler at the suction of the Automatic Gas Analyzer sample pump, a conversion of a temporary installation into a permanent installation. This modification provides increased reliability of the waste gas management system's automatic gas analyzer by removing condensate from the sample stream which could otherwise lead to failure of the sample pump.

Safet Evaluation The waste gas analyzer is not required to perform any safety function related to mitigating the consequences of an accident.

However, due to the potential for the release of gaseous radioactive effluent into the confines of the Reactor Auxiliary Building (RAB), this modification is considered Quality Related; The installation of the condensate recovery system off the refrigerant cooler inside the automatic gas analyzer room in the RAB has no impact on any safety related system or equipment. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

PLANT CHANGE/MODIFICATION 068-294 WIDE RANGE STEAM GENERATO LEVEL UPGRADE

~summa r This modification provides additional Wide Range Steam Generator Level (WRSGL) instrumentation (control room indication) for Post Accident Monitoring. This modification provides additional instrumentation to the original plant configuration and is being added to the FSAR Table 7.5-1. This modification fulfills a commitment to the NRC to add redundant wide range steam generator level indication in lieu of meeting all the requirements for RG la97 Category 1, Type D,variable.

Safet Evaluation:

The instrumentation added by this modification provides its function, thus there is no direct or indirect impact on the analysis of any design basis accident, nor is there an increased potential of any radiological hazards. In addition, the potential for an un-analyzed accident 'as 'not been increased by this modification. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

20

PLANT CHANGE/MODIFICATION 104-294 DELETION OF HPSZ LPSZ LO FLOW ALARMS

~Bummar This modification removed obsolete non-safety related ultrasonic flow monitors from HPSI and LFSI piping. These monitors were installed early in the life of the plant to provide additional annunciation in the control room of potential low flov conditions in the HPSI and LPSI pumps during shutdown cooling and post-LOCA.

Safet Evaluation:

The monitors did not perform any safety related functions and were not required to operate the plant. The monitoxs vere not required to mitigate the consequences of an accident, were not requixed to safely shutdown the plant, and did not provide input to any automatic function. The monitors were redundant to safety related indication and plant procedures for identifying and correcting low flow conditions. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

PLANT CHANGE/MODIFICATION 111 294 SAFEGUARDS BYPASS DISPLAY INDICATZON PANEL MODIFICATION

~Summar The Safeguards Bypass Indication System is a part of the plant annunciator system. The bypass panel provides the operator with information about the important safety related systems which are removed from service, tested or being repaired, or disabled. This modification corrected the anomaly that windows A7 and B7 were incorrectly labeled for the existing inputs by deleting the blanking and sparing windows A7 and B7.

inputs'o, Safet Evaluation:

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

22

PLANT CHANGE/MODXFXCATXON 125-294 CONTAINMENT VACUUM HX ALARM SETPOXNT CHANGES

~summa r This modification revised the setpoint for Containment to Annulus Differential Pressure Indicating Switches PDIS-25-11A and PDIS 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 high vacuum alarm before the technical specification limit is encountered. 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, this modification provides an early tripping of containment purge fans.

Safet Evaluation:

The Containment Purge System is designed to reduce the level of radioactive contamination in the containment atmosphere below the limits of 10 CFR 20 so as to permit personnel access to the containment during shutdown and refueling. The containment purge fans are designed to trip on high differential pressure between the containment and the annulus in order to prevent 'ncrease in containment vacuum. An early warning of containment high vacuum condition and the early tripping of the containment purge fans does not affect the opexation of Containment Vacuum Relief System and Containment Purge System. This modification did not constitute an unreviewed safety question or require changes to the plant

~

Technical Specifications. Therefore, prior NRC approval was not required for implementation of, this modification.

23'

PLANT CHANGE/MODIFICATION 148-294 MODIFY 3Y RELAYS IN AB BUS FOR XCW CCW 2C PUMPS

~Summa r Prior to this modification the 2C CCW and 2C ICW pumps automatically started immediately upon receipt of a SIAS if offsite power was available and started using the EDG load sequence timers if offsite power was not available. This modification changed the automatic start circuits of the 2C CCW and 2C ICW pumps such that the pumps automatically start using the EDG load seguence timers regardless of offsite power availability to the 4.16 kV 2AB bus.

This change was made by eliminating the function of relays 3YA1 and 3YB1 for the 2C CCW pump and relays 3YA2 and 3YB2 for the 2C ICW pump. Elimination of the function of these four relays results in a pump loading time delay for an automatic start signal, even under conditions with offsite power available. The manual startin'g of the pumps has not been affected (no time delay).

Safet Eva1uation:

The changes simplify the pump start circuits by effectively bypassing relays 3YA1, 3YA2, 3YB1, and 3BY2, which results in elimination of a premature start failure mode. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

24

PLANT CHANGE/MODIFICATION 008-295 REVISION REPLACEMENT OF THE EXCORE NEUTRON FLUX MONITORING AND 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 detectors RE-001-A2, B2, C2 & D2, along with the detector cable was replaced with Gamma-Metrics wide range detectors and detector cable.

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

3) The four (4) filter assemblies were removed assemblies did not (RX-110A, B.

require external C &

D). The new amplifier filters.

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

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

Safet Evaluation:

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

25

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PLANT CHANGE/MODIFICATION 023-295 NRC GENERIC LETTER 89 10 MOV CONTROL SNITCH SETTINGS

~8llEIIa This engineering package (EP) provided enhancements to motor operated butterfly valves for selected safety related valves to ensure that the valves will function as intended during maximum design basis conditions. These enhancements provide for: 1) field testing to verify that the MOV torque switches are properlythesetmost to allow the MOVs to generate sufficient torque to perform limiting stroke function without over stressing any valve or actuator component, and 2) modify certain MOVs to better facilitate diagnostic testing of the MOVs. This'P provides the specific design information necessary to set both the open and close torque switches for twenty-two (22) St. Lucie unit 2 motor operated butterfly, valves within the scope of NRC Generic Letteractuator 89-10.

Also within the scope of this EP is the replacement of gearing for ten (10) motor operated butterfly valves and the replacement of one (1) thermal overload.

Safet Evaluation:

The valve enhancements accomplished by this engineering package do not adversely affect the ability for these valves to perform their design functions. The MOVs operate in a manner identical to that prior to the modification. This modification did nottoconst'itute the an plant unrpviewed safety question or require changes Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

26

PLANT CHANGE/MODIFICATION 027-295 PRESSURIZER LI UID SPACE INSTRUMENT NORRLES REPLACEMENT

~Summa This Engineering Package (EP) modified the Pressurizer liquid space instrument nozzles E and F on the bottom of the Pressurizer and a temperature nozzle on the side. The replacement nozzles are less susceptible to primary water stress corrosion cracking (PWSCC).

The modification moved the partial penetration weld joint to the Pressurizer outside surface whereas the former joint was on the inside.

Safet Evaluation:

The replacement Pressurizer instrument nozzles are equivalent to the former nozzles and fabricated from acceptable material.

Welding of the weld buildup pads and partial penetration J welds were controlled by weld procedures which were reviewed and approved by FPL. The new nozzle design does not change, degrade, or prevent actions described in, or assumed to occur in the mitigation of any accident described in the SAR. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

27

PLANT CHANGE/MODIFICATION 029-295 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 System fluctuations, control signals are generated to start/stop charging pumps and to modulate the Letdown Flow Control Valves (LCV-2110P &

Q), accordingly. A drop in pressurizer level will require additional charging and reduced letdown flow. The following modifications were made to the Pressurizer Level Control System to improve the balance and smooth operation of the system:

a

~ Changing LIC-1110X & X's gain from 6.02 (16.6% proportional) to 5.14 (19.444 proportional).

~ Providing a' 20 second operating range for the lag component within the control loop to accommodate changes in the va'lve characteristics 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 letdown system as described in the FSAR. The mechanical functions for the letdown system remain unchanged by this design. The control changes enhance the operation of the system. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

28

0 PLANT CHANGE/MODIFICATION 035-295 CONDENSER TUBE CLEANING & DEBRIS FILTER SYSTEM PHASE I

~summa This modification installed a Debris Filter System (DFS) which mechanically removes entrained solid particles and a Continuous Condenser Tube Cleaning System (CTCS) that mechanically cleans the condenser tubes. The DFS operates continuously with a periodic automatic backwash to the circulating water discharge. The CTCS operates continuously to inject sponge-type balls into the condenser where they pass through (and clean) the tubes, are collected on the discharge, and returned for re-use. The net effect for both the DFS and the CTCS is to reduce condenser downtime for cleaning and prevent, a reduction in generating capability due to condenser fouling limitations. This PCM installs components (auxiliary piping/valves, instrumentation, and electrical components) of the DFS and CTCS. Taprogge America Corporation (TAC) is providing the modification and installation of the DFS and CTCS spool pieces in the circulating water inlet and outlet lines under PCM 036-295. Two PCMs were developed in order to differentiate the implementation scopes of the contract with TAC.

Safet Evaluation:

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

29

t I PLANT CHANGE/MODIFICATION 036-295 CONDENSER TUBE CLEANING & DEBRIS FILTER SYSTEM PHASE XX

~summa This modification, installed a Debris Filter System (DFS) which II mechanically removes entrained solid particles and a Continuous Condenser Tube Cleaning System (CTCS) that mechanically cleans the condenser tubes. The DFS operates continuously with a periodic automatic backwash to the circulating water discharge. The CTCS operates continuously to inject sponge-type balls into the condenser where they pass through (and clean) the tubes, are collected on the discharge, and returned for re-use. The net effect for both .the DFS and the CTCS is to reduce condenser downtime for cleaning and prevent a reduction in generating capability due to condenser fouling limitations. This PCM installed the DFS and CTCS spool pieces in the circulation water inlet and outlet lines by Taprogge America Corporation (TAC). PCM 035-295 installed components (auxiliary piping/valves, instrumentation, and electrical components) of the DFS and CTCS.

Two PCMs were developed in order to differentiate the implementation scopes of the contract with TAC.

Safet Evaluation:

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

30

PLANT CHANGE/MODIFICATION 063-295 HP TURBINE BLADE RING CHANGEOUT

~mumm ar This modification replace HP turbine blade ring, blade ring carrier bolting, blade ring pins, outer gland case and associated hardware.

These changes were made per a vendor recommendation to preclude premature failure of the components due to erosion or erosion-corrosion. The modified components are all non-rotating parts that direct steam flow over the rotor blade, provide internal support or seal the shaft area. The blade rings 'and their bolting aid in containing postulated turbine missiles.

Safet Evaluation:

The replacement components resulted in no degradation, either directly or indirectly, to any safety functions required for analyzed accidents, and do not increase any radiological hazards.

This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.

Therefore, prior NRC approval was not required for implementation of this modification.

31

I PLANT CHANGE/MODIFICATION 096-295 REPLACEMENT OF 2A & 2B UNDERGROUND DIESEL FUEL OIL TRANSFER LINES

~summa This modification rerouted and replaced the underground fuel oil transfer lines between the 2A & 2B diesel oil transfer pump and the expansion joint upstream of the 2A1, 2A2, 2B1, & 2B2 diesel oil day tanks. This modification .was essentially a like-for-like replacement of the existing piping, with a few enhancements added.

The new underground piping utilizes a cathodically protected guard pipe which provides"'improved corrosion protection and isolates any potential future fuel oil leakage from the environment. The new piping is provided with design, basis tornado missile protection in accordance with FSAR requirements. The old diesel fuel oil transfer lines have been abandoned in place.

Safet Evaluation:

This modification was essentially a like-for-like replacement. The EDG fuel oil system and the EDGs have not been effected. The design differences between the original design and the design of the new piping have been addressed and determined to be acceptable.

This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.

Therefore, prior NRC approval was not required for implementation of this modification.

32

PLANT CHANGE/MODZFZCATEON 132-295 V3439 & V3507 RELIEF VALVE REPLACEMENT

~Summar This modification replaced relief valves V3439 and V3507, increased the lift pressure from 500 to 535 psig, reduced the blowdown percentage from 104 to 6% 84, and increased the discharge piping size from 1/2" to 2" diameter. This modification was performed to reduce the potential for V3439 or V3507 to lift during normal shutdown cooling system (SDC) operation and reduce the time for the valve to reseat.

Safet Evaluation:

Valves V3439 and V3507 provide, protection against pressure developed due to fluid thermal expansion in an isolable section of the low pressure safety injection system (LPSI) headers 2A and 2B, respectively. The function of V3439 and V3507 was not changed by the installation of the new replacement valves, the increase in the-set pressure, the reduction in blowdown setting and the modification of the discharge piping. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not 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 the Letdown intermediate leg pressure controller PIC-2201 from 460 to 430 psig and associated alarm PA-2201 setpoints revised to 510 and 395 psig, and reduced the design blowdown value for the existing relief valve V2345 from 254 to 15 +0/-24, but maintained V2345's pressure setpoint at 600 psig. The modification was made to provide positive margin between the system normal operating pressure and the pressure at which V2345 will reseat following actuation.

Safet Evaluation:

Valve V2345 is installed downstream of the LCV-2110P&Q to provide overpressure protection for the .intermediate pressure letdown piping ,and letdown heat exchanger should PCV-2201P&Q close unexpectedly. The reduction of the Letdown Backpressure Control set pressure to 430 psig did not adversely affect the operation of the Letdown system. The bases of the backpressure control set pressure is to ensure that the fluid downstream of the letdown control valves does not flash to steam following a reduction in pressure 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 V2345 has not been changed. Adjustments of. the alarm setpoints ensures that adequate margin was provided to eliminate spurious alarm indications, while also ensuring that abnormal operating conditions are identified to plant operators through annunciation. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation, of this modification.

34

l PLANT CHANGE/MODIFICATION 135-295 THERMAL RELIEF VALVE BLOWDOWN MODIFICATION

~summar This modification reset thermal relief valve V3412's blowdown from 254 to 10-154 to ensure that the valve's minimum reseat pressure exceeds the maximum operating pressure'n HPSI header 2B. The lift pressure is unchanged at 1585 psig.

Safet Evaluation:

Reducing the blowdown from 25% to 10-154 ensures that valve V3412 will reseat even if there is a large unanticipated pressure transient following the valve lift which could prevent the valve from reseating. The function and the setpoint of the valve has not been changed. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

35

f PLANT CHANGE/MODIFICATION 136-295 SAFETY RELIEF VALVE V3417 SETPOINT AND BLOWDOWN MODIFICATION

~Summa' This engineering package (EP) modified the St. Lucie Unit 2 relief valve V3417. Valve V3417 provides protection against pressure developed due to charging pump discharge into an isolable section of the 2A high pressure header. The pressure setpoint of V3417 was increased from 2400 to 2485 psig, the blowdown was reduced from 254 to 10-12%'. This modification was performed to reduce the potential for V3417 to lift during system testing and off-normal operation and to reduce the potential of the valve failing to reseat at system operating pressures.'afet Evaluation:

The function of V3417 was not changed by this increase in the setpoint and the reduction in blowdown settings. This modification did not constitute an.unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

36

0 PLANT CHANGE/MODXFECATZON 152-295 RCS HOT LEG XNSTRVMENT NOZZLE REPLACEMENT

~Summa This engineering package(EP) modified the St. Lucie Unit 2 Reactor Coolant System (RCS) hot leg instrument and sample nozzles. The nozzles are designated as flow measurement nozzle J (quantity of eight) and sampling nozzle K (quantity of one) on drawing 2998-3793. These nozzles were made of a specific heat of material that is known to be susceptible to primary water stress corrosion cracking (PWSCC) and the flow measurement nozzle J for PDT-1121B was found to be leaking on 10/10/95. The purpose of these replacements was to remove and replace the sensing line nozzle for PDT-1121B and to minimize the possibility of a future nozzle failure. The modification moved the partial penetration weld joint to the hot leg outside surface where previously the joint was on the inside surface. The material of the nozzles was changed from Znconel 600 to Inconel 690.

Safet Evaluation:

This engineering package did not involve an unreviewed safety question, not did it recpxire a revision to the plant Technical Specifications. This modification did not effect plant safety and operation. Therefore, prior NRC approval was not required for implementation of this modification.

37

PLANT CHANGE/MODIFICATION 156-295 DELETION OF EDG AUTO START ON CXAS 6 CSAS

~summa This modification deleted the Containment Isolation Actuation Signal (CIAS) and Containment Spray Actuation Signal (CSAS)

Emergency Diesel Generator (EDG) start signals. The EDG start for Safety Injection Actuation Signal (SIAS) and Loss of Offsite power (LOOP) start signals remain. This modification was necessary to prevent operation of the EDGs with protective trips blocked while paralleled with offsite power. CIAS and CSAS are redundant trips to SIAS and the SIAS opens the EDG output breaker preventing operation of the EDG in parallel with offsite power and with disabled protective trips.

Safet Evaluation:

This modification did not affect the operation of the Containment Isolation System or the Containment Spray System. All functions of the CIAS and CSAS remain'ith the exception of the EDG start signals. The SIAS signal starts the EDG providing emergency electrical power. The purpose of this modification was to eliminate any possibility of damage to the EDGs due to spurious actuation of CIAS or CSAS relays, which may result from operating in parallel to offsite power with the'protective relay functions blocked. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.

Therefore, prior NRC approval was not required for implementation of this modification.

38

PLANT CHANGE/MODIFICATION 161-295 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 vapor seal drain system was reconfigured from a closed system that drained to the reactor drain tank to an open system that drains to the containment sump via open floor drains. This modification was accomplished to prevent accumulation of boric acid crystals on the tops of the RCP vapor seal assembly due to lack of proper drainage from the vapor seal assembly.

Safet Evaluation:

Operation of plant systems, including the RCPs and the liquid waste management system, was unaffected by this change; however, operators have been informed that RCP vapor seal leakage is now directed to a floor drain where atmosphere it may slightly affect containment radiation levels and reactor cavity sump level indication. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not require for implementation of this modification.

39

PLANT CHANGE/MODIFICATION 167-295 DRILLING OP VALVE DISK V3481

~Summa'his modification consisted of drilling a hole on the upstxeam or Reactor Coolant System (RCS) side of the valve disk of Shutdown Cooling (SDC) isolation valve V3481 in order to vent the bonnet of high pressure fluid, and thereby prevent potential pxessure locking. This modification was performed to satisfy the requixements of NRC Generic Letter 95-07 (endorsed under NUREG-1275 Vol. 9), and has been performed within the nuclear industry to address the pressure locking issue.

Safet Evaluation:

The functional safety related requirement of V3481 to open or close, or to maintain the integrity of the LPSI/SDC were not changed by drilling a hole on the upstream or'CS side of the disk.

This modification enhances the reliability of the valve and the SDC system since it will vent the bonnet of high pressure fluid and thereby eliminate the potential for a pressure locking condition.

This modification will not affect the structural integrity or the seismic qualification of the valve. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not 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 from normally locked closed to normally locked open to eliminate the potential for pressure locking the valve in its isolated position.

The "valve open<< annunciator was rewired to annunciate a "valve closed" position and the RTGB hand switch was reconfigured to permit the valve to be administratively locked open.

,Safet Evaluation:

The shutdown cooling suction cross-tie line was added to the Unit 2 system design to satisfy single failure criteria. A cross-tie isolation valve was included in the design to preserve the remove one train from service for maintenance or repairs.

ability'o Changing the normal valve position from normally shut to normally open does not affect the valves functional ability to isolate the cross-tie. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not. required for implementation of this modification.

41

PLANT CHANGE/MODIFICATION 194-295 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 Cooling Intake Cooling Water (ICW) pump and the 2A, 2B, and 2C Component Water (CCW) pump control circuits. This modification was implemented to delete the operational requirement of manually resetting the RTGB control switch to "stop" and then back to "auto" which may have been required to preserve the automatic SIAS start feature of the subject pumps. The resetting of the RTGB control switches was only required after a local push-button stop of a running pump which was started via the RTGB control switch.

Disconnection of the local push-button stations from the associated control circuit prevents the stopping of a running pump via the local push-button station and thus preserve the automatic SIAS start feature of the pumps without requiring a manual reset of the RTGB control switches.

Safet Evaluation:

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

42

PLANT CHANGE/MODIFICATION 216-295 EMERGENCY DIESEL GENERATOR LO'R COOLING PRE88URE ALARM SETPOINT

~summa@

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

Safet Evaluation:

This engineering package did not have an adverse effect on plant safety, security or operation, did not constitute an unreviewed safety question and did not require a change to plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

43

0 PLANT CHANGE/MODIFICATION 245-295 EVE 21 A&B CEDM COOLING FAN SETPOINT CHANGE

~811DBI K This modification changed the time delay setpoint in the starting logic for the'"Control Element Drive Mechanism (CEDM) cooling fans (HVE-21 A&B) from 10 seconds to 30 seconds. When a fan switch is taken to the start position, one fan starts immediately and then, after the start delay, the standby fan starts if a low flow condition exists. Before this modification the standby fan would frequently start. This modification prevents the inadvertent start of 'the standby fan when starting a fan from stop but does not affect the time the standby fan will automatically start after failure of a running primary fan.

Safet Evaluation:

This modification is to increase the setpoint of time delay relays 62/507 and 62/508 associated with the CEDM Cooling System (HVE-21 A&B). The CEDMs are referenced in the FSAR as Not Nuclear Safety.

In addition, the change has no effect on CEAs. This modification did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of this modification.

SECTION 2 SAFETY EVALUATIONS 45

SAFETY EVALUATION ZPN-PBL-SENS-94-018 REVISZON 0 OCHLORITE SYSTEM MODIFICATIONS

~summa This safety evaluation demonstrated that Hypochlorite (CL) system modifications performed in accordance with the guidance of this evaluation did not, adversely affect plant safety, security or operation. The CL system is a non-safety system common to St.

Lucie Units 1 & 2 that produces a sodium hypochlorite solution via electrolytic decomposition of filtered seawater. The hypochlorite solution is periodically injected into the suction side of the intake cooling water (ICW) and circulating water (CW) pumps for the control of biological fouling.

Safet Evaluation:

This safety evaluation addressed the technical and licensing requirements for the Hypochlorite (CL) system and concluded that the proposed plant modifications were bounded by the Technical Specifications and did not change the analysis of accidents addressed in the FSAR or the results and conclusions of any previous safety evaluations. The actions or changes identified and evaluated in this safety evaluation did not have any adverse effect on plant safety or plant operations. The actions or plant changes (procedures and/or hardware), identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.

46

I SAFETY EVALUATION JPN PSL-SEEP-94-035 REVISION 0 UPDATE FSAR LOOP ACCURACIES VARIOUS ELECTRICAL DISTRIBUTION SYSTEM METERS WITH CURRENT INFORMATION

~8llEEIK This safety evaluation demonstrated the acceptability of updated LOOP accuracy data for various meters which monitor electrical parameters (voltage, current, frequency, etc.) and are included in Table 7.5-1 (" Safety Related Display Instrumentation" ) of the St.

Lucie Unit 2 FSAR. PC/M 428-291M performed a like for like replacement of GE meters with meters manufactured by Yokogawa.

This safety evaluation considered the accuracy values for these new meters.

Safet Evaluation:

The accuracies for the electrical parameter meters are not used as an assumption in any analyzed accident of the accident analyses.

This update for the metering instrumentation neither involves an unreviewed safety question not requires changes to Technical Specifications. Therefore, this change and revision to the FSAR

~

were performed without prior NRC approval.

47

SAFETY EVALUATZON iTPN-PSL-SENP-94-037 REVISION SZT DISCHARGE LOOP CHECK VALVE STROKE TEST

~summar The purpose of this safety evaluation was to demonstrate the acceptability of performing a partial flow, full stroke test of the Safety Injection Tank (SIT) discharge/loop check valves.

Successful performance of the test was used to satisfy NRC requirements for SIT check valves as delineated in Generic Letter 89-04. There are four SITs with two check valves associated with each tank. The check valves tested were V-3215, V-3225, V-3245, V-3217, V-3227, V-3237, and V-3247. Partial flow, full stroke testing supersedes disassembly and inspection as a method to demonstrate acceptability. Plant restrictions applicable to the test were identified and listed.

Safet Evaluation:

This safety evaluation addressed the effect of the test on safety related components, including the fuel, steam generator nozzle dams, refueling water cavity clarity, and induced cyclical thermal stresses in the SIT. The actions or plant changes in procedures identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.

48

SAFETY EVALUATION ZPN-PSL-SENP-94-039 REVISION 1 PUMPER LIFTED TEAD FOR PDIS-2216

~Summa r This safety evaluation demonstrated the acceptability"of applying a .jumper to the pressure differential switch (PDIS-2216) in the chemical & volume control system (CVCS) letdown line. This switch measures the differential pressure across the regenerative heat exchanger as a means of detecting high fluid flow that would occur from a downstream line break. The jumper of this switch defeats closure on high differential pressure. Without the differential pressure switch, letdown isolation still occurs from a temperature element (TE-2221) located immediately downstream of the regeperative heat exchanger. TE-2221 senses high temperature (470 F) downstream of the regenerative heat exchanger from excessive letdown flow resulting from a line break.

Safet Evaluation:

a This safety evaluation addressed the technical and licensing requirements for the jumpering of pressure switch PDIS-2216 and

~

concluded that the proposed plant configuration and mode of operation was bounded by the Technical Specifications and did not change the analysis of accidents addressed in the FSAR or the results and conclusions of any previous safety evaluations. The actions or changes identified and evaluated in this safety evaluation did not have any adverse effect on plant safety or plant operations. The actions or plant changes in procedures, identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plant Technical Specifications.

Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.

49

SAPETY EVALUATION iTPN-PSL- JNO-95-001 REVISION 0 NUCLEAR PLANT CHEMISTRY PARAMETERS MANUAL REV. 18

~8UEN I This safety evaluation determined that Revision 18 to the Nuclear Plant Chemistry Parameters Manual had no adverse affect on plant safety.

Safet Evaluation:

This safety evaluation determined that the changes did not adversely affect the safety related equipment, plant operations, or safety functions. The actions or plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to'he plant Technical Specifications.

Therefore, prior NRC approval was not required for implementation of the actions or conditions identified within this evaluation.

50

SAFETY EVALUATION i7PN-PSL-SEMS-95-001 REVISION 0 LETDOWN PRESSURE CONTROLLER PIC-2201 SET PRESSURE REDUCTION

~Summa r This safety evaluation demonstrated the acceptability of operating with a reduced set pressure as low as 360 psig for St. Lucie Unit 2 Letdown Backpressure Controller PIC'2201 under controlled conditions. The normal operating set pressure is 450 psig with controller in the automatic mode. The reduced operating set pressure was necessary'o allow diagnostic testing of the Letdown Control Valves LCV-2110 P&Q control modules and valve response with dynamic flow conditions.

The conditions for operating at the reduced pressure were to support diagnostic testing of LCV-2110 P&Q and/or the starting or stopping of a second charging pump. An additional restriction was that 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 initiated.

if an RCS transient was Safet Evaluation:

This safety evaluation considered the effect on the operation of the Chemical Volume and Control System with a reduced set pressure on PIC-2201 under controlled conditions. It was determined that there were no adverse effects on plant safety under the conditions described in the safety evaluation. Therefore, the conditions or changes in procedures and design documents addressed under this evaluation did not involve an unreviewed safety question or require a change in plant Technical Specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified in this safety evaluation.

51

SAPETY EVALUATION i7PN-PSL-SEMS-95-002 REVISION 0 OPERATION OP THE UNIT 2 REPUELZNG WATER TANK DURZNG ZNBPECTZONB

~8UBQSEX'his safety evaluation documented the acceptability of performing an inspection of the inside of the Refueling Water Tank (RWT) during normal plant operations. The inspection by divers of the inside of the RWT was necessitated by failure of a retention element in the Unit 2 Spent Fuel Pool Ion Exchanger which resulted in a release of approximately 25 ft of resin which may have traveled to the RWT and settled to the bottom.

Safet Evaluation:

The evaluation determined that due to the RWT arrangement, piping sizes, and function an inspection by divers and/or submersible mechanical devices did not have an adverse effect on plant safety.

The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

52

0 0

SAFETY EVALUATION iTPN-PSL-SEMP-95-004 REVISION 1 REDUCED PRESSURIZER HEATER CAPACITY

~Summa This safety evaluation documented the acceptability operating the unit with a reduced pressurizer heater capacity of 1200 Kw. This permitted up to 6 pressurizer heaters, a total of 300 Kw, to be removed from service. The safety evaluation specified, the issuance of a PC/M to cover the physical modifications and that changes be made on the affected documents (control wiring diagram, TEDB, and power distribution data sheets). Table 1 of the safety evaluation i'ncludes the revisions to account for the heaters that have been retired 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). Because of the design margin of the pressurizer heaters, safe plant shutdown and the results of postulated events in the FSAR safety analyses are not adversely affected by operation with reduced heater capacity. he plant conditions identified in this safety evaluation did not constitute an unreviewed, safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified. within this evaluation.

53

SAFETY EVALUATION iTPN-PSL-SENS-95-005

,REVISION 0 EVALUATION FOR AN ALTERNATE REACTOR COOLANT GAS VENT SYSTEM ALIGNMENT

~summa This safety evaluation documented an alternate alignment for the reactor coolant gas vent system (RCGVS). The changes to the system that were evaluated are: 1) installation of a blank at the flanged connection downstream of V1465 for the purpose of arresting the leak through V1465, 2) providing 125V DC safety bus SA control power to V1464, and 3) depowering the normal SB bus control wiring to V1464. A blank fabricated from ASTM A-240 Type 316 SS (M&S 1032753110) 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 the system to meet the Technical Specification 3/4.4.10 LCO requirements. Operability of the RCGVS is not affected since two vent paths are available from both the pressurizer and the reactor vessel head. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

54

SAFETY EVALUATION O'PN-PSL-SENS-95-008 REVISION 0 DEENERGIZATION OF RAB VENTILATION DAMPERS D-5B & D-7B TO SUPPORT ACTUATOR REMOVAL FROM DAMPER D-7A

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

Safet Evaluation:

The safety functions of the RAB HVAC systems were not affected by this change since deenergization resulted in the proper post-accident system configuration. Although normal operation of the HVAC system was affected (loss of normal ventilation flow through D-5B & C-7B) the impact was insignificant. There were no plant operating restrictions required as a result of this evaluation.

The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the

~

plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

55

SAFETY EVALUATION J'PN-PSL-SENS-95-010 REVZSION 0 EVALUATZON FOR OPERATZON WITH A DIESEL OIL STORAGE TANK BUILDING MZSSILE DOOR REMOVED

~summa This safety evaluation documented acceptability of plant operation with a single Diesel Oil Storage Tank (DOST) building missile door removed. The two St. Lucie Unit 2 DOSTs are housed within a concrete enclosure which provides missile protection for the DOSTs.

Each section of the building has its own access door on the west side 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 door maintenance.

Safet Evaluation:

This evaluation concluded that operation of the plant with a single DOST building missile door removed did not impact plant safety and did not constitute an unreviewed safety question nor require a change to the technical specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

56

SAFETY EVALUATION JPN-PSL-SENS-95-013 REVISION 0 EVALUATION FOR OPERATION WITH DIESEL OIL TRANSFER PUMP 2B DISCHARGE ISOLATION VALVE V17216 CLOSED

~Summar This safety evaluation documented acceptability of plant operation with Diesel Oil Transfer Pump (DOTP) 2B discharge isolation valvewere V17216 in the closed position. Compensatory measures established to open the valve upon operation of the 2B Emergency Diesel Generator (EDG). V17216 is normally a LOCKED OPEN valve; however, due to a suspected leak in the underground piping downstream 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 valve 1I216 in the closed position did not impact plant safety and did not constitute an unreviewed safety question nor require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

57

SAFETY EVALUATION iTPN-PSL-SENS-95-021 REVISION 1 FUEL HANDLING E UIPMENT FSAR CHANGES

~BMUBRX'his safety evaluation justified and provided for Unit 2 FSAR changes identified as a result of an FSAR review conducted by the FPL Quality Assurance Department and documented in Audit No.

QSL-OPS-94-24.

Saf et Evaluation:

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

58

0 0

SAFETY EVALUATION a7PN-PSL-SENP-95-021 REVISION 0 DELETION OF SR-A1A EMBANKMENT SURVEY AND ANNUAL AERIAL PHOTOGRAPH OF BEACH AREA COMMITMENTS AT ST LUCZE SITE

~8ummar This safety evaluation provided justification for deletion of regulatory commitments to visually inspect the SR-A1A highway embankment following passage of any major storm and to perform an annual aerial photograph of the beach area. Additionally,, this safety evaluation provided the revision to the FSAR that documented these commitment changes.

Safet Evaluation:

These commitments were not required to achieve compliance with any rule, regulation or NRC order, were not made to minimize recurrence of an adverse condition, and were not relied upon in the original erosion analysis to assure protection to the plant from design basis flooding. The FSAR changes identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION JPN-PSL-SEFJ-95-022 REVISION 0 TEMPORARY USE OF DUMMY FUEL ASSEMBLY SKELETON REPLACING A CELL BLOCKING DEVICE FOR THE SPENT FUEL STORAGE RACK

~summar This safety evaluation documented acceptability of continued Spent Fuel Pool {SFP) operation with a dummy fuel assembly skeleton used as a physical barrier in place of a cell blocking device. A cell blocking device of the Region 2 storage cell, T15, was inadvertently knocked out of position during a recent SFP resin cleanup vacuuming. The blocking device prevented placing a fuel assembly in storage cell T15. A dummy fuel assembly skeleton was placed in cell T15 to prevent inadvertent use of the cell for fuel storage.

Safet Evaluation:

This safety evaluation discussed th'e SFP criticality physical mode and the effect of storage of a fuel assembly skeleton on rack reactivity. The evaluation concluded that the use of the dummy fuel assembly skeleton serving as a cell blocking device is technically justifiable. The plant conditions identified in, this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications.

Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION JPN PSL-SEMS-95-024 REVISION 1 ECCS HEADER 2A2 PRESSURIRATION DUE TO BACK LEAKAGE THROUGH V3217

~mmmm ar This safety evaluation evaluated the safety significance of the back leakage through RCS loop 2A2 ECCS cold leg injection header check valve V3217 and to allow the upstream piping segment to remain pressurized to RCS pressure until the next refueling outage in approximately 3 weeks. The configuration analyzed by this evaluation did not alter the RCS pressure boundary. The segment of piping pressurized to RCS pressure simply captured check valve leakage from V3217 which was determined to be within the Technical Specification limit of 1 gpm. The evaluation also included the use of a temporary pressure gage for monitoring purposes.

Safet Evaluation:

The safety evaluation considered the effects on RCS pressure boundary, piping design, ECCS operation, RCS leakage monitoring, containment integrity, and design integration. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION JPN-PSL-SEZP-95-031 REVISION 0 TEMPORARY USE OF AN ACOUSTIC FLON METER TO CORRECT THE DDPS FEED%PATER FLOP/ COEFFICIENT

~summa r This safety evaluation documented acceptability of using the Digital Data Processing System {DDPS) flow coefficients used in the calorimetric calculation to be corrected by the Leading Edge Flowmeter {LEFM) measurements. The feedwater flow measurement has been determined to be in error due to fouling within the venturi used to measure differential pressure. The LEFM was installed externally to the feedwater line where it is not subject to errors caused by pipe fouling. The more accurate flow readings provide a better calorimetric calculation.

Safet Evaluation:

The procedure covered by this safety evaluation did not change the intent of the feedwater flow measurement for either operator review or use in the calorimetric equation. The transmitters providing differential pressure across the feedwater venturi continue to be the primary instrument. providing indication. This evaluation does not 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 Power operation. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION ZPN-PSL-SENS-95-031 REVISION 0 TEMPORARY REMOVAL OF A CCW BUILDING MISSILE BARRIER

~smamar This safety evaluation documented acceptability of plant operation in MODE 5 or MODE 6 with one of the missile shield doors in the Component Cooling Water (CCW) building removed to perform maintenance activities. The function of the missile shield doors is to protect the CCW components during a hurricane/tornado from design basis missiles.

Safet Evaluation:

This safety evaluation demonstrated that temporarily removing a CCW building missile shield door while in MODE 5 or 6 did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not. required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION JPN-PSL-SENS-95-032 REVISION 0 REFUELING OPERATIONS NITH A STUCK REACTOR VESSEL STUD

~mamma This safety evaluation documented acceptability of leaving stuck reactor vessel stud f44 in place during refueling and until be replaced during the next scheduled refueling outage.

it can Reactor vessel stud f44 stuck during refueling. This safety evaluation allowed 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 vessel studs is with respect to corrosion concerns. A technical evaluation of the stuck stud concluded that the expected corrosion effects were minimal over a three week period. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions tha't were identified within this evaluation.

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~ ~

, SAFETY EVALUATION FRG f95-34 OPERATION OF NEW FUEL HANDLING CRANE FOR MAINTENANCE

~summa r This safety evaluation documented acceptability of operating the New Fuel Handling Crane for maintenance activities when no new fuel is stored in the Fuel Handling Building. The FSAR states that this crane is used exclusively for new fuel handling and is locked in position when fuel handling is not in progress. Several of the lights in the Fuel Handling Building had burned out bulbs which could only be accessible through the use of the installed new fuel handling crane.

Safet Evaluation:

The safety evaluation concluded that this change in operation of the New Fuel Handling Crane while there is no fuel in the new fuel storage area of the Fuel Handling Building does not adversely affect plant safety, security or operation, does not constitute an unreviewed safety question, nor does Technical Specifications.

it require changes to the Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION JPN-PSL-SEES-95-034 REVISION 0 EVALUATION FOR THE PROVISIONS TO TRIP EMERGENCY DIESEL GENERATOR OUTPUT BREAKER ON CIAS IN PLANT MODES 5 AND 6

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

Safet Evaluation:

This change did not affect plant operating practices and enhanced EDG protection in MODES 5 and 6. This evaluation concluded that safety question this modification did not constitute an unreviewed Specifications.

or require changes to the plant Technical Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION FRG-95-36 BYPASSING AUTOMATIC ESFAS ACTUATION DURING OPERATING MODES 5 AND 6

~Summar This safety evaluation documented acceptability of operating the plant in cold shutdown (MODE 5) and refueling (MODE 6) with SIAS, MSIS, CSAS, RAS and the high containment pressure initiated CIAS disabled as long as the administrative controls .placed on. the additional sets of ESFAS bypass keys are adequate to ensure that no more than one channel can be bypassed in MODES 1 through 4.

Additional sets of bypass keys were used to bypass the desired portions of the Engineered Safety Features Actuation System. The additional keys are maintained under strict administrative controls, providing assurance that they were removed prior to entering MODE 4 and were not available to the operators during the time the unit is not in MODES 5 OR 6.

Safet Evaluation:

This change in the operation of ESFAS does not adversely affect plant safety, security or operation, does not constitute an unreviewed safety question, nor does Technical Specifications.

it require changes to the Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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j ~

SAFETY EVALUATION JPN-PSL-SENS-95-037 REVISION 1 EVALUATION FOR CROSS-CONNECTING 480V LOAD CENTERS DURING MODES .5 AND 6

~summa r This safety evaluation documented acceptability of operating the A and B trains of the safety related 480V electrical system cross connected 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 connected with one of the safety 4.16kV busses (2A3 or 2B3) out-of-service for required maintenance while in MODES 5 or 6 neither involves an unreviewed safety question nor requires a change to plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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

REVISION 0 TEMPORARY REMOVAL OF DIESEL OIL LINE TORNADO MISSILE BARRIER

~Summa@

This safety evaluation documented acceptability of plant operation in MODE 6 with one section of the Diesel Oil transfer line tornado missile barrier removed to facilitate inspection and/or repair of the 2A/2B diesel oil transfer line. The discharge lines from the fuel oil transfer pumps exit. the Diesel Oil Storage Tank (DOST) above grade then run underground to the Diesel Generator building where they rise above grade and enter the Diesel Generator Building. Both the above grade and underground portions of the lines are protected from design basis tornado missiles. This evaluation concluded that the missile barriers are designed to support maintenance, inspection, and repair on the fuel oil lines and that temporary removal of that section during MODE 6 is acceptable.

Safet Evaluation:

This safety evaluation demonstrated that temporarily removing the diesel oil transfer line missile lid while in MODE 6 neither involves an unreviewed safety question nor requires a change to plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION JPN-PSL-SENS-95-043 REVISION 0 EVALUATION FOR THE TEMPORARY INSTALLATION OF COUBTICAL MONITORING E UIPMENT

~mumm ar This safety evaluation documented acceptability of mounting acoustic sensors to the pressurizer code safety valves (SRVs),

adjacent piping and additional locations as approved by Engineering in order to monitor for potential valve seat leakage. The sensors and/or their associated mounting studs may remain installed as necessary to support data acquisition through the current operating cycle, at which time they will be removed.

Safet Evaluation:

This evaluation concluded that the ,operation of the plant as described does not represent and unreviewed safety question, does not require a change to plant Technical Specifications and does not adversely affect plant operation or safety. Therefore, prior NRC approval was not required for the conditions that. were identified within this evaluation.

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SAFETY EVALUATION ZPN-PSL-SEEP-95-045 REVISION 0 POST ACCIDENT MONITORING INSTRUMENTATION COVERED BY TECHNICAL SPECIFICATIONS

~summar This safety evaluation documented acceptability of updated information for Table 7.5-1 (" Safety Related Display Instrumentation" ) of the St. Lucie Unit 2 FSAR, as to which instruments are used for post accident monitoring. The new FSAR information is consistent with that provided to the NRC. No plant configuration changes were made or analyzed by this safety evaluation.

Safet Evaluation:

This evaluation determined that the updated information, specifying the post accident monitoring instrumentation, neither involved an unreviewed safety question nor required a change to the Technical Specifications. Therefore, this update to the FSAR may be performed without prior NRC approval.

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SAFETY EVALUATION JPN-PSL-SENP-95-046 REVISION 0 EVALUATION FOR THE REMOVAL OF THE UNIT 2 PRESSURIZER MISSILE SHIELD ROOF

~sum'mar 'h This safety evaluation documented acceptability of permanently removing the St. Lucie Unit 2 pressurizer missile shield roof. The purpose of the pressurizer missile shield roof is to prevent credible pressurizer missiles from penetrating and 2" puncturing the containment vessel. The containment vessel is a thick steel right hand, circular cylinder with a 1" thick steel hemispherical dome. The evaluation concluded that there does not exist any credible missile which require the pressurizer missile shield roof to protect the containment vessel. The plant benefits include identified with removal of the pressurizer missile shield roof the reduction in 'risk from dropping a heavy load, reduction in thermal aging effects on components located in the pressurizer cubicle and it facilitates, inspection activities while the unit is at normal operating pressure and temperature.

Safet Evaluation:

The safety evaluation concluded that the removal of the pressurizer missile shield roof does not adversely affect safe operation of the plant and does not constitute an unreviewed safety question and

'o does not require a change the Technical Specifications.

Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION FRG 95-48 JUMPER LIFTED LEAD 2-95-003

~Summar This safety evaluation documented acceptability of jumpering and lifting lead 2-95-008 in order to eliminate a ground on the pressurizer heater back-up B-1 power panel 226.

Safet Evaluation:

The safety functions of the pressurizer heaters were not affected by this change. There were no plant operating restrictions recgxired as a result of this evaluation. The plant conditions identified in this safety evaluation did not constitute an unreviewed safety questi*on or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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0 SAFETY EVALUATION JPN PSL-SENP-9S-094 REVISION 0 DELETION OF THE CONDENSATE DEGASIFIER

~Summa r This safety evaluation documented acceptability of removing the condensate degasifier from the Unit 2 turbine building. The degasifier was ineffective in removing gases from demineralized water and was abandoned in place since startup of Unit 2.

Safet Evaluation:

This safety evaluation concluded that the removal of the condensate degasifier did not represent an unreviewed safety question, did not require a change to plant Technical Specifications, and did not adversely affect plant operation or safety. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION ZPN-PBL-SENP-95-096 REVISION 0 CLASSIFICATION OF EMERGENCY DIESEL GENERATOR AUXILIARIES AND FUEL OIL PIPING ASME DESIGN UALZTY GROUP

~summa This safety evaluation documented acceptability of reclassifying the emergency diesel generator (EDG) auxiliaries and fuel oil system as owner optional upgrade Quality Group C. These components are the EDG air start system, the EDG fuel oil system, the EDG lube oil system, the EDG cooling water designed system, and the intake and and constructed to a exhaust system. These systems were higher ASME Quality Classification to increase reliability, however, there were no provisions made to allow for individual component ZST testing which was optional since they were owner upgrades. These sub-components of the EDG sets are tested when the EDGs 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 notoil systems did not represent an unreviewed safety question and did require a change to plant Technical Specifications and did not adversely affect plant operation or safety. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION JPN FRG 95-145 USING COMPUTER SOFTWARE PROGRAM AND ASSOCIATED SENSORS FOR TESTING SETTING AND CALIBRATING PRIMARY RELIEF VALVES AND OTHER RELIEF VALVES

~Summar This safety evaluation documented acceptability of using a computer software program and associated sensors and electronics for the purpose of testing, setting and calibrating primary safety relief Corporation was verified by Dunn's Valve Tester's Inc. under the oversight of representatives of the St. Lucie I&C Department and the Mechanical Maintenance Engineering Group.

Safet Evaluation:

This safety evaluation concluded that the use of this computer software did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SAFETY EVALUATION JPN-PSL-SESP-96-054 REVISION 0 EVALUATION OF PRESSURIZED THERMAL SHOCK OF REACTOR VESSEL 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 beltline materials. This information is required for 10 CFR 50.61 submittal. 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 question because it does not involve a change: to the plant Technical Specifications; to plant equipment or procedures; and does not change the plant's licenses. Furthermore, the plants beltline materials meet the 10 CFR 50.61 requirements. Therefore, prior NRC approval was not required for the conditions that were identified within this evaluation.

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SECTION 3 RELOAD SAFETY EVALUATIONS 78

0 PLANT CHANGE/MODIFICATION 112-295 LUCIE UNIT 2 CYCLE 9 RELOAD

~summa r This engineering package provided the reload core design of the St.

Lucie Unit 2 Cycle 9. The Cycle 9 core is designed for cycle lengths up to 13,009 and 12,627 EFPH, depending upon variation in the cycle 8 length of between 12,284 and 11,284 EFPH, respectively.

The cycle lengths for Cycle 9 included an end of cycle inlet temperature coastdown to 535 F followed by a coastdown in power to approximately 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 the replacement of 84 irradiated fuel assemblies with fresh Region L fuel assemblies. The fuel is arranged in a low leakage pattern with no significant differences between the Cycle 9 loading pattern and the Cycle 8 design. The mechanical design of Region L fuel is the 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 length on poison rods) in lieu of A1~03-B<C Burnable Absorber.

b) Changeover from TIG to laser welded Zircaloy intermediate spacer grids.

Safet Evaluate.on:

The safety analysis of this design was performed by Asea Brown Boveri Combustion Engineering Nuclear Operations (ABB CENO) and independently reviewed by Florida Power and Light Co. It has been determined that the operation of the Cycle 9 reload core does not pose an unreviewed safety question and can be implemented with no changes to the St. Lucie Unit 2 Technical Specifications.

Therefore, prior NRC approval was not required for implementation of this modification.

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