ML17352A302

Annual 10CFR50.59 Rept on Changes Test & Experiments for 920701-930601.W/931118 Ltr

ML17352A302
Person / Time
Site:	Turkey Point
Issue date:	11/18/1993
From:	Plunkett T FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-93-290, NUDOCS 9312020007
Download: ML17352A302 (200)

Text

FORD 1 REGULA Y INFORMATION DISTRIBUTIO YSTEM (RIDS)ACCESSION NBR: 9312020007 DOC: DATE: 93/06/Ql NOTARIZED:

NO FACIL: 50-250 Turkey Point Plantz Unit 3z Florida Power and Light C 50-251 Turkey Point Plantz Unit 4I Florida Power and Light C AUTH.NAME AUTHOR AFFILIATION PLUNKETTI T.F.Fl or i da P ower 5 Light Co.REC IP.NAME RECIP IENT AFFILIATION DOCKET 05000250 05000251 R

SUBJECT:

"Annual 10CFR50.59 Rept on Changes Test 8.Experiments" for 920701-930601.

W/931118 ltr.I DISTRIBUTIDN CODE: ZERTD COPIES RECEIVED: LTR ENCL I SIZE: TITLE: 50.59 Annual Report of Changesz Tests or Experiments Made W/out*pprov NOTES: RECIPIENT ID CODE/NAME PD2-2 PD COPIES LTTR ENCL 1 0 REC IP IENT ID CODE/NAME RAGHAVANZ L COPIES LTTR ENCL 2 1 INTERNAI: AEOD/DOA AEOD/DSP/TPAB G>PI'L 02 AEOD/DSP/ROAB NRR/DRCH/HHFB RGN2 FILE 01 EXTERNAL: NRC PDR NSIC 0 NOTE TO ALL RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL D" SK.ROOM P I-37 (EXT.504-2065)TO ELIMLNATE YOUR NAhIE FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T HEED!TOTAL NUMBER OF COPIES REQUIRED: LTTR 1 1 ENCL 10 P.O.Box 029100, Miami, FL, 33102-9100 NOY 18 893 L-93-290 10 CFR 50'9 (b)(2)U.S.Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.20555 Gentlemen:

Re: Turkey Point Units 3 and 4 Docket No.50-250 and 50-251 10 CFR 50.59 Re ort Florida Power&Light, Company's Report on"Changes, Tests and Experiments Made Without Prior Commission Approval" for the period July 1, 1992 through June 1, 1993 is attached.Very truly yours, T.F.Plunkett Vice President Turkey Point Nuclear TFP/GS/rt Attachment cc: Stewart D.Ebneter, Regional Administrator, Region II, USNRC T.P.Johnson, Senior Resident Inspector, USNRC, Turkey Point Plant 280087 9312020007 930601 PDR ADDCK 05000250 3 R PDR'y an FPL Group company ANNUAL 10 CFR 50.59 REPORT FLORlM POSZR&LIGHT COMPASS TURD<3" POINT UNITS 3&4 TURKEY POINT PLANT UNITS 3 AND 4 DOCKET NUMBERS 50 250 AND 50 251 CHANGES'ESTS AND EXPERIMENTS MADE AS ALLOWED BY 10,CFR 50'9 FOR THE PERIOD OF JULY 1g 1992 THROUGH JUNE 1g 1993 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 as described in the SAR, and iii)tests and experiments not described in the SAR which are conducted without prior Commission approval be reported to the Commission at least annually.This report is intended to meet this requirement for the period of July 1, 1992, through June 1, 1993.This report is divided into five (5)sectionsg the first, changes to the facility as described in the SAR performed by a Plant Change/Modification (PC/M);the second, changes to the facility or procedures as described in the SAR not performed by a PC/M and tests and experiments not described in the SAR)the third, a summary of any fuel reload evaluationsg the fourth, a list of Power Operated Relief Valve (PORV)actuations, which is submitted in accordance with FPL's commitment to comply with the requirements of Item II.K.3.3 of NUREG 0737;the fifth, a summary of the findings of Steam Generator tube inspections.

Both Unit 3 and Unit 4 had Steam Generator tube inspections during this reporting period.

TABLE OF CONTENTS PAGE SECTION 1 PLANT CHANGE MODIFICATIONS 84-83 85-54 85-105 85-154 86-045 86-79 87-140 87-220 87-245 87-279 87-409 88-087 88-099 88-100 88-346 CATHODIC PROTECTION FOR CCW AND TPCS HEAT EXCHANGERS

-05/27/93 FUEL TRANSFER SYSTEM CABLE DRIVEMODIFICATIONS

-05/28/93 MAIN FEEDWATER BYPASS AIR SUPPLY SOLENOID VALVES-02/05/93 PROTECTIVE DOORS FOR 3D01 DC DISTRIBUTION PANEL BREAKERS-02/10/93 AUXILIARY FEEDWATER TURBINE EXHAUST SILENCER CONDENSATE REMOVAL-08/20/92 SIMULATOR TRAINING FACILITY-04/24/93 HVAC DUCTWORK TEST HOLE COVERS-12/16/92 5KV ELECTRICAL PENETRATIONS MP-0728 LEAK RATE REVISION-12/21/92 VELAN VALVE TAG NO.3-333 COMPONENT SUBSTITUTION

-05/28/93 ELECTRICAL AND I&C DRAWING SEPARATION AND RENUMBERING

-07/23/92 INSTALLATION OF SPARE SAFETY INJECTION PUMP MOTOR'-05/15/93 PRESSURE SPRAY VALVE PRESSURE CONTROLLER (PC-444C/D)

LOW LIMIT CIRCUIT DEFEAT 01/15/93 PRIMARY WATER STORAGE DEAERATED'WATER TRANSFER PUMP HI LEVEL START SWITCH MODIFICATION

-03/16/93 PRIMARY WATER STORAGE DEAERATED WATER TRANSFER PUMP HI LEVEL START SWITCH MODIFICATION

-03/16/93 TURBINE PLANT COOLING WATER ISOLATION VALVE MODIFICATION

-05/10/93 12 13 14 15 16 17 18 19 20 21 22 23 24 25 - 88-450 88-527 88-534 89-095 89-100 89-475 89-491 89-512 t 89-542 90-193 90-194 90-239 90-240 90-396 90-397 90-445 90-446 RCP MOTOR REFURBISHMENT AND UPGRADE-04/12/93 RESOLUTION OF DRAWING CHANGES ASSOCIATED WITH 5610-T-E-4503

-07/23/92 DRAWING DISCREPANCIES ON 5610-T-E-4534, SHEETS 1 AND 2 CONTAINMENT VENTILATION SYSTEM-12/29/92 DRAWING UPDATE 5610-T-E-4065, SHEETS 2 AND 3 LUBE WATER AND CIRCULATING WATER SYSTEM 08/18/92'RAWING UPDATE 5610-T-E-4061, SHEET 1 LUBE WATER AND CIRCULATING WATER SYSTEM 07/23/92 DRAWING DISCREPANCIES ON 5610-T-E-4501, SHEET 1 REACTOR COOLANT-04/23/93 STRUCTURAL STEEL ATTACHMENT FIREPROOFING REQUIREMENTS

-01/11/93 NUCLEAR OPERATIONS/CHEMISTRY BREAK AREA AND CONTROL POINT-05/13/93 DRAWING UPDATE-ISI QUALITY GROUP CLASSIFICATIONS/BOUNDARIES

-05/10/93 ADDITION OF APPENDIX R BYPASS SWITCH FOR LCV-3-460-10/08/92 ADDITION OF APPENDIX R BYPASS SWITCH FOR LCV-4-460-04/19/93 C BUS SWITCHGEAR CONTROL AND PROTECTION POWER ISOLATION FOR APPENDIX R-05/10/93 RAD-3-6417 SAMPLE LINE END CABINET MODIFICATION

-11/17/92 NIS RECORDER CHANNEL SELECTOR SWITCHES 10/16/92 NIS RECORDER CHANNEL SELECTOR SWITCHES 05/03/93 DRILLING OF VALVE WEDGE FOR MOV-3-872 11/16/93 WATER TREATMENT PLANT IN-LINE MONITORING 09/23/92 26 27 28 29 30 31 32 33 34 35 37 39 40 41 42 43 44 90-449 91-128 CCW AREA PIPE TRENCH FLOODWALLS

-08/13/92 480V UNDERVOLTAGE PROTECTION SCHEME MODIFICATION 128 45 46 91-130 91-133 91-166 91-198 92-004 92-033 92-034 92-040 92-054 92-057 92-058 92-059 92-063 92-073 92-074 92-075 92-079 PROCESS RADIATION MONITORING SYSTEM R-3-11 AND R-3-12 REPLACEMENT

-10/21/92 REPLACEMENT OF 480 VOLT MOTOR CONTROL CENTER 3E-07/31/92 REPLACEMENT OF SEAL TABLE FITTINGS AND THIMBLE TUBE LENGTHENING

-11/29/92 REPAIR AND MODIFICATIONS OF THE UNIT 3 INTAKE STRUCTURE-11/14/92 UPGRADING PLANT PAGE AUDIBILITY

-05/25/93 EMERGENCY BUS LOAD SEQUENCER MODIFICATION 10/21/92 EMERGENCY BUS LOAD SEQUENCER MODIFICATIONS 05/03/92 ADDITION OF REVERSE POWER RELAY AND MAIN GENERATOR PROTECTION MODIFICATIONS

-12/01/92 480V UNDERVOLTAGE PROTECTION SCHEME MODIFICATION

-05/01/93 HHSI THERMAL RELIEF VALVE-11/09/92 PROCESS RADIATION MONITORING SYSTEM R-4-12 AND R-4-12 REPLACEMENT

-05/04/93 CONTROL ROOM AIR CONDITIONING AND VENTILATION SYSTEM CONTROL MODIFICATION

-11/16/92 REACTOR COOLANT PUMP 3B MOTOR REFURBISHMENT/

UPGRADE-11/11/92 ADDITION OF REVERSE POWER RELAY AND MAIN GENERATOR PROTECTION MODIFICATION

-05/25/93 CORE EXIT THERMOCOUPLE SEAL UPGRADE-11/08/92 CORE EXIT THERMOCOUPLE SEAL UPGRADE-05/11/93 REPAIR AND MODIFICATION OF THE UNIT 4 INTAKE STRUCTURE-0-4/30/93 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 92-097 92-102 92-108 92-110 92-124 92-163 92-166 92-181 93-009 93-020 ALTERNATE SAFETY INJECTION THERMAL RELIEF VALVE MODIFICATION

-05/12/93 REPLACEMENT OF RAW WATER STORAGE TANK I (T63A)-03/15/93 REPLACEMENT OF RAW WATER AND SERVICE WATER SYSTEM DAMAGED BY HURRICANE ANDREW-03/19/93 INSTALLATION OF A DUCT BANK FROM MH 610 TO MH 324-12/29/92 OFFSITE RADIO COMMUNICATIONS PROJECT 03/30/93 REPLACEMENT OF SEAL TABLE FITTINGS AND THIMBLE TUBE LENGTHENING

-05/22/93 NIS SOURCE RANGE DETECTOR REPLACEMENT 04/15/93 ELIMINATION OF TURBINE RUNBACK ON DROPPED ROD-04/30/93 INSTALLATION OF JIB CRANE IN THE UNIT 4 CONTAINMENT BUILDING AT ELEVATION 58'-0 05/07/93 REACTOR COOLANT PUMP 4A MOTOR REFURBISHMENT/

UPGRADED-05/21/93 64 65 66 67 68 69 70 71 72 73 SECTION 2 SAFETY EVALUATIONS86-011 86-033 86-067 86-433E 87-384 SAFETY EVALUATION FOR CPWOs86-017 AND 86-018 UNIT 4 REPLACEMENT OF NORMAL&EMERGENCY CONTAINMENT COOLER DRIP PANS-03/15/93 SAFETY EVALUATION FOR CPWO 86-086 RELOCATION OF EMERGENCY DIESEL GENERATOR COOLER SYSTEM DRAIN VALVES 293A AND 293B-05/13/93 SAFETY EVALUATION FOR CPWOs86-163 PASS CHLORIDE REAGENT AND CALIBRATION STANDARD PUMPS SUBSTITUTION

-03/29/93 SAFETY EVALUATION FOR CPWO 86-035 UNIT 4 VALVE POSITIONER REPLACEMENT FOR PCV-4-455 A&B-03/23/93 SAFETY EVALUATION FOR CPWOs87-060 AND 87-061 PRMS DRAWER REPLACEMENT

-03/29/93 75 76 77 78 79 DE-ENERGIZATION OF UNIT 4 4160 VOLT SAFETY RELATED BUSSES-04/08/93 SAFETY EVALUATION OF THE DELETION OF FIRE HOSE STATIONS IN THE RADWASTE BUILDING-08/11/92 DE-ENERGIZATION OF UNIT 3 4160 VOLT SAFETY RELATED BUSSES-10/08/92 THE CONDUCT OF ZNTEGRATED SAFEGUARDS TESTING ON A SHUTDOWN UNIT WITH THE OPPOSITE UNIT AT POWER-08/06/82-03/11/93 SAFETY EVALUATION FOR LOAD CENTER AND RELAY SETTINGS CHANGES-01/07/93-05/21/93 EVALUATION OF IMPACT OF ACCUMULATOR DISCHARGE TEST ON FUEL AND REACTOR INTERNALS-08/20/92 04/16/93 EVALUATION OF ACCUMULATOR DISCHARGE TEST WITH REACTOR VESSEL HEAD INSTALLED-09/25/92 TEMPORARY LEAD SHIELDING INSTALLATION SPECIFICATION SPEC-C-003

-08/04/92 UNIT 4 TWENTIETH YEAR CONTAINMENT TENDON SURVEILLANCE

-08/11/92 THE DEMOLITION OF THE TURKEY POINT FOSSIL UNIT 1 CHIMNEY-09/03/92-09/04/92 10/02/92 FREEZE SEAL EVALUATION FOR REPLACEMENT OF VALVES 3-777, 3-834 AND 3-833-08/13/92 SAFETY EVALUATION RELATED TO THE TURKEY POINT FOSSIL UNITi2 CHIMNEY-09/19/92-09/24/92 INTERIM FIRE PROTECTION SYSTEM CONFIGURATION TO SUPPORT UNIT 4 STARTUP-09/24/92 09/25/92-09/30/92 SAFETY EVALUATION RELATED TO THE TURKEY POINT FOSSIL UNITS 1 AND 2 CHIMNEY CONSTRUCTION ACTIVITIES

-11/27/92 SAFETY EVALUATION RELATED TO THE NEW TURKEY POINT FOSSIL UNIT 1 CHIMNEY AND UNIT 2 CHIMNEY REINFORCEMENT

-12/08/92 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 92-044 92-045 MANUAL OVERRIDE OF MOV-*-626 DURING RCP SEAL FAILURE-09/30/92 FREEZE SEAL INSTALLATION ON THE HHSI ALTERNATE HOT LEG INJECTION CROSS-TIE PIPING-08/21/92 95 96 92-052 92-056 92-059 92-060 92-061 SAFETY EVALUATION FOR ICW VALVE REPLACEMENT 10/15/92 INSTALLATION OF COMMUNICATION ANTENNAS-09/22/92 98 97 UNIT 3 REFUELING OUTAGE CONTINGENCY PLAN FOR EMERGENCY POWER TO THE SFP PUMPS-10/08/92 INSTALLATION AND USE OF AN ABB/CE RCCA INSPECTION STATION AT TURKEY POINT-10/09-92 99 100 EVALUATION FOR TSA 03-92-06-12 FIRE WATER PUMP TRIP 101 UPON LOOP DURING 4160 VOLT BUS 3A DE-ENERGIZATION 10/12/92 92-063 92-066 92-070 92-071 92-072 93-007 93-009 93-010 93-011 93-015 THE INSTALLATION OF COMMUNICATION ANTENNAS (TP-907)-10/30/92 FREEZE SEAL SAFETY EVALUATION FOR REPAIR OF CV-3-244-11/06/92 REPLACEMENT OF CRDM 4A COOLER FAN MOTOR AT POWER OPERATION-ll/24/92 SAFETY EVALUATION FOR ALLOWING A MAN-BASKET TO REMAIN WITHIN CONTAINMENT DURING ALL MODES OF OPERATION-11/20/92 SAFETY EVALUATION FOR LT-3-494 VENT PATH MODIFICATION

-11/27/92 TEMPORARY REMOVAL OF STEAM GENERATOR 4C THRUST BEAM-03/30/93 MACHINING OF MOTOR OPERATED VALVE STEMS FOR INSTALLATION OF STRAIN GAUGES-SPECIFICATION SPEC-M-009

-03/16/93 INTAKE COOLING WATER VALVE REPLACEMENTS AND B HEADER CRAWL THROUGH INSPECTION

-03/25/93 04/08/93 OMS SETPOINT DURING RCP OPERATION-05/11/93 SAFETY EVALUATION FOR ACCEPTABLE UPPER AND LOWER TIME DELAY LIMITS FOR ECC 4A AND ECF 4A AGASTAT LOAD SEQUENCING RELAYS-05/19/93 102 103 104 105 106 107 108 109 110 93-017 93-018 93-019 EVALUATION FOR LOOSE OBJECTS IN THE SECONDARY SIDE OF STEAM GENERATOR C AT TURKEY POINT UNIT 4-05/06/93 SAFETY EVALUATION FOR STEAM GENERATOR C SECONDARY SIDE FOREIGN OBJECTS-05/01/93 SAFETY EVALUATION FOR STEAM GENERATOR A SECONDARY SIDE FOREIGN OBJECTS-05/01/93 112 113 114 SECTION 3 RELOAD SAFETY EVALUATZONS91-108 92-045 TURKEY POINT UNIT 3 CYCLE 13 RELOAD SAFETY AND LICENSING CHECKLIST-12/18/92 TURKEY POINT UNIT 4 CYCLE 14 RELOAD SAFETY AND LICENSING CHECKLIST-05/21/93 116 117 SECTION 4 ANNUAL REPORT OF POSER OPERATED RELZEF VALVE PORV ACTUATIONS UNITS 3 AND 4 119 t SECTION 5 STEAM GENERATOR TUBE INSPECTZONS FOR TURKEY POINT UNIT 3 UNIT 4 121 136 0

SECTION PLANT CHANGE/MODIFICATIONS 10 PLANT CHANGE MODIFICATION 84-83 UNIT TURN OVER DATE 3 05/27/93 CATHODZC PROTECTZON FOR CCW AND TPCW HEAT EXCHANGERS

~mumm ar This design package covered the installation of an impressed cathodic protection system for each of the Turbine Plant Cooling Water (TPCW)Heat Exchangers and the Component Cooling Water (CCW)Heat Exchangers for Turkey Point Unit 3.Included in the modification was the installation of two reference cell electrodes in each of the CCW Heat Exchangers and the TPCW Heat Exchangers.

The button anodes originally installed in the TPCW Heat Exchangers channel covers were replaced with probe anodes.No modification was required in regards to the anodes installed in the CCW Heat.Exchangers.

Safet Evaluation:

The cathodic protection system was design to improve the longterm useability of the equipment.

Its misuse or failure would not hinder the functional ability of the heat exchangers to mitigate the consequences of an accident or to maintain safe shutdown conditions, since sufficient administrative controls existed to control and identify operating problems.In addtion, the impairment of either heat exchanger's functional ability would be a longterm process recognizable during routine maintenance activities.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

11 PLANT CHANGE MODIFICATION 85-54 UNIT TURN OVER DATE 4 05/28/93 FUEL TRANSFER SYSTEM CABLE DRIVE MODIFICATIONS

~summa r This design package provided the engineering and design necessary to upgrade the Turkey Point Unit 4 fuel transfer system.The fuel transfer system was modified to provide a more reliable operating system and reduce the amount of equipment under water.This resulted in a more ALARA effective system.These modifications involved the following:

(1)fuel transfer system traverse drive source modifications; (2)addition of upender lifting frame counterweights, winch cable and bushings;and, (3)addition of upender winch load monitors and quick-disconnect control consoles.Safet Evaluation:

As described in Appendix 5A of the Turkey Point Plant Units 3 and 4 Updated FSAR, the fuel transfer system does not perform a safety related function.However, there is a very low probability that a fuel handing accident could result as described in Chapter 14.2 of the Updated FSAR.In order to minimize the effects of these potential events,.the fuel handling system modifications were designed to withstand all applicable load combinations, including seismic loads, in accordance with Updated FSAR criteria.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.

12 0 0 0 PLANT CHANGE MODIFICATION 85-105 UNIT~3 TURN OVER DATE: 02/05/93 m MAIN FEEDPFATER BYPASS AIR SUPPLY SOLENOID VALVES~summa This modification replaced the existing main feedwater bypass air supply solenoid valves.The existing model of ASCO solenoid valves were replaced with another model of ASCO solenoid valves, which were qualified and had a better temperature rating.The replacement solenoid valves have the same overall dimensions and no system alterations were required.Safet Evaluation:

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

13 PLANT CHANGE MODIFICATION 85-154 UNIT~3 TURN OVER DATE: 02/10/93 PROTECTIVE DOORS FOR 3D01 DC DZSTRZBUTZON PANEL BREAKERS~summa r This design package provided for the installation of protective doors on three of the six subpanels of the 3D01 DC distribution panel.This DC distribution panel is located east of the Cable Spreading Room in the Auxiliary Building and east of the Unit 3 Motor Generator sets.These protective doors cover the breakers located in the lower half of the subpanels to prevent the inadvertent closing of these breakers which may cause the unit to trip.The doors were made of expanded metal sheets with a sheet metal frame.The doors were connected to the exterior sheet metal skin of the subpanels using sheet metal tapping screws and three steel hinges.The hinges and fasteners were required to support the deadweight of the door and seismic loads.The stresses induced in the metal door panels and the sheet metal frames were found to be within the allowable capacities of the materials used.Safet Evaluation:

The installation of protective doors on the 3D01 DC distribution panel did affect the electrical function of the panel and therefore, did not perform a safety related function.However, the doors were designed and installed so as not to inhibit the safety functions of the DC distribution panel itself.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

14 II PLANT CHANGE MODIFICATION 86 045 UNITS TURN OVER DATE 3&4 08/20/92 AUXILIARY FEEDRATER TURBINE EXHAUST SZLENCER CONDENSATE REMOVAL~8llRBIR The modifications provided in this engineering package will ensure the adequate removal of condensate from the turbine exhaust line, precluding the discharge of hot condensate from the silencers upon AFW pump actuation.

The modifications direct the turbine exhaust drain discharge through a 1-1/2" header to an area where it will not be a hazard to personnel.

The modifications in this engineering package consisted of locking the existing manual isolation valves in the open position, enlarging the discharge piping downstream of the existing steam orifices, and connecting the discharges to a new 1-1/2 inch drain header.These modifications were required to ensure adequate condensate removal from the turbine exhaust lines to preclude the discharge of hot condensate from the silencers.

The existing turbine exhaust drains currently discharge to a drainage trench in the Auxiliary Feedwater pump area.The modifications direct the turbine exhaust piping drain discharge to a storm drain in the Unit 3 Steam Generator Blowdown Tank Area by way of 1-1/2 inch header.The isolation valves were locked in an open position, which will prevent the inadvertent closure of the valve.Safet Evaluation:

The modifications for connecting existing turbine casing and exhaust drains and the Trip and Throttle valve stem packing high pressure leakoff drains to a new 1-1/2 inch header were required to ensure adequate condensate removal for personnel protection.

The modifications did not have an adverse impact on the Auxiliary Feedwater (AFW)System.In the case of the turbine exhaust drains, the AFW turbine was only removed from service within the conditions allowed by the plant Technical Specifications.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

15 e

PLANT CHANGE MODIFICATION 86 79 UNITS TURN OVER DATE 3&4 04/24/93 8ZMULATOR TRAZNING FACILITY~8UBUIIR This engineering package addressed the addition of the Turkey Point Simulator Training Facility.This facility was constructed in order to satisfy NRC training requirements.

The structure provides a facility to train the nuclear plant operators in a simulated control room, as well as, provide training for other operations and maintenance activities.

The facility is located on the southwest corner of the site, outside of the plant security fence.The facility is a two story reinforced concrete and masonry structure.

The building contains the simulator control room, computer room, classrooms, offices, library and maintenance training areas.Building utilities (power, telecommunications, sanitary, potable water, and fire water)are tied to existing site systems.A paging system designed to extend the Site Evacuation Alarm into the building is also included.8afet Evaluation:

The Simulator Training Facility engineering package did not modify or affect any plant nuclear safety related systems nor.does it perform an automatic nuclear safety related function.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

16 PLANT CHANGE MODZPZCATZON 87 140 UNITS TURN OVER DATE 3 6 4 12/16/92 HVAC DUCTWORK TEST HOLE COVERS~mamma This engineering modification provided for the installation of HVAC flow test hole covers in the ductwork of all existing HVAC systems in the Units 3 and 4 Containment, Auxiliary Building, Radwaste Building, Fuel Handling Building, Turbine Building and Control Building.The test hole covers provided a 1-1/8 inch access port for a portable flow measurement probe.The addition of test hole covers enhanced the ability to obtain flow measurements, assess HVAC system performance, and allow for proper balancing of the HVAC systems.The installation and location of the test hole covers was performed in accordance with ASHRAE Standards.

Safet Evaluation:

The test hole covers are designed and constructed consistent with existing HVAC systems and ASHRAE standards.

A number of factors were considered in the evaluation of this modification.

Among them was the increase of bulk material inventory inside the containment which was expected to have a negligible effect on the ECCS heat sink analysis and on the potential for hydrogen generation as stated in Updated FSAR.The modifications in this Engineering Package did not have any adverse effect on plant safety or plant operations.

This modifications 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 87-220 UNITS 3&4 TURN OVER DATE: 12/21/92 5KV ELECTRICAL PENETRATIONS MP 0728 LEAK RATE REVISION~summar This engineering package was issued to increase the leak rate acceptance threshold for the 5kV electrical penetrations to make leakage testing practicable.

This PC/M evaluated the acceptability of a change in the leakage acceptance criteria and provided the basis for changes in the plant maintenance procedure MP-0728.The leak rate testing performed by MP-0728 is identified in this procedure as a 10 CFR 50, Appendix J"Type B test" and was intended to detect local leaks across the pressure boundary formed by the electrical penetration assembly.It was discovered that the leakage criteria that was originally contained in Maintenance Procedure MP-0728 actually originated from IEEE Std.317-1983, which prescribes requirements for post-installation testing and was not intended for post-maintenance testing.Safet Evaluation:

Since the leakage criteria of the IEEE standard is less than one thousands of one percent of the total allowable leakage from all containment penetrations, a nominal increase in the leakage rate of each 5kV electrical penetration was considered negligible compared to the size of the total allowable leak rate specified for all containment penetrations.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 MODIPICATION 87-245 UNIT TURN OVER DATE 3 05/28/93 VELAN VALVE TAG NOa 3-333 COMPONENT SUBSTITUTION

~mumm ar This engineering package provided the engineering basis for a change in material for the packing washer on the Velan valve 3-333.The subject Class 2 safety related valve is in a CVCS charging line to the RCS Loop A Cold Leg that is required to function to provide one means of reactivity control to satisfy NRC requirements.

The function of the packing washer, is to precede the packing for effective seating and sealing at the base of the packing gland.Velan originally supplied the subject 3-inch bonnetless bypass valve around HCV-121 with a packing washer made from ASTM-A276 (SS304).Velan recommended a change in the material of this washer to ASTM-A 564 (SS603)and this engineering package was developed to facilitate this change.Safet Evaluation:

The above part was evaluated as to its appropriate ASTM standard and found to have equal or better.strength characteristics as compared to the originally specified material.The differences in chemical composition between the existing and proposed material are negligible, thus, are acceptable.

The corrosion resistance of the new material was considered to be comparable to the original material and could be used in the RCS.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

19 PLANT CHANGE MODIFICATION 87-279 UNITS 3&4 TURN OVER DATE: 07/23/92 ELECTRICAL AND Z&C DRA'HING SEPARATION AND RENUMBERING

~summa The scope of this engineering package was the enhancement of electrical and I&C drawings under a program which maintains traceability to and the continuing fidelity of plant drawings.The drawing enhancement and renumbering provided by this engineering package was initiated with the intent of improving the usefulness of the existing plant drawings for plant personnel.

This also was intended to ensure that future modifications were properly designed and implemented, as well as making interpretation of the drawings much easier.To alleviate possible confusion resulting from the continued use of common drawings for the implementation of unit specific design changes, the plant drawings were split into separate drawings for Unit 3 and Unit 4.Additionally, many drawings were further split, such that, only one piece of equipment was depicted on each drawing.The drawings were enhanced by the addition of reference information that was typically missing from the existing plant drawings.Title blocks were standarized so that equivalent information was shown in the same place on every title block.A parallel drawing numbering system was used, such that, a number used on one unit is reserved for the same purpose on the other unit.A drawing cross-reference which ties existing drawing numbers to new drawing numbers was created.Included under the drawing enhancement scope was the incorporation of"as-built" NCRs associated with the Emergency Load Sequencers and Emergency Diesel Generators that were dispositioned during the 1987 Unit 3 and Unit 4 outages.In all instances, the new drawings were prepared and verified by engineering personnel using established project procedures.

Safet Evaluation:

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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

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

20

PLANT CHANGE MODIF1CATION 87-409 UNITS 3&4 TURN OVER DATE: 05/15/93 INSTALLATION OF SPARE SAFETY INJECTION PUMP MOTOR~summa This engineering package provided for the replacement of any of the four safety injection pump motors at Turkey Point Units 3 and 4 with a spare motor purchased by FPL from Westinghouse.

This engineering package ensured that the spare motor could be installed to replace any one of the four existing motors in the event of a failure or for maintenance on an existing motor.The availability of a spare motor will preclude any lengthy outage while an installed motor is required or a new motor is purchased.

In addition, analysis was performed and documentation provided to ensure the spare motor meets all seismic and environmental qualification requirements.

This engineering package also ensured the interchangeability of the safety injection'ump motors among themselves.

Motors which have been replaced by the new spare motor may themselves eventually become a spare motor used to replace any of the installed pump motors.Safet Evaluation:

The safety injection pump motors are Class I equipment that power safety injection pumps, which are intended to automatically deliver cooling water to the reactor core in the event of a loss-of-coolant-accident or main steam line break.The spare motor was purchased to safety related requirements and is seismically and environmentally qualified for its intended application.

In addition, the spare motor is capable of meeting the safety injection pump performance characteristics identified in the Updated FSAR.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

21 PLANT CHANGE MODIFICATION 88-087 UNIT~4 TURN OVER DATE: 01/15/93 PRESSURIZER SPRAY VALVE PRESSURE CONTROLLER PC-444C D LOS LIMIT CIRCUIT DEFEAT~8UDBSRK This engineering package provided for the defeat of the low limit circuit internal to pressurizer spray valve pressure controllers (PC-444C and D).This was accomplished by lifting one lead for this circuit.The pressurizer spray valves were originally equipped with an electro/pneumatic (I/P)converter which converts an input signal (from PC-444C and D)to a corresponding pressure output signal.This pressure signal is then used in conjunction with a valve positioner to precisely control PCV-455A and B.The I/P converter is located inside containment.

Due to ambient temperature variations, the I/P converter had a tendency to drift causing air to be supplied to the spray valves continuously.

This air supply caused the spray valves to remain open even with a zero percent open signal present.Safet Evaluation:

The pressurizer spray valve controls are not required for any design basis event, do not perform a safety related function, and do not interface with safety related systems or components.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 MODIFICATION 88-099 UNIT~3 TURN OVER DATE: 03/16/93 PRIMARY RATER 8TORAGE DEAERATED WATER TRANSFER PUMP HI LEVEL 8TART 8WITCH MODIFICATION

~summa This engineering modification retagged and repulled cables for level switches LS-1529 and LS-1532, located on the primary water storage deaerator.

This was implemented to correct a mismatch in relative elevations of the deaerator level switch (associated with LT-1532)versus the deaerator water transfer pump start-signal level switch (LS-1529).

The elevation of level switch LS-1529 was higher than the upper range of the deaerator level transmitter (LT-1532), which prevented the transfer pump from starting automatically without installing a temporary jumper cable.The elevation of level switch LS-1529 was also near the vacuum pump high level trip (LS-1552).

After this modification, the proper automatic operating scheme for the deaerator was restored, with the water level in the primary water storage deaerator rising above the elevation require to start the transfer pumps without causing the vacuum pump to trip.8afet Evaluation:

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 MODZFICAT10N 88-I.OO UNIT TURN OVER DATE 4 03/16/93 PRIMARY WATER STORAGE DEAERATED WATER TRANSFER PUMP HZ LEVEL START SWITCH MODZFICATION

~summa This engineering modification retagged and repulled cables for level switches LS-1529 and LS-1532, located on the primary water storage deaerator.

This was implemented to correct a mismatch in relative elevations of the deaerator level switch (associated with LT-1532)versus the deaerator water transfer pump start-signal level switch (LS-1529).

The elevation of level switch LS-1529 was higher than the upper range of the deaerator level transmitter (LT-1532), which prevented the transfer pump from starting automatically without installing a temporary jumper cable.The elevation of level switch LS-1529 was also near the vacuum pump high level trip (LS-1552).

After this modification, the proper automatic operating scheme for the deaerator was restored, with the water level in the primary water storage deaerator rising above the elevation require to start the transfer pumps without causing the vacuum pump to trip.Safet Evaluation:

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 modif ication.24

PLANT CHANGE MODXFXCATXON 88 346 UNIT~4 TURN OVER DATE: 05/10/93 TURBINE PLANT COOLXNG WATER XSOLATXON VALVE MODXFXCATXON

~summa This modification provided for the addition of controls to Turbine Plant Cooling Water (TPWC)Isolation Valves 50-4-314 and 50-4-334, such that, these valves would close on a Safety Injection Actuation Signal (SIAS).TPCW isolation valves with pneumatic operators were installed to replace manually operated valves by an earlier engineering modification.

However, the early modification did not provide for the connection of the valve control circuits.This engineering package installed the controlsg instrument air and electrical control circuits to operate existing pilot solenoid valves.Automatic isolation of TPCW during accident conditions ensures required Intake Cooling Water (ICW)flow to the Component Cooling Water (CCW)-heat exchangers.

Consequently, the modification resolved the single failure concerns associated with valve CV-4-2201 as described in Justification for Continued Operation (JCO)86-003 and provided a basis for eliminating Unit 4 from the corrective action requirements of this JCO.Safet Evaluation:

Isolation of the TPCW system occurs following a design basis accident (DBA)to ensure adequate Intake Cooling Water (ICW)flow is diverted to the Component Cooling Water (CCW)heat exchangers for design basis accident heat load removal in the event of a single failure that results in one ICW pump being available.

Valves 50-4-314 and 50-4-334 are spring-to-close, fail-closed on loss of air and electrical power, and are provided with 125 VDC pilot solenoid valves.The modification ,in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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

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

25 0

PLANT CHANGE MODIFICATION 88-450 UNIT~4 TURN OVER DATE: 04/12/93 RCP MOTOR REFURBISHMENT AND UPGRADE~summa This engineering package provided for the refurbishment and upgrade of the 4B reactor coolant pump motor.The design bases established in the Updated FSAR were reviewed and determined to be unaffected, because thee modifications met all FSAR criteria stipulated for the original design.In addition, these modifications did not impact any Technical Specifications.

The original installed motor was replaced with a spare motor which was refurbished at the Westinghouse Electro-Mechanical Division facility.This refurbishment consisted of inspection and maintenance activities performed to the existing design specifications.

In addition, two upgrade modifications were performed, concurrent with the refurbishment, to ensure consistency with the latest RCP technology and to realize additional reliability and availability.

These modifications consisted of an upgrade to the oil lift system and a redesign of the lower cooling coil.In the past, the lower cooling coil had been susceptible to handling damage due to the use of bronze flanges, copper pipes, and soldered/brazed joints.These were replaced with steel pipe and fittings with welded joints and heavier 90/10 copper/nickel cooling coil tubing.-The oil lift system upgrade included such improvements as stainless steel lines, flow control valves, elimination of the 3-way valve system pressure switch settings change, and an enhanced oil lift pump.Safet Evaluation:

This package is classified as safety related, since it performs work on the lower bearing cooling coil which is considered part of the safety related CCW system.The design bases established in the Updated FSAR were reviewed and determined not to be affected, because these modifications meet all Updated FSAR criteria stipulated for the original design.In addition, these modifications did not impact any Technical Specifications.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

26 0 0 PLANT CHANGE MODIFICATION 88-527 UNITS TURN OVER DATE 3&4 07/23/92 RESOLUTION OF DRYING CHANGES ASSOCIATED RITH 56i.O-T E-4503~summa r This engineering package provided a basis to evaluate and resolve the existing outstanding Requests for Engineering Assistance (REAs)and Non-Conformance Reports (NCRs)associated with Units 3 and 4 safety related systems.Zn this way, discrepancies between the as-built condition of the plant and the existing Plant Operating Documents were reconciled.

This engineering package documented drawing changes associated with 5610-T-E-4503, Sheet 1, which resulted from discrepancies identified by three REAs and an NCR.The changes requested by the REAs included in this package only involved redesignation of valve type or classification.

Safet Evaluation:

The engineering evaluation for this modification concluded that the subject would not alter the plant's design basis and were bounded by existing design analysis.Further, there would be no adverse effect on the plant's systems, structures or components.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 88-534 UNITS 3&4 TURN OVER DATE: 12/29/92 DRA'NING DISCREPANCIES ON 5610-T E-4534 SHEETS 1 AND 2 CONTAINMENT VENTILATION SYSTEM~summa r This engineering package was developed to correct drawing discrepancies identified on Operating Diagram 5610-T-E-4534, Sheets 1 and 2.A field verification walkdown of theContainment Ventilation System and a review of Operating Diagram 5610-T-E-4534, Sheets 1 and 2, revealed several drawing discrepancies.

These discrepancies involved valve positions for normal operation, the lack of a drawing showing service air, HVAC damper and ductwork locations, flow direction discrepancies, and valve-type discrepancies.

The changes to resolve all discrepancies were evaluated by Engineering using nonconformance reports and found to be acceptable.

Safet Evaluation:

Based on an evaluation contained in this engineering package, these discrepancies were determined not to adversely impact plant systems, structures or components.

Further, these drawing changes did not alter the plant's design basis and were bounded by existing design analyses.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 PLANT CHANGE MODIFICATION 89-095 UNIT TURN OVER DATE 4 08/18/92 DRAWING UPDATE 5610-T-E-4065 SHEETS 2 AND 3 LUBE WATER AND CIRCULATING WATER SYSTEM~Smmmar This engineering package was developed to correct drawing discrepancies on Operating Diagram 5610-T-E-4065, Sheets 2 and 3.A field verification walkdown of the Unit 4 Lube Water and Circulating Water Systems and a review of Operating Diagram 5610-T-E-4065, Sheets 2 and 3, revealed several drawing discrepancies.

These discrepancies involved setpoints, value positions, test connection locations, and valve types.The changes to resolve all discrepancies were evaluated by Engineering using nonconformance reports and found to be acceptable.

Safet Evaluation:

Based on an evaluation contained in this engineering package, these discrepancies were determined not to adversely impact plant systems, structures or components.

Further, these drawing changes did not alter the plant's design basis and were bounded by existing design analyses.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.

29

PLANT CHANGE MODIFICATION 89-100 UNIT~4 TURN OVER DATE: 07/23/92 DRAWING UPDATE 5610-T E-4061 SHEET LUBE WATER AND CIRCULATING WATER SYSTEM~summa This engineering package was developed to correct drawing discrepancies on Operating Diagram 5610-T-E-4061, Sheet 1.A field verification walkdown of the Unit 4 Main Steam System and a review of Operating Diagram 5610-T-E-4061, Sheet 1, revealed several drawing discrepancies.

These discrepancies involved steam trap identification tagging, steam trap isolation and drain valves, small valves tagging designations, and small bore piping and valve configuration for the Moisture Separator Reheater nitrogen system.The changes to resolve all discrepancies were evaluated by Engineering using nonconformance reports and found to be acceptable.

Safet Evaluation:

a Based on an evaluation contained in this engineering package, these discrepancies were determined not to adversely impact plant systems, structures or components.

Further, these drawing changes did not alter the plant's design basis and were bounded by existing design analyses.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.

30 PLANT CHANGE MODIFICATION 89-475 UNITS TURN OVER DATE 3&4 04/23/93 DRAWING DISCREPANCIES ON 5610 T E-4501 SHEET REACTOR COOLANT~summa This engineering package was developed to correct drawing discrepancies identified on Operating Diagram 5610-T-E-4501, Sheet 1.A field verification walkdown of small bore piping for the Reactor Coolant System and a review of Operating Diagram 5610-T-E-4501, Sheet 1, revealed several drawing discrepancies.

These discrepancies involved incorrectly labeled valves;the exact locations of several small valves, blind flanges, and instrument taps;and piping caps.The changes to resolve all discrepancies were evaluated by Engineering using nonconformance reports and were found to be acceptable.

Safet Evaluation:

Based on an evaluation contained in this engineering package, these discrepancies were determined not to adversely impact plant systems, structures or components.

Further, these drawing changes did not alter the plant's design basis and were bounded by existing design analyses.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 PLANT CHANGE MODIFICATION 89 491 UNITS 3&4 TURN OVER DATE: 01/11/93 STRUCTURAL STEEL ATTACHMENT FZREPROOFING RE UIREMENTS~summa This engineering package reevaluated the fireproofing requirements applicable to structural steel attachments that penetrate the fireproofing envelope and implemented drawing changes which required modifications to the structural steel fireproofing in four rooms.This modification also provided a basis for changes in plant inspection and maintenance procedures associated with structural steel fireproofing requirements.

During the periodic reinspections of structural steel fireproofing, as required by Plant Procedure O-SMM-016.3"Fire Barriers and Structural Steel Fireproofing Inspection," questions were raised relative to implementation requirements for fireproofing attachments which penetrate the structural steel fireproofing envelope.Although details for attachments are shown on drawings 5610-A-181, Sheet 1 and 2, the requirements for attachments penetrating fireproofing were reevaluated and clarified to ensure a consistent treatment of all cases.A compartment heat load analysis was performed for all rooms containing fireproofed structural steel.Safet Evaluation:

The fireproofing material specified was identical to that originally installed.

The changes which were made in this engineering package were determined not to adversely impact plant systems, structures, or components.

Furthermore, these changes did not alter the plant's overall design basis, which was bounded by existing design analyses.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 MODIFICATION 89-512 UNITS 3&4 TURN OVER DATE: 05/13/93 NUCLEAR OPERATIONS CHEMISTRY BREAK AREA AND CONTROL POINT~summar This engineering package provided a break area/control point for Nuclear Operations (NO)and Chemistry personnel, since they are not permitted to eat inside the Radiation Control Area (RCA).Locating a shelter outside the RCA, in the Turbine Building/yard area on the west side of the existing RCA fence, provided a convenient RCA entry/exit point for NO and Chemistry personnel as well as a break area.This shelter was constructed as a pre-fabricated non-combustible shelter.Additional fencing was provided to direct personnel from the access point to the break area/control point.This required the installation of a personnel contamination monitor and a hand frisker.Both pieces of equipment were relocated from the existing control point/guard shack located north of this new control point.Access to and from the Radiation Control Area (RCA)through the original control point/guard shack was no longer permitted, and access was transferred to this new control point.Safet Evaluation:

This shelter and related components do not provide any nuclear safety functions.

The shelter was located in a relatively clear area of the Turbine Building/yard area, and was designed to withstand the wind and roof loading requirements of the South Florida Building Code.The shelter and related components were reviewed for the seismic requirements.

These design basis requirements were implemented to preclude any potential interaction with future safety related systems, structures, and components.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 MODIFICATION 89-542 UNITS TURN OVER DATE 3&4 05/10/93 DRYING UPDATE-ISI UALITY GROUP CLASSIFICATIONS BOUNDARIES

~summa r This engineering package provided a basis for correcting design document deficiencies within the existing series of quality group classification

(" Code Boundary")drawings that were identified during previous QA audits.This engineering package corrected these deficiencies by adding the quality group classifications/

boundaries on selected POD T-E documents which best represent present day plant configurations and by providing bases for classifications/boundaries based on current regulatory commitments and other guidelines.

This plant change did not involve any physical plant configuration change.The accurate, up to date drawings were required in order to properly establish pressure test program requirements and appropriate procedures for In-service Inspection and Testing.This included ASME Section XI ISI code testing and support design requirements for repair and modifications to the piping systems.Safet Evaluation:

The Quality Group Classification bases evaluated in this engineering package were not considered an operability concern based on design equivalence.

They did not impact system operation or create any safety related concerns.They were drawing additions of Quality Group Classifications which were evaluated to be acceptable.

No new equipment or components were installed.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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

PLANT CHANGE MODIFICATION 90-193 Page 1 of 2 UNIT TURN OVER DATE 3 10/08/92 ADDIT1ON OF APPENDIX R BYPASS SWITCH FOR LCV 3 460~summa r This engineering package provided a keylocked bypass switch located on the main control board 3C03, which defeated the electrical interlock between Chemical Volume Control System (CVCS)valves LCV-3-460 and CV-3-200A, B, and C.The bypass switch would be used only in the event that a fire causes a hot short to spuriously open one of the CV-3-200A, B, or C valves and prevent closure of LCV-3-460.The addition of the bypass switch replaced the original requirement.

for pulling the control fuses associated with CV-3-200A, B, and C to defeat the circuit interlock.

These modifications ensured the availability of LCV-3-460 to perform its safe shutdown function for postulated fire scenarios causing spurious opening of CV-3-200A, B, and C.Credit is taken during certain Appendix R fire scenarios, including Alternate Shutdown, for LCV-3-460 to provide CVCS letdown isolation during safe shutdown.LCV-3-460 is a DC solenoid controlled valve which has circuit interlocks with downstream orifice isolation valves, CV-3-200A, B, and C.This interlock is intended to prevent potential damage to the regenerative heat.exchanger and relief valve RV-3-203 due to pressure transients in the line between LCV-3-460 and the CV-3-200 valves.CV-3-200A, B, and C are DC solenoid-controlled valves that close on loss of electrical power or loss of control air.Spurious opening of any one of the CV-3-200 valves due to a hot short would prevent closure of LCV-3-460, because of the electrical interlock between the valves.This condition was not a concern for Alternate Shutdown but was valid for other fire zones.This engineering package modification served to correct this potential issue.Safet Evaluation:

This modification enhanced the capability of the control room operator to defeat the interlock between LCV-3-460 and CV-3-200A, B, and C and mitigate the consequences of a fire in areas outside the alternate shutdown areas caused by a hot short.The operators 35 PLANT CHANGE MODIFICATION 90-193 Page 2 of 2 ADDZTZON OF APPENDZX R BYPASS 8'PITCH FOR LCV 3 460 ability was enhanced by providing a bypass switch, located adjacent to the existing control switch for LCV-3-460 to defeat the interlock.

This bypass switch provided the operator a quicker and more desirable method to mitigate the consequences of a fire and is also consistent with NRC guidance for actions required to achieve hot shutdown.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 PLANT CHANGE MODXFZCATXON 90-194 Page 1 of 2 UNIT 4 TURN OVER DATE: 04/19/93 ADDZTXON OF APPENDXX R BYPASS SNITCH FOR LCV-4 460~8ummam This engineering package provided a keylocked bypass switch located on the main control board 4C03, which defeated the electrical interlock between Chemical Volume Control System (CVCS)valves LCV-4-460 and CV-4-200A, B, and C.The bypass switch would be used only in the event that a fire causes a hot short to spuriously open one of the CV-4-200A, B, or C valves and prevent LCV-4-460 closure.The addition of the bypass switch replaced the original requirement for pulling the control fuses associated with CV-4-200A, B, and C to defeat the circuit interlock.

These modifications ensured the availability of LCV-4-460 to perform its safe shutdown function for postulated fire scenarios causing spurious opening of CV-4-200A, B, and C.Credit is taken during certain Appendix R fire scenarios, including Alternate Shutdown, for LCV-4-460 to provide CVCS letdown isolation during safe shutdown.LCV-4-460 is a DC solenoid controlled valve which has circuit interlocks with downstream orifice isolation valves, CV-4-200A, B, and C.This interlock is intended to prevent potential damage to the regenerative heat exchanger and relief valve RV-4-203 due to pressure transients in the line between LCV-4-460 and the CV-4-200 valves.CV-4-200A, B, and C are DC solenoid-controlled valves that close on loss of electrical power or loss of control air.Spurious opening of any one of the CV-4-200 valves due to a hot short would prevent closure of LCV-4-460, because of the electrical interlock between the valves.This condition was not a concern for Alternate Shutdown but was valid for other fire zones.This engineering package modification served to correct this potential issue.Safet Evaluation:

This modification enhanced the capability of the control room operator to defeat the interlock between LCV-4-460 and CV-4-200A, B, and C and mitigate the consequences of a fire in areas outside the alternate shutdown areas caused by a hot short.The operators 37

PLANT CHANGE MODIFICATION 90-194 Page 2 of 2 ADDITION OF APPENDIX R BYPASS SWITCH FOR LCV 4-460 ability was enhanced by providing a bypass switch, located adjacent to the existing control switch for LCV-4-460 to defeat the interlock.

This bypass switch provided the operator a quicker and more desirable method to mitigate the consequences of a fire and is also consistent with NRC guidance for actions required to achieve hot shutdown.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 lp 0 PLANT CHANGE MODIFICATION 90-239 UNIT~4 TURN OVER DATE: 05/10/93 C BUS SliFITCHGEAR CONTROL AND PROTECTION PORER ZSOLATZON FOR APPENDZX R~Summa This modification provided for the installation of a molded case circuit breaker at the C Bus switchgear to connect or disconnect the control and protection power for the switchgear.

This eliminated the present fuse pulling requirement and, thereby, reduced the operator burden associated with supplying power to the Standby Steam Generator Feedwater (SSGF)pumps.The C Bus switchgear provides power to the Standby Steam Generator Feedwater Pumps (SSGFPs).These pumps provide an alternate source of feedwater to the steam generators.

In the event of a postulated fire which could render the Auxiliary Feedwatr System (AFW)inoperable the SSGF pumps are utilized to provide feedwater.

During this condition credit is taken to power the C Bus switchgear from the Units 1 and 2 Cranking Diesels after tripping the C Bus switchgear breakers.The required safe shutdown breakers are then manually aligned.Safet Evaluation:

The modification provided'by this engineering package installed a molded case circuit breaker on the switchgear door.The use of a breaker reduced the number of actions required to connect or disconnect the C Bus switchgear control and protection power.In addition, since the breaker was flush mounted to the switchgear door, operator entry into the panel to pull fuses was eliminated.

Therefore, this modification enhanced the means by which the control and protection power was isolated.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

39

PLANT CHANGE MODIFICATION 90-240 UNIT TURN OVER DATE 3 11/17/92 RAD 3 6417 SAMPLE LINE END CABINET MODIFICATION

~mamma This engineering package eliminated the chillers and drain tanks located in the Unit 3 Steam Jet Air Ejector exhaust system.The chiller system had been installed to remove entrained water and water vapor from the exhaust flow prior entering radiation monitor RaD-3-6417.

The original equipment was incompatible with the exhaust constituents (a caustic mixture of air,'ater, and ammonial), which had resulted in a history of material deterioration, failure and excessive maintenance requirements.

The original chillers, drums, connecting piping and associated equipment were removed.The sample line to RaD-3-6417 was re-routed to take the sample from the existing (plugged)threaded connection on the air ejector exhaust gooseneck.

The sample line from the existing water separator to the monitor was heat traced and insulated to heat the sample prior to its passage through the monitor.This decreased the sample relative humidity and prevented condensation in the detectors, without the complications.

and maintenance requirements imposed by the existing chiller system.Safet Evaluation; The modification increased the reliability of the subject monitor, which decreased the likelihood of entering a Technical Specification Action statement.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

40

,4 0 PLANT CHANGE MODIFICATION 90-396 UNIT TURN OVER DATE 3 10/16/92 NIS RECORDER CHANNEL SELECTOR SWITCHES~81HOKR This engineering package involved the modification of the NIS recorder channel selector switches in the control room.This modification consisted of replacing the existing twelve position NIS recorder channel selector switches with eight position switches.In addition, wiring associated with these four unused switch positions located between the first terminal block in panel 3C01 and the selector switches was removed.During the detailed control room design review, the unused positions of the nuclear instrumentation system recorder selector switches were identified as a human engineering deficiency (HED No.TA-40).Florida Power and Light committed to resolve this HED by eliminating the unused switch positions and changing the escutcheon plates.Safet Evaluation:

The replacement eight position switches were equivalent in all respects to the existing twelve position switches.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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 90 397 UNIT~4 TURN OVER DATE: 05/03/93 NIS RECORDER CHANNEL SELECTOR SWITCHES.~SllRRR This engineering package involved the modification of the NIS recorder channel selector switches in the control room.This modification consisted of replacing the existing twelve position NIS recorder channel selector switches with eight position switches.In addition, wiring associated with these four unused switch positions located between the first terminal block in panel 4C01 and the selector switches was removed.During the detailed control room design review, the unused positions of the nuclear instrumentation system recorder selector switches were identified as a human engineering deficiency (HED No.TA-40).The nuclear instrumentation system recorder provides the control room operator with trending information.

This data is particularly valuable during reactor startup and other power transients.

If these selector switches were inadvertently placed in one of the unused positions the resulting display could confuse the control room operator.Florida Power and Light committed to resolve this HED by eliminating the unused switch positions and changing the escutcheon plates.Safet Evaluation:

The replacement eight position switches were equivalent in all respects to the existing twelve position switches.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

42 PLANT CHANGE MODIFICATION 90 445 UNIT TURN OVER DATE 3 11/16/93 DR1LLING OF VALVE WEDGE FOR MOV-3 872~sammar This engineering modification provided for the drilling of a small pressure relieving hole in the valve wedge on the Reactor Coolant System (RCS)side of MOV-3-872.

MOV-3-872 is a component of the Alternate Low Head Safety Injection flowpath and is classified as safety related.As documented in INPO SOER 84-7, system pressure in the valve bonnet area may become trapped causing a high differential pressure across the valve disc/wedge and resultant binding during valve opening.These INPO reported failures have prevented safety related systems from functioning when called upon to operate.A subsequent engineering analysis determined that MOV-3-872 is not affected, however, as a long-term precautionary measure it was recommended that MOV-3-872 be modified to prevent the potential for pressure locking.This modification eliminated the potential for such binding.Safet Evaluation:

This modification did not affect the function of MOV-3-872 nor the operation of any plant systems.INPO SOER 84-7 documented the disc drilling modification as an acceptable solution to the potential for pressure locking and Velan concurred with drilling location and size of the hole.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

43 PLANT CHANGE MODIFICATION 90 446 UNITS 3&4 TURN OVER DATE: 09/23/92 WATER TREATMENT PLANT ZN-LINE MONITORS~summa This engineering package provided for the modifications of the Nuclear Chemistry Building and Water Treatment Plant to address INPO Finding CY.3-2.The modifications consisted of routing tubing from the cation, anion and mixed bed demineralizers and final effluent station to the Nuclear Chemistry Building, and connecting service water and drains to the Nuclear Chemistry Building lab sink.These modifications provided a central location for water chemistry analysis of the demineralized water produced by the Water Treatment Plant.Safet Evaluation:

The modifications performed by this engineering package involved the Water Treatment Plant System.There were no safety related systems affected by the implementation of this engineering package.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

44 PLANT CHANGE MODIFICATION 90-449 UNITS TURN OVER DATE 3&4 08/13/92 CCW AREA PIPE TRENCH FLOODRALLS

~summa This engineering package provided engineering and design required to install new reinforced concrete floodwalls, and repair existing floodwalls in pipe trenches located on the east side of the Auxiliary Building in the Component Cooling and Safety Injection areas.Technical Issue No.4 under the FPL Systematic Design Investigation (SDI)Program identified four pipe trenches, located on the east side of the Auxiliary Building in the Component Cooling and Safety Injection areas, as potential points of flood water intrusion into the flood protected area in the event of a hurricane surge tide.Flood water intrusion into the plant could adversely affect equipment or components important to safety.The new floodwalls were installed in the Unit 3 and 4 component cooling pipe trenches, and the existing floodwalls in the Unit 3 and 4 safety injection pipe trenches had gaps sealed.Flexible pressure boots were installed at large bore pipe penetrations in the component cooling water pipe trenches to provide a barrier against flood water intrusion while allowing for the design pipe movement.Safet Evaluation:

The effects of this modification on the external flood protection system were reviewed and no adverse effects will result from this implemented modification.

The ability of the external flood protection system to perform its design function was'enhanced by the installation and repair of floodwalls.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

45 PLANT CHANGE MODIFICATION 91-128 UNIT TURN OVER DATE 3 10/20/92 480V UNDERVOLTAGE PROTECTION SCHEME MODIFICATION

~summa This engineering package modified the 480V load center non-safety injection degraded voltage schemes.These modifications were required to improve the repeatability of the existing degraded voltage relays to their specified setpoints.

The original logic did not provide detection of auxiliary relay coil failures, and did not allow the circuit to be placed in the trip mode under the conditions of such an event.Redesign of the original degraded voltage scheme mitigated these conditions.

This modification also installed a bypass switch which will allow one channel of the degraded voltae scheme to be placed in the trip mode when one or both of the relays of that channel are removed for relay testing or calibration.

To accommodate the modifications provided by this engineering package, provisions for a minor change to the Technical Specifications were instituted.

This change deleted the specific reference to the inverse time relay.A HOLD POINT was placed on this engineering package to restrict any changes to the plant configuration as described in the Technical Specification until NRC approval was received.Safet Evaluation:

This modification was evaluated under the requirements of 10 CFR 50.59 and did not constitute an unreviewed safety question.This activity does not change the operation, function or design bases.of any structure, system or component important to safety as described in the SAR.In particular, the undervoltage protection scheme still used relays to detect a degraded voltage condition and to actuate sequencer trip using a two-out-of-two logic.The Technical Specification requirements applicable to this modification were not affected.However, the method of satisfying the Technical Specification requirements was affected.While the setpoint values are unchanged, the means of satisfying these setpoint requirements was accomplished by a new definite time delay undervoltage relay.The analysis performed to support this'ircuit modification confirmed that the new relays, at the existing setpoints, provided the necessary degraded voltage protection.

46 PLANT CHANGE MODIFICATION 91-130 UNIT TURN OVER DATE 3 10/21/92 PROCESS RADIATION MONITORING SYSTEM R-3 11 AND R 3-12 REPLACEMENT

~summar This engineering package was issued to replace containment particulate and noble gas radiation monitors R-3-11 (particulate) and R-3-12 (gaseous)and associated displays and controls in control room panel 3QR66 with new ones, which were expected to be more reliable and easier to maintain.R-3-11 and R-3-12 provide a means for monitoring the Unit 3 containment atmosphere for radioactivity released from normal operation, anticipated transients, and accident conditions.

These instruments are part of the Engineered Safety Features Instrumentation.

The new equipment performs similar functions to the original equipment with the exception of the capability to monitor the common plant vent;Since R-14 and RaD-6304 monitored the common plant vent, the need for R-3-11 and R-3-12 to monitor the common plant vent was no longer necessary and was deleted.Safet Evaluation:

After the monitor replacements, monitors R-3-11 and R-3-12 continue to perform their safety functions, whereby the high radiation level for the channel initiates closure of the containment purge supply and exhaust duct valves and instrument air bleed valves, and initiates control room ventilation isolation, as described in the Updated FSAR.The provisions for monitoring the plant vent using R-3-11 and R-3-12 was no longer necessary, since this function continued to be accomplished by monitors R-14 and RaD-6304.Wide range monitor RaD-6304 was used to satisfy the monitoring requirements of Regulatory Guide 1.97.In addition, monitors R-3-11 and R-3-12 provided early detection capability, since containment atmosphere is monitored directly.These functions fulfilled the leakage detection system requirements and the Engineered Safety Features Instrumentation requirements as described in the Updated FSAR.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

47 PLANT CHANGE MODZRICATZON 91-133 UNIT TURN OVER DATE 3 07/31/92 REPLACEMENT Op 480 VOLT MOTOR CONTROL CENTER 3E~summa This Engineering Package provided the engineering and design necessary to replace MCC 3E with a new Motor Control Center.Rust and corrosion of the internal structure of the non-nuclear safety related 480 Volt Motor Control Center (MCC)3E, located outdoors at the intake structure, had resulted in structural degradation extensive enough to require replacement.

The cause of this degradation was related to the utilization of noncorrosion-resistant materials in the original design.Also, the MCC was obsolete and obtaining replacement parts was becoming difficult.

The new MCC was designed and constructed as a standard MCC installed in a stainless steel (type 304L)NEMA 4X enclosure, providing increased corrosion protection.

Moisture entry from a manhole below was blocked by a stainless steel bottom on the enclosure and cable entry openings were sealed after installation of the cables.Safet Evaluation:

Motor Control Center 3E does not supply power to or control any nuclear safety related plant equipment and is normally powered, via MCC 3F, from Load Center 3F, which in turn is powered from non-safety related 4160 VAC Switchgear 3C.None of the equipment in the vicinity of MCC 3E was Nuclear Safety Related.MCC 3E cannot be powered from the safety related emergency diesel generators, and is not powered from the vital 125 VDC system.The modifications in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

48 PLANT CHANGE MODIFICATION 91-166 UNIT TURN OVER DATE 3 11/29/92 REPLACEMENT OF SEAL TABLE FITTINGS AND THIMBLE TUBE LENGTHENING

~summa r This engineering modification provided for the replacement of the existing seal table fittings with new fittings that contained an integral low pressure seal for refueling.

The original fittings were frequently a source of primary leakage during plant startup and operation.

The original low pressure refueling seals were difficult to assemble and often leaked.The new design eliminated the original guide tube ferrule which was frequently the location of leakage when the seal was configured for plant operation.

The replacement design was welded to the guide tube and had a tapered machined sealing surface that replaced the ferrule.The new refueling seal utilized the same sealing technique as the original seal, i.e., compression of an elastomer to form the seal.The difference in the refueling seal was that the new fitting had the seal permanently installed and used an internal nut for compression.

This modification also provided for lengthening of thimble tubes to compensate for reduced core inserted thimble end elevations.

This condition was associated with new fitting stackup dimensions and previous thimble tube shortening activities.

Safet Evaluation:

The new seal table fittings, thimble tube extensions and replacement guide tubes were similar to the existing hardware and incorporate some improvements.

The capping abandoned thimble tubes and provided an improved and more reliable seal design than the original isolation valve, in the event of a thimble tube leak inside the RCS pressure boundary.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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

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

49 PMNT CHANGE MODIFICATION 91-198 UNIT~3 TURN OVER DATE: 11/14/92 REPAIR AND MODIFICATION OF THE UNIT 3 INTAKE STRUCTURE~summar This engineering package restored, as required, the concrete slabs supporting Unit 3 ICW 3A and 3B Pumps and the Screen Wash Pumps to a condition which met the original design bases, and ensured acceptable long-term performance.

This was accomplished for the ICW 3A and Screen Wash Bays by removing deteriorated concrete and reinforcing steel, protecting uncovered reinforcing steel from future corrosion, and replacing concrete with material of equal or greater strength.Repairs were implemented using Nonconformance Reports, which were then evaluated and dispositioned by Engineering providing appropriate repairs for each bay.Safet Evaluation:

Upon completion of the modifications, the structural integrity of the Intake Structure slabs were restored to withstand all applicable loads in accordance with the requirements for Class I structures identified in the Updated FSAR, including operating loads of the pumps and associated components, thereby meeting the original design intent of the slab.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

50 0

PLANT CHANGE MODXFXCATXON 92-004 UNITS TURN OVER DATE 3&4 05/25/93 UPGRADXNG PLANT PAGE AUDXBXLXTY

~summa r This engineering package provided the design to supplement the plant page alarms in the high noise areas by replacing 13 existing blue lights with high intensity strobe lights and also added 31 high intensity strobe lights at various locations.

The Turkey Point Public Address System provides normal plant paging capabilities and is also utilized to broadcast site evacuation and containment evacuation alarms throughout the plant.The new strobe lights would be activated during any emergency alarm.The new strobe lights were powered from 120 VAC paging power supply.This improved awareness to the emergency alarm in all plant areas.The original separate fire alarm system wiring was abandoned and fire horns and fire alarm control relays were removed.The new tone generator would broadcast the fire alarm over the existing public address system speakers.Safet Evaluation:, No credit is taken for the public address system to support operator actions to accomplish safe shutdown or accident mitigation, to prevent uncontrolled release of radioactivity or to perform a fire protection function.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

51 PLANT CHANGE MODIFICATION 92-033 UNIT TURN OVER DATE 3 10/21/92 EMERGENCY BUS LOAD SE UENCER MODIFICATIONS

~summa This engineering package modified the Unit 3 programmable emergency bus load sequencers to eliminate the root cause of the failure that resulted in Sequencer 4A aborting an Auto Test and to address the results of the Failure Mode and Effects Analysis identified in Engineering Report No.JPN-PTN-SEIS-92-010.

This engineering package also upgraded the sequencers by installing a new output module for diagnostic purposes.The intent of the diagnostics was to create an error message code and to provide additional information to facilitate troubleshooting.

Other modifications included in the scope of this engineering package consisted of programming modifications to eliminate a nuisance alarm and the delay (eleven cycles)of the signal from the Auxiliary Transformer breaker position and rewiring of some of the blocking relays to ensure that failure of a relay to de-energize would be displayed in the front panel.Safet Evaluation:

The sequencer modifications were tested in both the simulator and in the plant sequencers and demonstrated that the sequencer safety functions had not been affected and confirmed the proper interactions between the modified sequencers and plant equipment.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

52 PLANT CHANGE MODIPICATION 92-034 UNIT TURN OVER DATE 4 05/03/92 EMERGENCY BUS LOAD SE UENCER MODIRICATIONS

~mamma r This engineering package modified the Unit 4 programmable emergency bus load sequencers to eliminate the root cause of the failure that resulted in Sequencer 4A aborting an Auto Test and to address the results of the Failure Mode and Effects Analysis identified in Engineering Report No.ZPN-PTN-SEIS-92-010.

This engineering package also upgraded the sequencers by installing a new output module for diagnostic purposes.The intent of the diagnostics was to create an error message code and to provide additional information to facilitate troubleshooting.

Other modifications included in the scope of this engineering package consisted of programming modifications to eliminate a nuisance alarm and the delay (eleven cycles)of the signal from the Auxiliary Transformer breaker position and rewiring of some of the blocking relays to ensure that failure of a relay to de-energize would be displayed in the front panel.Safet Evaluation:

The sequencer modifications were tested in both the simulator and in the plant sequencers and demonstrated that the sequencer safety functions had not been affected and confirmed the proper interactions between the modified sequencers and plant equipment.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

53 PLANT CHANGE MODZF1CATION 92 040 UNIT TURN OVER DATE 3 12/01/92 ADDITION OF REVERSE POWER RELAY AND MAZN GENERATOR PROTECTION MODIFICATIONS

~summa This engineering package modified and upgraded the Turkey Point Unit 3 main generator protection schemes.These modifications realigned the existing main generator protection schemes to mitigate the effects of a partial loss of vital DC power on their protection capability.

These modifications included the realignment of the vital DC control power supplying these schemes and addition of a reverse power relay.In addition, this modification upgraded the existing generator protection by regrouping existing main generator primary and backup protective functions, adding an out-of-step relay and main generator output circuit breaker emergency trip control switch.These design upgrades provided additional backup capability and enhancements to the generator protection schemes.The INPO Significant Operating'xperience Report (SOER)81-15 made specific recommendations regarding the ability of plants to manage and recover from a loss of a vital DC bus.Implementation of these modifications satisfied INOP SOER 81-15 recommendation 1C.Safet Evaluation:

These modifications to the main generator protection schemes enhanced and upgraded existing non-safety related main generator protection system, which was not required for safe shutdown during a design basis accident.The generator protection system remained functionally the same as a result of this modification, and the effects of a failure of any relay remained unchanged from the original design.The safety related function of the 125V vital DC Buses 3D01 and 4323 which feed these circuits were unaffected by this modification, since any power feed realignment performed would be downstream from the 125V vital DC bus isolation circuit breakers.The function of the auxiliary relays, which interface with the safety related diesel generator, remained unchanged and unaffected by this modification.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

54 0

PLANT CHANGE MODZFZCATZON 92-054 UNIT~4 TURN OVER DATE: 05/01/93 480V UNDERVOLTAGE PROTECTZON 8CHEME MODZFZCATZON

~summa This engineering package modified the 480V load center non-safety injection degraded voltage schemes.These modifications were required to improve the repeatability of the existing degraded voltage relays to their specified setpoints.

The original logic did not provide detection of auxiliary relay coil failures, and did not allow the circuit to be placed in the trip mode under the conditions of such an event.Redesign of the original degraded voltage scheme mitigated these conditions.

This modification also installed a bypass switch which would allow one channel of the degraded voltage scheme to be placed in the trip mode when one or both of the relays of that channel are removed for relay testing or calibration.

To accommodate the modifications provided by this engineering package provisions for a minor change to the Technical Specification were instituted.

The Technical Specification change deleted the specific reference to the inverse time relay.This change was approved by the NRC as Amendments No.152 and No.147 for Units 3 and 4, respectively.

8afet Evaluation:

This activity does not change the operation, function or design bases of any structure, system or component important to safety as described in the Updated FSAR.In particular, the undervoltage protection scheme still used relays to detect a degraded voltage condition and to actuate sequencer trip using a two-out-of-two logic.This modification was evaluated under the requirements of 10 CFR 50.59 and did not constitute an unreviewed safety question.The Technical Specification requirements applicable to this modification are not affected.However, the method of satisfying the Technical Specification requirements was affected.While the setpoint values were unchanged, the means of satisfying these setpoint requirements was accomplished by the new definite time delay undervoltage relay.The analysis performed to support this circuit modification confirmed that the new relays, at the original setpoints, provide the necessary degraded voltage protection.

55 PLANT CHANGE MODIFICATION 92-057 UNIT TURN OVER DATE 3 11/09/92 HHSZ THERMAL RELIEF VALVE~summar This modification to Unit 3 Containment Penetration No.18 piping consisted of adding a new relief valve for the overpressure protection of this piping.The original design of Containment Penetration No.18 was to relieve thermal and valve leak-by-overpressure conditions through the use of cross-tied relief valve RV-3-859, located in adjacent Penetration No.17.This relief scheme was eliminated by isolation of the manual cross-tie valve (3-849A)to alleviate operational problems experienced during routine SI accumulator filling operations.

Safet Evaluation:

This modification was for the installation of a relief valve in the portion of the SI system that was originally overpressure protected by relief valve RV-3-859.The new relief valve was equivalent to the relief valve RV-3-'859, and performed the same function as this original valve.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

56 PLANT CHANGE MODIFICATION 92-058 UNIT, 4 TURN OVER DATE: 05/04/93 PROCESS RADIATION MONITORING SYSTEM R 4Ž1I.AND R-4-12 REPLACEMENT

~Summa This engineering package was issued to replace containment particulate and noble gas radiation monitors R-4-11 (particulate) and R-4-12 (gaseous)and associated displays and controls in control room panel 4QR66 with new ones, which were expected to be more reliable and easier to maintain.R-4-11 and R-4-12 provide a means for monitoring the Unit 4 containment atmosphere for radioactivity released from normal operation, from anticipated transients and from accident conditions.

These instruments are part of the Engineering Safety Features Instrumentation.

The new equipment performed similar functions to the original equipment, with the exception of the capability to monitor the common plant vent.Since R-14 and RaD-6304 originally monitored the common plant vent, the need for R-4-11 and R-4-12 to monitor the common plant vent was no longer necessary and this function was deleted.Safet Evaluation:

After the monitor replacements, monitors R-4-11 and R-4-12 continued to perform their safety functions, whereby the high radiation level for the channel initiates closure of the containment purge supply and exhaust duct valves and instrument air bleed valves, and initiates control room ventilation isolation as described in the Updated FSAR.The provision for monitoring the common plant vent using R-4-11 and R-4-12 was no longer necessary, since this function continued to be accomplished by monitors R-14 and RaD-6304.Wide range monitor RaD-6304 was used to satisfy the monitoring requirements of Regulatory Guide 1.97.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

57

PLANT CHANGE MODIFICATION 92-059 UNITS 3 6 4 TURN OVER DATE: 11/16/92 CONTROL ROOM AZR CONDITIONING AND VENTILATZON SYSTEM CONTROL MODIFICATION

~summar This engineering package modified the control room Air Conditioning and Ventilation System to allow for the independent operation of the three control room air conditioning trains (i.e., air conditioner and air handler unit).This was accomplished by removing the existing common thermostat, controller and control switches, and providing common independent thermostats, one for each air conditioner units (i.e., compressor/condenser).

Also, the air handler motor starters were removed and replaced with fuses for circuit and motor protection.

The air handlers may run continuously and their operation only depends on their respective A/C train's power source.In addition, the Firestat sensors were disconnected.

These modifications eliminated the potential single failure concerns and associated temporary corrective actions for the control room air conditioning system addressed in the Justification for Continued Operation, as identified in JPE-L-86-113.Safet Evaluation:

This new design provided electrically independent circuits and components for each air conditioning train, so that, any postulated single failure of a circuit or component would only disable its associated air conditioning train.This modification eliminated existing single failure concerns and did not create any new failure modes that could impact nuclear safety.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

58 PLANT CHANGE MODIFICATION 92-063 UNIT~3 TURN OVER DATE: 11/11/92 REACTOR COOLANT PUMP 3B MOTOR REFURB18HMENT UPGRADE~Summa r As part of the on-going program to improve the reliability and performance of all Reactor Coolant Pumps (RCP)at Turkey Point, this engineering package documented the upgrade of the 3B Reactor Coolant Pump Motor.The originally installed motor was replaced with a spare motor which was refurbished at the Westinghouse Electro-Mechanical Division facility.This standard factory refurbishment consisted of inspection and maintenance activities performed to the existing design specifications.

In addition, modifications were conducted, concurrent with the refurbishment, to ensure consistency with the latest RCP technology and to realize additional reliability and availability.

Upon completion of the modifications, the motor was assembled, balanced and tested and shipped back to PTN.Safet Evaluation:

The RCP motor does not perform any safety related function, with the exception of providing sufficient inertia (through its flywheel)to ensure sufficient coastdown of the Reactor Coolant Pump after an RCP trip.The RCP motor modifications did not affect the coastdown characteristics of the motor.The modification in this Engineering Package did not have any adverse, effect on plant safety or plant operations.

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.

59 0

PLANT CHANGE MODIFICATION 92 073 UNIT TURN OVER DATE 4 05/25/93 ADDITION OF REVERSE POWER RELAY AND MAIN GENERATOR PROTECTION MODIFICATIONS

~81HSltIR1'his engineering package modified and upgraded the Turkey Point Unit 4 main generator protection schemes.These modifications realigned the existing main generator protection schemes to mitigate the effects of a partial loss of vital DC power on their protection capability.

These modifications included the realignment of the vital DC control power supplying these schemes and added a reverse power relay.In addition, this modification upgraded the existing generator protection by regrouping existing main generator primary and backup protective functions, added new relays for out-of-step and 1004 ground protection and an emergency control switch to trip the main generator output circuit breakers.Also included as part of this upgrade was the addition of new protection schemes for inadvertent connection, string bus differential and automatic synchronizing.

The Institute of Nuclear Power Operations (INPO)Significant Operating Experience Report (SOER)81-15 made specific recommendations regarding an operating plant's ability to manage and recover from a loss of a vital DC bus.Implementation of.these modifications satisfied INPO SOER 81-15 recommendation 1c.Safet Evaluation:

These modifications to the main generator protection schemes enhanced and upgraded the original non-safety related main generator protection system.No credit is taken for these protection features to accomplish safe shutdown during a design basis accident.The generator protection system remained functionally the same as a result of this modification and the effect of failure of any relay remained unchanged from the original design.Also, the function and performance capability of the'uxiliary relays which interface with the safety related diesel generator remained unchanged and unaffected by this modification.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

60 PLANT CHANGE MODIFICATION 92-074 UNIT~3 TURN OVER DATE: 11/08/92 CORE EXIT THERMOCOUPLE SEAL UPGRADE~summa In order to reduce the potential for primary boundary leakage around the core exit thermocouple (CET)nozzles, this engineering package was implemented to install a new, single piece head port adapter on each CET nozzle, thereby, eliminating the lower Conoseal joint and replacing the upper Conoseal joint with a Grafoil graphite seal.The upper Grafoil seal cartridge was softer and more forgiving than the Conoseal and wil be replaced each time the seal is reassembled.

This seal also allowed for a one-joint disassembly for head removal at each outage, which resulted in significant time and radiation exposure savings.The four original core exit thermocouple nozzles on the Unit 3 reactor vessel closure head each have two primary pressure boundary"Conoseal" metal seals that must be disassembled at each refueling outage.The Conoseal installation techniques and surface finish on sealing faces were extremely critical in preventing degradation of the sealing capability for preventing RCS leakage.In recent years, Turkey Point has had several leaks at the Conoseal upon returning to service after an outage.Repair of these leaks has required significant unplanned outage time and lost power generation.

The problems with the Conoseal design originally installed were attributed to the difficulty in assembling the seal and degradation of the sealing surfaces which had occurred during the many times they were disassembled and reassembled.

Safet Evaluation:

This modification replaced an existing component with one providing the same function, containing the same basic pressure retaining components, and designed to meet or exceed the original ASME Boiler and Pressure Vessel code requirements.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

61 PLANT CHANGE MODIFICATION 92 075 UNIT TURN OVER DATE 4 05/12/93 CORE EXIT THERMOCOUPLE SEAL UPGRADE~summa In order to reduce the potential for primary boundary leakage around the core exit thermocouple (CET)nozzles, this engineering package was implemented to install a new, single piece head port adapter on each CET nozzle, thereby, eliminating the lower Conoseal joint and replacing the upper Conoseal joint with a Grafoil graphite seal.The upper Grafoil seal cartridge was softer and more forgiving than the Conoseal and will be replaced each time the seal is reassembled.

This seal allowed for a one-joint disassembly for head removal at each outage, which resulted in significant time and radiation exposure savings.The f'our original core exit thermocouple nozzles on the Unit 3 reactor vessel closure head each have two primary pressure boundary"Conoseal" metal seals that must be disassembled at each refueling outage.The Conoseal installation techniques and surface finish on sealing faces were extremely critical in preventing degradation of the sealing capability for preventing RCS leakage.In recent years, Turkey Point has had.several leaks at the Conoseal upon returning to service after an outage.Repair of these leaks has required significant unplanned outage time and lost power generation.

The problems with the Conoseal design originally installed were attributed to the difficulty in assembling the seal and degradation of the sealing surfaces which had occurred during the many times they were disassembled and reassembled.

(Safet Evaluation:

This modification replaced an existing component with one providing the same function, containing the same basic pressure retaining components, and designed to meet or exceed the original ASME Boiler and Pressure Vessel code requirements.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

62 PLANT CHANGE MODIFICATION 92-079 UNIT~4 TURN OVER DATE: 04/30/93 REPAIR AND MODIFICATION OR THE UNIT 4 INTAKE STRUCTURE~8UIIUII&This engineering package restored the concrete slabs supporting Unit 4 ICW 4A and 4C Pumps and the Screen Wash Pumps to a condition which met the original design bases, and ensured acceptable long-term performance.

This was accomplished for the ICW 4A and Screen Wash Bays by removing deteriorated concrete and reinforcing steel, protecting uncovered reinforcing steel from future corrosion, and replacing concrete with material of equal or greater strength.Repairs were implemented using Nonconformance Reports, which were then evaluated and dispositioned by Engineering providing appropriate repairs for each bay.Safet Evaluation:

Upon completion of the modifications, the structural integrity of the Intake Structure slabs were restored to withstand all applicable loads in accordance with the requirements for Class I structures of Updated FSAR, Appendix 5A, including operating loads of the pumps and associated components, thereby meeting the original design intent of the slab.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

63 PLANT CHANGE MODIFICATZON 92-097 UNIT TURN OVER DATE 4 05/12/93 a ALTERNATE SAFETY INJECTION THERMAL RELZEP VALVE MODIPICATION

~summa This modification consisted of adding a new relief valve in the piping near Unit 4 Penetration No.18 to provide overpressure protection for this piping.This new relief valve was installed with the inlet line connected directly to the 2-inch alternate safety injection piping and the discharge connected to the existing 3-inch discharge line downstream of relief vale RV-4-382.This new relief valve provided the same overpressure protection function as did RV-4-859, which was part of the original design the for Safety Injection System (SIS).The relief function for thermal and valve leak-by type overpressure conditions at Penetration No.18 was originally performed by relief valve, RV-4-859, located on adjacent Penetration No.17.This relief scheme was eliminated through isolation of the manual cross-tie valve, 4-849A, to alleviate relief valve lifting problems experienced during routine SI accumulator filling operations.

This engineering package also provided recommendations for Operating Procedures which were revised to eliminate or minimize the Safety Injection System hydraulic transients that may occur during the safety injection accumulator filling operation.

Safet Evaluation:

This modification did not change the operation, function or design bases of any structure, system or component important to safety as described in the Updated FSAR.The components installed under this modification met or exceeded the requirements on the system where they were installed.

Also, the affected portion of the SI system was returned to a configuration equivalent to its original design;therefore, no new flow path was created.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

64

PLANT CHANGE MODIFICATION 92 102 UNITS TURN OVER DATE 3&4 03/15/93 REPLACEMENT OF RAN WATER STORAGE TANK I T63A~Summa This engineering package provided the design documentation necessary to replace Raw Water Storage Tank I and to provide appropriate repairs to the damaged foundation.

On August 24, 1992, Turkey Point Nuclear site sustained wind damage from Hurricane Andrew.During this event, Raw Water Storage Tank I (T63A), which supplied water to the plant fire protection system and provided potable and service water to the plant, was demolished when the nearby elevated water tank collapsed on top of it.'a The Raw Water Storage Tank I does not perform a safety related function.It is, however, part of the plant fire protection system.The suction nozzle location for the new raw water and service water system remains unchanged on the replacement tank, and thus, the fire protection water reserve capacity is unaffected by the tank replacement.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

65

PLANT CHANGE MODIFICATION 92 108 UNITS TURN OVER DATE 3&4 03/19/93 REPLACEMENT OF RAW WATER AND SERVICE'WATER SYSTEM DAMAGED BY HURRICANE ANDREW~mamma r This engineering package provided the design and evaluation required for replacement of the damaged portions of the service water and raw water system.On August 24, 1992, Hurricane Andrew passed over the Turkey Point Power Plant.An elevated water tank located between the fossil plant intake and the nuclear plant intake collapsed due to the storm and damaged equipment located beneath it, including the Raw Water Tank I, portions of the Fire Protection system, raw water booster (service water)pumps, raw water pumps, associated piping, valves, instrumentation, and power supply.The elevated storage tank was eliminated from the system design;however, a diesel engine driven service water pump was added to the system to provide an alternate water source in the event of a loss of electric power to the service water pumps.Safet Evaluation:

The service water and raw water systems do not provide any safety related functions, however, connections to the Fire Protection system were restored in accordance with the original system design.The changes provided in this PC/M do not alter the functions of the service water or raw water systems, and there was no change to the overall operation of the plant.The modification in this Engineering Package did not have any adverse effect on plant safety, or plant operations.

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.

66 PLANT CHANGE MODIFICATION 92-110 UNITS 3&4 TURN OVER DATE 12/29/92 INSTALLATION OF A DUCT BANK FROM MH 610 TO MH 324~summa This engineering package provided design details for the construction of a duct bank which formed part of the raceway for new non-safety related cables routed from Load Center 3F and 3G to service water pumps P235A, P235B and P235C (the new cables will be installed under PC/M 92-108).On August 24, 1992, Turkey Point Nuclear site sustained wind damage from a Category A hurricane, designated Hurricane Andrew.During this event, service water pumps P235A, P235B and P235C, and the associated power supply were severely damaged, when the nearby elevated water tank collapsed.

The power for the non-safety related service water was originally supplied from fossil Units 1 and 2.However, due to the security separation of the fossil and nuclear units, power for the pumps was supplied from the Load Center 3F and 3G.Safet Evaluation:

The existing manholes MH 324 and MH 610, the duct bank, and the new non-safety related cables pulled through the duct bank to supply power and control functions to service water pumps P235A, P235B and P235C did perform any safety related functions and had no potential for interaction with safety related equipment/systems.

The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

'67 0

PLANT CHANGE MODZFZCATXON 92 124 UNITS TURN OVER DATE 3 6 4 03/30/93 OFFSXTE RADXO COMMUNXCATXONS PROSPECT~Summa In order to provide a more reliable, permanent installation for the Offsite Radio Communications System, this engineering package provided the design and evaluation for a new permanent system that will remain functional before, during, and after an event similar to Hurricane Andrew.On August 24, 1992 Hurricane Andrew passed over the Turkey Point Power Plant.The original Offsite Radio Communication System design did not remain functional due to the hurricane winds experienced during Hurricane Andrew.The original Offsite Radio Communication System design was not provided with a reliable power source in the event of loss of offsite power (LOOP), andits design did not provide for redundancy or diversity to ensure offsite communications would remain available.

Based on the results of the tests, performed throughout the PTN Site, the five radio systems and locations provide reliable and acceptable offsite communications.

Three systems are within the project scope of this modification.

This design provided diverse and wireless path of communications to local and remote FPL facilities, the emergency operations centers of local counties, and the State and Federal agencies during times of off-normal and emergency events.These systems will remain functional during any foreseeable natural events that are within the design envelope of the plant.Safet Evaluation:

The function of the new Offsite Radio Communications System is similar to the original system, to provide offsite communications between the Plant and various external and internal organizations in the event of an emergency and loss of normal communications.

The new design is intended to provide a more reliable Offsite Radio Communications System capable of withstanding an event similar to Hurricane Andrew.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

68 PLANT CHANGE MODIFICATION 92-163 UNIT a 4 TURN OVER DATE: 05/22/93 REPLACEMENT OF SEAL TABLE FITTINGS AND THIMBLE TUBE LENGTHENING

~summa This engineering modification provided for the replacement of the existing seal table fittings with new fittings that contained an integral low pressure seal for refueling.

The original fittings" were frequently a source of primary leakage during plant startup and operation.

The original low pressure refueling seals were difficult to assemble and often leaked.The new design eliminated the existing guide tube ferrule which was frequently the location of leakage when the seal is configured for plant operation.

The replacement design was welded to the guide tube and had a tapered machined sealing surface that replaced the ferrule.The new refueling seal utilized the same sealing technique as the original seal, i.e., compression of an elastomer to form the seal.The difference in the refueling seal was that the new fitting had the seal permanently installed and used an internal nut for compression.

This modification also provided for lengthening of thimble tubes to compensate for reduced core inserted thimble end elevations.

This condition was associated with new fitting stackup dimensions and previous thimble tube shortening activities.

Safet Evaluation:

The new seal table fittings, thimble tube extensions and replacement guide tubes were similar to the existing hardware and incorporated some improvements.

The method of capping abandoned thimble tubes provided an improved and more reliable seal design than the original isolation valve in the event of a thimble tube leak inside the RCS pressure boundary.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

69 PLANT CHANGE MODIFICATION 92-166 I UNIT TURN OVER DATE 4 04/15/93 NIS SOURCE RANGE DETECTOR REPLACEMENT

~summar The modifications provided by this engineering package cover the replacement of the original BF3 detector model WL-23706 with the new BF3 model NY-10032 in both source range channels for Unit 4.The replacement of the existing BF3 detector model WL-23706 with the new BF3 improved model NY-10032 was required because of the high number of failures experienced with the existing detectors, mainly during refueling outages, and the limited service life of the existing detectors.

The replacement detector was an integral cable proportional counter assembly with similar dimensions and parameters as the original detector.The new detector with titanium housing was corrosion resistant and is expected to increase the service life of the detector and reduce the number of detector failures.Safet Evaluation:

The replacement of the detectors with an improved model did not change the function of the NIS source range channels.Therefore, the system will perform its safety functions as originally designed.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

70 PLANT CHANGE MODIFICATION 92-181 UNIT TURN OVER DATE 4 04/30/93 ELIMINATZON OF TURBZNE RUNBACK ON DROPPED ROD~mamma This engineering package provided the necessary design documentation to remove the turbine runback selector switch HS-4-6686 and associated relays from the Control Room Panel 4C02 to eliminate the activation of a turbine runback on a dropped rod event.The automatic turbine runback feature of Units 3 and 4 is designed to provide protective action in the event of a dropped RCCA or dropped bank.Detection of a dropped RCCA or bank occurs by either a rod-on-bottom signal or by a chnage in neutron flux as seen by the NIS excore power range detectors.

The design of the automatic turbine runback on a dropped rod was prone to spurious runbacks (i.e., runbacks not caused by an RCCA drop), because there was no coincidence logic used in the initiation of the runback.Thus, a single failure of an electrical component could cause a turbine runback when it was not needed.Due to the fact that the majority of the spurious runbacks had resulted from failures in the flux rate input to the runback logic, this input was deleted during normal operation.

The NIS switch position was only used for short time intervals while performing periodic maintenance or tests.This modification to the Turbine Runback System was analyzed in Appendix 14C of the updated FSAR.This evaluation concluded that deletion of the flux rate portion of the Turbine Runback System is acceptable.

The reactor can be maintained in automatic rod control, since auto rod withdrawal had been previously eliminated.

Bafet Evaluation:

The RCCA Drop analysis in the FSAR is currently analyzed with the protective action of turbine runback.The Dropped RCCA transient assuming no turbine runback was analyzed by Westinghouse using a detailed digital simulation of the Turkey Point Plant.The results of the analysis confirmed that the departure from nucleate boiling (DNBR)remains above the limiting value for both standard and optimized fuel types.Thus, it was concluded that eliminating turbine runback following a dropped rod event would not have an adverse impact on plant safety.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

71

PLANT CHANGE MODIFICATION 93-009 UNIT~4 TURN OVER DATE: 05/07/93 INSTALLATZON OP iVIB CRANE ZN THE UNIT 4 CONTAINMENT BUILDING AT ELEVATION 58~0'8ammar This engineering package provided the necessary engineering requirements and details for the procurement and installation of a jib crane in the Unit 4 Containment Building.Originally, the polar crane was used to move relatively small loads between two containment building elevations.

Tool boxes, small equipment, etc., were staged and removed from the 58'-0" elevation during early and latter parts of a refueling outage.At these times, the use of the polar crane was limited to handling large containment loads.For, lighter loads, personnel were used to move the loads via the stairs or the containment elevator.This engineering package removed these ineffective and potentially unsafe load transfer activities by installing a jib crane.The jib crane is capable of hoisting loads up to 1,500 pounds.The crane is 10 feet high, with an eight foot rotating boom.The jib crane can be used during Modes 5, 6 and defueled.The jib crane will not be used in Modes 1, 2, 3 or 4 due to potential for a load drop accident scenario causing unsafe plant conditions.

8afet Evaluation:

The jib crane, installed along the west side of the equipment hatch was limited to 1,500 pounds to avoid heavy load requirements which would result in substantial structural modifications due to large factors of safety.The jib crane was equipped with boom rotation stops to prevent striking the containment liner plate with the end of the boom.The jib crane was not a single failure proof device.As a result the jib crane operation will be limited to Modes 5, 6 or defueled in order to preclude the potential for an inadvertent load drop scenario.The modification in this Engineering Package did not have'any adverse effect on plant safety or plant operations.

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

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

72 PLANT CHANGE MODIFICATION 93-020 UNIT TURN OVER DATE 4 05/21/93 REACTOR COOLANT PUMP 4A MOTOR REFURBISHMENT UPGRADE~summar As part of the on-going program to improve the reliability and performance of,all Reactor Coolant Pumps (RCP)at Turkey Point, this engineering package documented the upgrade of a spare motor to be installed in the 4A RCP slot.The original motor was replaced with the rotated spare motor, which was refurbished at the Westinghouse Electro-Mechanical Division facility.This standard factory refurbishment consisted of inspection and maintenance activities performed to the existing design specifications.

In addition, modifications were conducted, concurrent with the refurbishment, to insure consistency with the latest RCP technology and to realize additional reliability and availability.

Upon completion of the modifications, the motor was assembled, balanced and tested and shipped back to PTN.Safet Evaluation:

The RCP motor does not perform any safety related function, with the exception of providing sufficient inertia (through its flywheel)'to ensure sufficient coastdown of the Reactor Coolant Pump after an RCP trip.The RCP motor modifications did not affect the coastdown characteristics of the motor.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

73 SECTION 2 SAPETY EVALUATIONS 74 1

SAFETY EVALUATION ZPE-M-86 011 REVISION 0 UNIT TURNOVER DATE 4 03/15/93 SAFETY EVALUATION FOR CPSOs86-017 AND 86-018 UNIT 4 REPLACEMENT OF NORMAL&EMERGENCY CONTAINMENT COOLER DRIP PANS~8UBlRIL This safety evaluation was written to support the plant changes implemented under PC/M 86-017 and PC/M 86-018, whose implementation activities were completed and turned over to the plant by March 15, 1993.The purpose of these PC/Ms was to fabricate and replace the existing drip pans for the Unit 4 Normal and Emergency Containment Coolers inside containment.

The original galvanized steel drip pans collected condensate from the cooling coils and were in a corroded condition.

These galvanized drip pans were replaced with stainless steel pans of the same thickness, which did not require any additional supports or restraints due to the equivalency of weights of both assemblies.

All rework duplicated, the existing drip pan design with the exception of material.Similarly, all carbon steel fittings were replaced with equivalent stainless steel fittings.Safet Evaluation:

All replacement drip pans duplicated the existing drip pan design with the exception of material.The removal and reinstallation of the Normal and Emergency Containment Cooler drip pans did not affect any other system nor require any other component to be taken out of service.The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations.

The plant changes in hardware 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 changes identified in this safety evaluation.

75 SAFETY EVALUATION aI'PE-M 86-033 REVISION 0 UNITS TURNOVER DATE 3&4 05/13/93 SAFETY EVALAUTION FOR CPRO 86-086 RELOCATION OF EMERGENCY DIESEL GENERATOR COOLING SYSTEM DRAIN VALVES 293A AND 293B~summa This safety evaluation was written to support the plant changes implemented under PC/M 86-086, whose implementation was completed and turned over to the plant by May 13, 1993.The purpose of this PC/M was to relocate Emergency Diesel Generator (EDG)A&B cooling system drain valves 293A and 293B from their existing location outside the vital area barrier to within the barrier confines for the EDGs.In addition, valves 292A and 292B were located on the same drain lines from the diesel radiator shells, but were located on inside the vital area barriers.Under normal operating conditions, all valves are normally closed with 293A and 293B locked closed.Safet Evaluation:

The relocation of the drain lines did not affect the intended function of the EDG radiator shell drain lines or associated isolation valves.A seismic evaluation performed on the revised configuration of the radiator drain line with isolation valves installed demonstrated that this.revised configuration could withstand a design basis seismic event.The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations.

The plant changes in hardware 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 changes identified in this safety evaluation.

76 SAFETY EVALUATION aTPE-M 86-067 REVISION 0 UNITS TURNOVER DATE 3&4 03/29/93 SAFETY EVALUATION FOR CPROs86-163 PASS CHLORIDE REAGENT AND CALIBRATION STANDARD PUMPS SUBSTITUTION

~mumm ar This safety evaluation was written to support the plant changes implemented under PC/M 86-163, whose implementation activities were completed and turned over to the plant by March 29, 1993.The purpose of this PC/M was to substitute two existing Post Accident Sampling System (PASS)chloride reagent and calibration standard positive displacement metering pumps with pumps of a different type.The original positive displacement pumps design discharge pressure range was 20 and 30 psig, respectively; while, the system pressure range was approximately 50 psig.This excessively high pressure differential was considered to directly contribute to the high failure rate of the original type of pumps.The replacement of the original pumps with diaphragm-type pumps was considered to be an appropriate substitution for this system.After replacement pump testing was performed to verify that no leakage was present.Safet Evaluation:

The replacement of the original pumps with diaphragm-type pumps was considered to be an appropriate substitution for this system.The work performed under this PC/M did not affect any plant features necessary to assure the integrity of the reactor coolant pressure boundary, nor did it hamper the capability to shutdown the reactor and maintain in safe shutdown condition following design basis events.The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations.

The plant changes in hardware 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 changes identified in this safety evaluation.

77 SAFETY EVALUATION aTPES PTP 86 433E REVISION 0 UNIT TURNOVER DATE~4 03/23/93 SAFETY EVALUATION POR CPRO 86-035 UNIT 4 VALVE POSITIONER REPLACEMENT FOR PCV-4-455 A 5 B~summa This safety evaluation was written to support the plant changes implemented under PC/M 86-035, whose implementation activities were completed and turned over to the plant by March 23, 1993.The purpose of this PC/M was replacement of the valve positioner on pressurizer spray valves PCV-4-455A and PCV-4-455B.

The original Bailey positioner model was no longer available and a suitable replacement approved for nuclear service was selected, i.e., Barton Conoflow positioner.

Replacing the original positioner with the Barton Conoflow positioner enhanced the reliability of the pressure spray valve operation, since the replacement positioner had better documented qualifications than the original positioner, which was supplied as an accessory on the Copes-Vulcan valves.In addition, since the Copes Vulcan valves had been relocated outside of the pressurizer cubicle with its normally high radiation levels, the expected dose to replacement positioner would be substantially reduced.Safet Evaluation:

The conclusion of the safety evaluation was that the positioner replacement was considered to enhance the reliability of the pressure spray valve operation, since the replacement positioner had better documented qualifications and was located in a lower radiation environment.

The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations.

The plant changes in hardware 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 changes identified in this safety evaluation.

78 SAFETY EVALUATION ZPES-E-87-384 REVISION 0 UNIT TURNOVER DATE 4 03/29/93 SAFETY EVALUATION FOR CPROS 87 060 AND 87 061 PRMS DRAWER REPLACEMENT

~8URRR This safety evaluation was written to support the plant changes implemented under PC/M 87-061, whose implementation was completed and turned over to the plant by March 29, 1993.The purpose of this PC/M was to replace several PRMS drawers which required maintenance.

Each PRMS drawer, contained the electronics and hardware to power its associated radiation detector, to process the signal, and to provide digital and analog outputs.Due to schedular problems, the original equipment could not be supplied in a timely fashion and the vendor offered an upgraded model of drawer.This seismically qualified upgraded PRMS drawer was installed as a replacement for existing drawers in Unit 4 PRMS channels R-11, R-12, R-15, R-17A, R-17B, and R-19.The replacement drawers were equivalent in form, fit and function to the original equipment.

Performance specifications for the new drawers indicated that parameters, such as, accuracy, response time, and repeatability were equivalent to the original drawers.Safet Evaluation:

The upgraded PRMS drawer replacement did not change the system functional design basis or the system configuration.

In addition, the replacement drawers were equivalent in form, fit and function to the original equipment.

The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations.

The plant changes in equivalent hardware 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 changes identified in this safety evaluation.

79 i 0' SAPETY EVALUATION JPN-PTN-SEEJ 88 042 REVISION 2 UNIT~4 APPROVAL DATE: 04/08/93 DE ENERGIZATION OP UNIT 4 4l.60 VOLT SAFETY RELATED BUSSES~slmmR This evaluation was developed to establish the requirements and restrictions which must be placed on the operation of Units 3 and 4 and their equipment when a Unit 4 4160 volt bus was de-energized and Train A and B load centers were cross-connected.

Also examined were technical and licensing concerns associated with de-energizing safety related equipment and removing on EDG from service concurrent with a Unit 4 4160 volt bus de-energization.

The de-energization of a Unit 4'160 volt safety related bus, with Unit 4 in cold or refueling shutdown (Modes 5 or 6)or de-fueled and Unit 3 at power operation (Mode 1)or below, is sometimes necessary to allow for periodic maintenance, testing, or design modifications of the 4160 volt switchgear.

De-energization of a 4160 volt bus would cause de-energization of the 480 volt load centers and motor control centers powered from that bus, if any, and a loss of power to equipment which may be required to maintain cold/refueling shutdown, perform outage'elated activities, or support safe shutdown and accident mitigation on the opposite Unit.This condition was alleviated by closing the tie-breakers between opposite train 480 volt load centers, while one 4160 volt bus was de-energized or by ensuring that alternate equipment was available.

Safet Evaluation:

This safety evaluation addressed the technical and licensing requirements for the de-energization of each Unit 4 4160 volt bus 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 (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.

80 SAFETY EVALUATION aTPN PTN SEMaT-88 090 REVISION 0 UNIT a 3 TURNOVER DATE: 08/11/92 SAPETY EVALUATION OP THE DELETION OF PIRE HOSE STATIONS IN THE RADWASTE BUILDING J~sammar This safety evaluation was written to support the plant changes implemented under PC/M 88-603, whose implementation activities were completed and turned over to the plant by August 11, 1992.This PC/M was developed to correct drawing descrepancies identified on Operating Diagram 5610-T-E-4072, Sheet 1.A field verification walkdown of the fire protection system and a review of 5610-T-E-4072, Sheet 1 revealed several drawing discrepancies.

This drawing showed to fire hose stations in the Radwaste Building as part of the fire protection system.However, field walkdowns confirmed that these two hose stations were actually part of the non-safety related service water system and not the fire protection system.This drawing was corrected to match the existing field confirguration.

This safety evaluation verified the ability of the existing fire protection system to meet licensing basis criteria from the Updated FSAR.Safet Evaluation:

The investigation confirmed that the Radwaste Building was not required to be covered under the fire protection program to meet NRC requirements, and therefore hose stations were not required in that building.Therefore, the Updated FSAR and engineering drawings may be updated accordingly without affecting the validity of the plant fire protection program and the overall safe operation of the plant.The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations.

The document changes 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 changes identified in this PC/M and safety evaluation.

81 0

SAFETY EVALUATION i7PN PTN SEEaT-89-085 REVISION 3 UNIT APPROVAL DATE 3 10/08/92 DE ENERGIZATION OF UNIT 3 4160 VOLT SAFETY RELATED BUSSES~summa This evaluation was developed to establish the requirements and restrictions which must be placed on the operation of Units 3 and 4 and their equipment when a Unit 3 4160 volt bus was de-energized and Train A and B load centers were cross-connected.

Also examined were technical and licensing concerns associated with de-energizing safety related equipment and removing an EDG from service concurrent with a Unit 3 4160 volt bus de-energization.

The de-energization of a Unit 3 4160 volt safety related bus, with Unit 3 in cold or refueling shutdown (Modes 5 or 6)or de-fueled and Unit 4 at power operation (Mode 1 or below)is sometimes necessary to allow for periodic maintenance, testing, or design modifications.

De-energization of a 4160 volt bus would cause de-energization of the 480 volt load centers and motor control centers powered from that bus, and a loss of power to equipment which may be required to maintain cold/refueling shutdown, perform outage related activities, or support accident mitigation on the opposite unit.This condition was alleviated by closing the tie-breakers between opposite train 480 volt load centers, while one 4160 volt bus was de-energized or by ensuring that alternate equipment was available.

Safet Evaluation:

This safety evaluation addressed the technical and licensing requirements for the de-energization of each Unit 3 4160 volt bus.It concluded that the proposed plant configuration and mode of operation was bounded by the Technical Specifications and did not change the analysis for 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.

82 SAFETY EVALUATION JPN-PTN-SENP-92-001 REVISIONS 1&2 UNITS 3&4 APPROVAL DATES: Rev.1 08/06/92 Rev.2 03/11/93 THE CONDUCT OF INTEGRATED SAFEGUARDS TESTING ON A SHUTDOWN UNIT'RITH THE OPPOSITE UNIT AT POWER~summa The purpose of this evaluation was to identify and resolve technical and licensing concerns associated with the performance of integrated safeguards testing on one unit with the opposite unit at power.This evaluation established the acceptability of performing train-by-train integrated testing on the shutdown unit with the opposite unit at power or any other mode of operation, while confirming that all technical specification requirements were met.The principal objectives of the revised test procedure as evaluated in this safety evaluation was to satisfy Technical Specification surveillance and test requirements for the onsite emergency power system and safeguards equipment on a train-by-train basis, while allowing continued, uninterrupted power operation of the non-test unit without placing either unit in an unanalyzed or unsafe condition.

Revision 1 of this evaluation incorporated a review of the current changes to the IST procedure and incorporated minor plant comments.Revision 2 of this evaluation incorporated'hanges to the EDG loading charts, which address limitations experienced with the Unit 3A EDG during surveillance testing and provided additional acceptance criteria.All other portions of this safety evaluation remained unchanged.

Safet Evaluation:

Based on the requirements, restrictions and precautions specified and discussed in this safety evaluation and Technical Specifications, the performance of the proposed revised integrated safeguards testing in accordance with the revised plant procedures, 3/4-0SP-203.1 and 3/4-0SP-203.2, did not have any adverse effect on plant safety or plant operations.

The actions and 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 and changes identified in this safety evaluation.

83 SAPETY EVALUATION aTPN-PTN-SEEP>>92 008 REVISION 0 UNITS 3&4 TURNOVER DATES: 92-030 01/07/93 92-,032 05/21/93 SAPETY EVALUATION FOR LOAD CENTER AND RELAY SETTING CHANGES~mumm ar This safety evaluation was written to support the plant changes implemented under PC/M 92-030 and PC/M 92-032, whose implementation activities were completed and turned over to the plant by January 7, 1993 and May 21, 1993, respectively.

These PC/Ms were developed to resetting one overcurrent relay per unit and provide overcurrent circuit breaker settings for safety related load center breakers.The overcurrent relay which was reset on each unit was for the Train"A" 4160 VAC switchgear circuit breaker feed from the adjacent unit's start-up transformer.

This alternate electrical feed provided each unit with a second source of offsite emergency power which was capable of supporting the loads necessary for achieving and maintaining safe shutdown.Providing the overcurrent circuit breaker settings for the safety related load center breakers was intended to document the engineering specified settings and incorporated these settings into the controlled drawing system.This safety evaluation established the basis for any relay setting changes that were necessary.

Safet Evaluation:

Overcurrent relay and circuit breaker settings served to protect equipment during electrical fault and abnormal overload conditions.

The overcurrent device settings were based on engineering calculations which ensured electrical protection and coordination.

These changes did not add any new component or change the function, operation, and design basis of any existing equipment described in the SAR.The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations.

The document changes 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 changes identified in these PC/Ms and safety evaluation.

84 SAFETY EVALUATION iTPN PTN-SEFZ-92-012 REVISIONS 1&3 UNITS 3&4 APPROVAL DATES: Rev.1 08/20/92 Rev.3 04/16/93 EVALUATION OP ZMPACT OF ACCUMULATOR DZSCHARGE TEST ON PUEL AND REACTOR ZNTERNALS~8'mm8 This safety evaluation documented the acceptability of performing the accumulator discharge test with fuel and upper intenals in the reactor vessel.The test allowed the full stroke exercise of the accumulator discharge check valves.It was performed by discharging the accumulator water volume into the RCS at a predetermined pressure that was sufficient to fully open the downstream check valves.This test was previously done on Unit 3 without fuel in the reactor vessel.In this evaluation, Nuclear Fuels investigated any adverse effects on the fuel and reactor vessel internals which could occur as a result of this test.Revision 1 to this evaluation examined the performance of the test with the upper internals in the reactor vessel and addressed industry experience with this test, such as, the Wolf Creek Plant contamination incident of 1988.Revision 3 provided tolerances and clarification of the 50 second requirements to close the accumulator isolation valve after the initiation of the test.Safet Evaluation:

The proposed configuration was a normal plant evolution in Mode 6, in which mode the test was performed.

The effects of the accumulator discharge test on the fuel and the reactor internals were bounded by normal plant evolutions.

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.

85

SAFETY EVALUATZON ZPN-PTN-SEFZ-92-0I.S REVISION 0 UNITS 3 6 4 APPROVAL DATE: 09/25/92 EVALUATZON OF ACCUMULATOR DZSCHARGE TEST RZTH REACTOR VESSEL HEAD ZNSTALLED~SUlHE This safety evaluation documented the acceptability of performing an accumulator discharge test with fuel and upper internals in the reactor vessel and the reactor vessel head installed.

The test allowed the full stroke exercise of the accumulator'discharge check valves.It was performed by discharging accumulator water volume into the RCS at a predetermined pressure that was sufficient to fully open the downstream check valves.This test was previously done on Unit 3 without fuel'n the reactor vessel and the reactor vessel head removed.The impact on fuel and reactor internals of the increased flow experienced in the reactor vessel during the opening of the accumulator isolation valve, and the potential for release of nitrogen into the reactor vessel had been evaluated previously in another safety evaluation.

This safety evaluation focused on additional potential issues resulting from the reactor vessel head being installed.

These areas of concern were identified as follows: (1)RCS pressurization at low temperature and the impact on the circumferential weld of the reactor vessel;(2)the flow increase experienced in the pressurizer and the impact on the pressurizer heaters;(3)over-filling of the pressurizer and spilling into the Pressurizer Relief Tank (PRT);and (4)the impact on accumulator thermal stresses.Safet Evaluation:

The proposed configuration was a normal plant evolution in Mode 6, in which mode the test was performed.

The effects of the accumulator discharge test on the fuel and the reactor internals were bounded by normal plant evolutions.

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.

86 0

SAFETY EVALUATION aTPN-PTN-SECS 92 018 REVISION 0 UNITS 3&4 APPROVAL DATE: 08/04/92 TEMPORARY LEAD SHIELDING INSTALLATION SPECIFICATION SPEC-C 003~mumm ar This engineering evaluation examined Specification SPEC-C-003,"Temporary Lead Shielding Installation Specification", which provided engineering guidance and requirements for the installation of temporary lead shielding at Turkey Point Units 3 and 4.Lead shielding could be in the form of blankets, sheets or bricks which could be configured to form temporary lead shielding barriers.These barriers could be supported from permanent or temporary plant structures, or could be applied directly to piping systems.The specification prohibited the use of these barriers in Modes 1, 2, 3 or 4, and allowed their use in Modes 5, 6 or defueled, only if the specific set of implementation instructions accompanying each shielding barrier allowed it.The intent of the specification was to present a convenient set of temporary plant configurations which had been assessed by engineering for impact on nuclear safety.This was done to avoid performing repetitive engineering evaluations for the installation of frequently used lead shielding barriers.Safet Evaluation:

Temporary lead shielding barriers covered under the scope of the specification did not perform safety related functions, nor did they alter plant operations, design bases or technical specifications.

Their installation was considered a temporary change to the facility which was evaluated to satisfy plant licensing requirements.

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.

87 0

SAFETY EVALUATION iTPN PTN SECP 92 021 REVISION 0 UNIT APPROVAL DATE 4 08/11/92 UNIT 4 TWENTIETH YEAR CONTAINMENT TENDON SURVEILLANCE

~summa The purpose of this safety evaluation was to address construction and surveillance activities associated with the Unit 4 twentieth year containment structure tendon surveillance.

The tendon surveillance program is an in-service physical inspection of the concrete containment post-tensioning system, which satisfied plant Technical Specification requirements for the twentieth year surveillance.

This surveillance program is a systematic means of assessing the continued performance of the containment post-tensioning system.Technical Specification Section 4.6.1.6.1 required that three dome, five hoop (horizontal), and four vertical tendons be selected for the twentieth year surveillance based on a random and representative selection process.Also, other inspection activities were performed for data collection purposes, but were not required to satisfy Technical Specifications.

Safet Evaluation:

The performance of this tendon surveillance did not compromise the containment structural integrity, because the conditions for testing, as described in Updated FSAR Section 5.1.7.4 and Technical Specification Sections 4.6.1.6.1 and 4.6.1.6.2 were maintained.

Similarly, this activity did not create any spatial or functional adverse interaction with any structure, system or component important to safety or safe plant operation.

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.

88 SAFETY EVALUATION ZPN PTN-SENP 92 032 REVISIONS 0-2 UNITS 3&4 APPROVAL DATES: Rev.0 09/03/92 Rev.1 09/04/92 Rev.2 10/02/92 THE DEMOLITION OF THE TURKEY POINT FOSSZL UNIT 1 CHZMNEY~81HOEIR The purpose of this safety evaluation was to analyze the effect on the nuclear units during a controlled demolition of the Turkey Point Fossil Unit 1 chimney.Hurricane Andrew hit south Florida causing damage to equipment and structures on both the nuclear and fossil units.Selected site damage that was incurred included visible structural damage to the Fossil Unit 1 chimney.Although the Unit 1 chimney remained standing, the extensive nature of the damage raised concerns that the chimney may not have survived another high wind event.The chimney could have fallen in an uncontrolled manner and represented a potential hazard to personnel and the Turkey Point Nuclear Units.Conventional dismantling of the chimney would taken several months and the required technique itself represented a personnel safety hazard.Therefore, the Unit 1 chimney was raised in a controlled manner using precision demolition to cause it to fall in a safe and predictable direction.

Based on a detailed inspection, the Unit 2 chimney, which is the closest chimney to the nuclear units, did not suffer any significant structural damage.t Revision 1 clarified the limitations on wind speed and direction applicable to the demolition plan.Revision 2 of the safety evaluation was issued to provide a final report, as described in attachments to the safety evaluation.

Safet Evaluation:

This activity has been evaluated and equipment important to safety will remain functional during and following the felling of the Unit 1 chimney.No adverse interactions involving safety related equipment would be created by the demolition of the Unit 1 chimney.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.

89 0

SAFETY EVALUATION a7PN-PTN SEMS-92 033 REVISION 0 UNIT APPROVAL DATE 3 08/13/92 FREERE SEAL EVALUATION FOR REPLACEMENT OF VALVES 3-777 3-834 AND 3 833~summa This safety evaluation addressed the use of freeze seals, temporary supports, and a seismic evaluation while replacing the Component Cooling Water (CCW)valves 3-777, 3-834 and 3-833.CCW from the non-regenerative heat exchanger bypass throttle valve 3-834, CCW supply to non-regenerative heat exchanger isolation valve 3-777, and TCV-3-144 inlet isolation valve 3-833 had all deteriorated to the point where they could not perform their normal functions due to excessive seat leakage.The maintenance was performed in two phases.The first phase, performed in Modes 5 and 6 during the Unit 3 refueling outage, replaced valves 3-777 and 3-834, which required closing valve 3-781 and establishing freeze seals to isolate the work area.The second phase of maintenance replaced valve3-833, using valves 3-777, 3-834 and 3-780 as boundaries.

This phase of the maintenance could have been performed any time during the refueling outage after the completion of the first phase but prior to returning the non-regenerative heat exchanger to service.Ik Safet Evaluation:

Reduced water inventory operations are sensitive to a loss of decay heat removal capability.

However, administrative controls on the freeze seal operation and operator actions precluded the loss of decay heat removal capability during the maintenance activity any adverse interactions with equipment important to safety.Additionally, analysis was performed on the piping and pipe supports to ensure that acceptable loadings were not exceeded any time during the maintenance.

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.

90 SAFETY EVALUATION iXPN-PTN SENP-92 033 REVISIONS 0&1 UNITS 3&4 APPROVAL DATES: Rev.0 09/19/92 Rev.1 09/24/92 SAFETY EVALUATION RELATED TO THE TURKEY POINT FOSSIL UNIT 2 CHIMNEY~summa The purpose of this safety evaluation was to analyze the effect on safety of operating the nuclear units with the Unit 2 chimney in its current post-hurricane condition.

Hurricane Andrew hit south Florida causing damage to equipment and structures on both the nuclear and fossil units.Specific site damage that was incurred included minor vertical and horizontal cracking to the Unit 2 chimney, which stands next to the nuclear units.In most cases the cracks were hairline cracks with little or no spalling or signs of distress within the lower 150 feet.An analysis conducted by Failure Analysis Associates (FaAA)conservatively modeled the cracks and showed that the stack would not fail under the original design load of 55 psf (approximately equivalent to 145 mph wind)or following a 0.15g seismic event.The results of this evaluation were corroborated by a second independent evaluation.

Revision 1 to this safety evaluation incorporated the results of additional structural analyses, which showed structural margins for the Unit 2 chimney of 554 for a 55 psf wind load, 404 for a 145 mph FSAR wind load, and 254 for a 225 mph FSAR tornado wind load.These analyses demonstrated the ability of the Unit 2 chimney to withstand the design basis natural phenomenon (hurricane, tornado, and seismic)without interacting adversely with the nuclear units.Safet Evaluation:

Although the likelihood of a Unit 2 chimney failure resulting in damage to equipment important to safety is a low-probability event, this remote possibility was evaluated to determine the consequences of such an extraordinary event.This evaluation concluded that even the worst case scenarios of equipment damage could be accommodated with core damage and that current plant procedures are in effect to cope with equipment damage events.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 changes, 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.

91 SAFETY EVALUATION aTPN-PTN-SEMaT 92 034 REVISIONS 0-2 UNITS 3&4 APPROVAL DATES: Rev.0 09/24/92 Rev.1 09/25/92 Rev.2 09/30/92 INTERIM FIRE PROTECTION SYSTEM CONFIGURATION TO SUPPORT UNIT 4 STARTUP~summa The purpose of this safety evaluation was to identify the Fire Water Supply System licensing and design basis requirements and determine what system configuration requirements were needed for Unit 4 startup following Hurricane Andrew.High winds associated with Hurricane Andrew caused the Turkey Point (PTN)Raw Water System high tower to collapse.As a result of the collapsed high tower, portions of the PTN Fire Water Supply System were damaged, including the electric driven fire pump, both fire water jocket pumps, and portions of the fire protection piping system.Although several plant modifications were under development to restore the Fire Protection facilities, these modifications were not fully implemented in time to support startup of Turkey Point Unit 4.This evaluation examined the interim Fire Protection System configuration which was evaluated against the system operability requirements specified in the Turkey Point Technical Specifications and the updated FSAR.The necessity for supplemental fire protection equipment to meet system design requirements was also determined.

Revision 1 of the evaluation provided additional information regarding the performance capabilities of the screen wash pumps.Revision 2 of the evaluation clarified the operational requirements of the jockey pumps to support unit startup.Safet Evaluation:

As discussed in this safety evaluation, the interim Fire Water Supply System utilizing the Raw Water Storage Tank II with the permanent electrical and diesel driven fire pumps and the intake canal with the screen wash pumps was capable of delivering the required fire water flow and remained capable of mitigating the effects of a fire.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.

92 i

SAFETY EVALUATION iXPN-PTN-SECP-92 038 REVISION 0 UNITS APPROVAL DATE 3&4 11/27/92 SAFETY EVALUATION RELATED TO THE TURKEY POZNT FOSSIL UNITS 1 AND 2 CHIMNEY CONSTRUCTZON ACTIVZTZES

~summa The purpose of this safety evaluation was to address the construction activities associated with the erection of a new chimney for Turkey Point Fossil Unit 1 and the reinforcement of the Fossil Unit 2 chimney which were damaged during Hurricane Andrew on August 24, 1992.Specific site damage that was incurred included visible structural damage to the Turkey Point Fossil Unit 1 chimney and minor cracking to the Unit 2 chimney.The damage to the Unit 1 chimney was sufficiently severe to require its demolition.

The post-hurricane condition of the Unit 2 chimney was evaluated in another safety evaluation.

This evaluation concluded that the chimney had sufficient remaining capacity to withstand the Turkey Point Updated FSAR loads for Class I structures without adversely interacting with the nuclear units.However, due to the long term corrosion problems a new sheath would be constructed around the Unit 2 chimney.The scope of this evaluation was limited to the erection of the Unit 1 chimney and preparations for the reinforcement of the Unit 2 chimney up to, but not including, placement of concrete.Safet Evaluation:

The construction activities associated with the new Unit 1 chimney and reinforcement of the Unit 2 chimney up to, but not including, Unit 2 concrete placement were reviewed.All equipment and materials were confined to the Units 1 and 2 side of the site and would not affect the nuclear units.The new Unit 1 chimney was analyzed to show that it wwould be able to withstand the wind and seismic loads defined in the FSAR for Class I structures without interacting with the nuclear units.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.

93

SAFETY EVALUATION a7PN PTN-SECP-92-040 REVISION 0 UNITS 3 6 4 APPROVAL DATE 12/08/92 SAFETY EVALUATION RELATED TO THE NEW TURKEY POINT FOSSIL UNIT 1 CHIMNEY AND UNIT 2 CHIMNEY REINFORCEMENT

~summa r The purpose of this safety evaluation was to document the design criteria which were used in the design of a new Unit 1 chimney and the reinforcement of the original Unit 2 chimney.The original fossil chimney's were damaged during Hurricane Andrew on August 24, 1992.The criteria used ensured that the new and repaired chimneys could withstand the loads defined in the Turkey Point Updated FSAR for Class I structures without interacting with the nuclear units.This evaluation addressed the potential effects of the Unit 2 concrete placement on the safe operation of the nuclear units, since failure of the chimneys and/or construction accidents could potentially affect nuclear safety related equipment.

Safet Evaluation:

The criteria used in the design of the new Unit 1 chimney and the reinforcement of the Unit 2 chimney ensured compliance with all existing building codes, and also ensured that there was no potential for interaction with the nuclear units under the wind and seismic loads defined in the Updated FSAR for Class I structures.

The design of the new and reinforced chimneys was verified by an independent consultant (Failure Analysis Associates).

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.

94 SAFETY EVALUATION iJPN PTN-SENS92-044 REVISION 1 UNITS APPROVAL DATE 3&4 09/30/92 MANUAL OVERRIDE OF MOV*626 DURING RCP SEAL FAILURE~summa This safety evaluation examined the effects of manually overriding the automatic operation of MOV-*-626 (by opening/verifying open the valve)following an reactor coolant pump (RCP)seal failure and using an operator dedicated to restoring electric power to the MOV when required.MOV-*-626 is the CCW return isolation valve common to all the RCP thermal barrier heat exchangers.

MOV-*-626 is part, of the containment isolation scheme for contaiment Penetration No.43.A Westinghouse bullentin described how the failure of a No.1 RCP seal could result in a loss of CCW to all RCP thermal barrier heat exchangers due to the automatic closure of MOV-+-626.

This would result in a loss of CCW cooling to all RCP thermal barrier heat exchangers, which potentially leads to the failure of the unaffected RCP seals.This safety evaluation redefined the design basis for containment Penetrations No.3, 4 and 43 to allow the closed system inside containment to be one of the required barriers.Safet Evaluation:

The redefined design bases for containment Penetrations No.3, 4, and 43 satisfy the two barrier criterion for containment isolation and were successfully evaluated against the Updated FSAR single active failure criterion.

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.

95 SAFETY EVALUATION ZPN PTN-SENS-92-045 REVISION 0 UNIT APPROVAL DATE 3 08/21/92 FREEZE SEAL ZNSTALLATZON ON THE HHSZ ALTERNATE HOT LEG ZNZECTZON CROSS-TIE PIPING~summa r This evaluation examined the installation of a freeze plug on the alternate hot leg injection cross-tie header for the performance of flow testing of the High Head Safety Injection (HHSI)pumps.This testing was performed in a mode when the safety injection system was not required to be operable by technical specifications.

Normal maintenance or testing performed on a system not required to be operable by the technical specifications does not generally require evaluation under the provisions of 10 CFR 50.59.However, site policy governing evolutions for the use of freeze seals was under development, and it was considered prudent at the time to evaluate the piping configuration against the criteria of 10 CFR 50.59.Safet Evaluation:

The freeze seal was installed on the HHSI alternate hot leg injection cross-tie piping which was not required to be operable per the Technical Specifications.

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.

96 SAPETY EVALUATION ZPN PTN-SEMS-92-052 REVISION 0 UNIT APPROVAL DATE 3 10/15/92 SAPETY EVALUATION FOR ICW VALVE REPLACEMENT

~summa'his safety evaluation covered isolation, removal, and replacement of ICW header cross-connect valves 3-50-307 and 3-50-350.The purpose of this safety evaluation was to assess all potential safety concerns associated with activities for the replacement of these valves.The replacement work was performed with Unit 3 in Mode 6 or with the reactor defueled and all the spent fuel stored in the spent fuel pool (SFP).The valves were replaced due to excessive seal leakage, making isolation of the ICW headers during maintenance crawl-through inspections difficult.

8afet Evaluation:

The ICW configurations that were established during the valve replacement maintenance were analyzed to ensure that the operable portions of the ICW system remained seismically qualified.

The ability of the ICW system to support Residual Heat Removal (LHSI)and SFP cooling during Mode 6, or SFP cooling with the reactor defueled and all fuel in the SFP was not adversely impacted by the valve replacement activities.

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.

97 SAFETY EVALUATION ZPN PTN-SECS-92-056 REVISION 0 UNITS APPROVAL DATE 3&4 09/22/92 INSTALLATION OF COMMUNICATION ANTENNAS~SlHMSR This evaluation addressed the acceptability of installing two fiberglass whip antennas on the control room roof under Temporary System Alteration (TSA)3-92-1-23.

The antennas were attached to the missile barrier separating the computer room HVAC units.As a result of Hurricane Andrew, offsite communications were interrupted due to loss of all communication paths from the site due to equipment damage.In order to increase the capability of the offsite communications system and significantly increase the probability of maintaining offsite communications paths during an event similar'to Andrew, a VHF and UHF.radio system was installed.

This safety evaluation addressed the mounting of the antennas and their potential interaction with safety related structures, systems and equipment because of their location and attachment to the control room missile barrier.Safet Evaluation:

This safety evaluation concluded that two antennas could be installed on the subject missile barrier provided that all of the requirements stipulated within this evaluation were followed.The evaluation also concluded that this activity will have no adverse impact on plant operations, and will not compromise the licensing bases for Turkey Point.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.

98 SAFETY EVALUATION a7PN-PTN-SENS-92-059 REVISION 0 UNIT APPROVAL DATE 3 10/08/92 UNIT 3 REFUELING OUTAGE CONTINGENCY PLAN FOR EMERGENCY POSER TO THE SFP PUMPS~SUEIIIE This safety evaluation provided a basis for a contingency plan to provide a source of emergency (back-up)power to the Unit 3 spent fuel pool (SFP)cooling pump motor transfer switch in the unlikely event of loss of normal power from Load Center 3C.The SFP cooling system did not include emergency power as a design requirement, and a SFP boiling analysis demonstrated that offsite doses will remain well within 10 CFR 100 limits.The contingency plan evaluated in this safety evaluation required that a cable of sufficient length be installed (~onl upon a loss of the normal power supply)between the SFP motor transfer switch and breaker 42116 in cubicle 4E of MCC 4H.The interconnecting cable is for use only during the Unit 3 refueling outages.Safet Evaluation:

The installation of a temporary cable has been evaluated electrically and seismically and will not adversely affect the SFP cooling system and adds reliability to the system during Unit 3 refueling outages.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.

99 SAFETY EVALUATION a7PN-PTN-SEMS-92-060 REVISION 0 UNIT~3 APPROVAL DATE: 10/09/92 INSTALLATION AND USE OP AN ABB CE RCCA INSPECTlON STATION AT TURKEY POINT~summa This safety evaluation evaluated the consequences of installation of an ABB/CE rod cluster control assembly (RCCA)inspection device at Turkey Point.The inspection device was installed on top of the spent fuel storage racks.This evaluation included the effect of the inspection stand, and the RCCA while in the stand, on the racks only.It did not consider the process of RCCA removal, storage, evaluation, subsequent RCCA disposal or re-insertion in the fuel.Safet Evaluation:

The potential safety issues associated with installing this equipment'ere enveloped by postulated accidents previously evaluated in the Updated FSAR.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.

100 SAFETY EVALUATION JPN-PTN-SEES-92 061 REVISION 0 UNITS APPROVAL DATE 3&4 10/12/92 EVALUATION FOR TSA 03 92 06-12 FIRE RATER PUMP TRIP UPON LOOP DURING 4160 VOLT BUS 3A DE-ENERGIZATION

~81HDKil For the Unit 3 refueling outage, this evaluation was developed in support of Temporary System Alteration (TSA)03-92-06-12, which provided a trip circuit scheme for the electric-driven Fire Water Pump (FWP)during the 3A 4kV bus outage.The FWP was powered from the 480 volt Load Center (LC)3C.The FWP was designed to trip upon the loss of voltagei however, the trip circuit would be disabled when the 3A 4kV load sequencer was removed from service as part of the 3A 4kV bus outage, and therefore allow the FWP to be auto-connected to EDG 3B in the first load block.The TSA required that wires of sufficient length be installed within LC 3C between spare contacts of two undervoltage relays, and FWP breaker control circuit.This would preclude the FWP from being automatically loaded onto Emergency Diesel Generator 3B upon initiation of a loss of offsite power.Safet Evaluation:

The temporary use of an alternate relay provided undervoltage protection to trip the fire water pump breaker open in the event of an undervoltage condition restored compliance with the design basis for the fire water pump, while the 3A 4kV bus was de-energized for maintenance.

The installation of a temporary jumper did not adversely interact with the 3B EDG or any equipment important to safety.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.

101 0,

SAFETY EVALUATION iTPN-PTN-SECS-92-063 REVISION 0 UNITS 3&4 APPROVAL DATE: 10/30/92 THE ZNSTALLATZON OF COMMUNICATION ANTENNAS TP-907~summa This evaluation addressed the acceptability of installing two antennas;one fiberglass whip antenna on the control room roof and one loop antenna on the Unit 4 EDG Building.As a result of Hurricane Andrew, offsite communications were interrupted due to loss of all communication paths from the site because of equipment damage.A comprehensive wireless system was being considered for installation in order to preclude communication losses in the future.Under Test Procedure TP-907, various communications tests were performed to assess the feasibility and performance of various antenna/radio systems.The subject antennas were installed on a temporary basis in order to accumulate test data pertaining to the acceptability for proposed antenna locations.

Safet Evaluation:

This safety evaluation addressed the mounting of the antennas and their potential interaction with safety related structures, systems and equipment and concluded that the antennas can be installed provided that all of the requirements stipulated within this evaluation were followed.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.

102 0

SAPETY EVALUATION ZPN-PTN-SEMS-92-066 REVISION 0 UNIT~3 APPROVAL DATE: 11/05/92 PREEZE SEAL SAPETY EVALUATION POR REPAIR OP CV 3-244~mamma r This safety evaluation addressed the use of a freeze seal in order to repair valve CV-3-244 at the discharge of the Chemical and Volume Control System (CVCS)demineralizers.

In order to perform corrective maintenance, a freeze seal was utilized to provide isolation from the letdown bypass path around the CVCS demineralizers.

The purpose of the freeze seal was to allow for the continued use of letdown.This maintenance was performed during Modes 5 and/or 6.Safet Evaluation:

During the maintenance activity, equipment important to safety required for accident mitigation remained available to perform its required safety functions.

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.

103 SAFETY EVALUATION ZPN-PTN-BECB-92 070 REVISION 1 UNIT~4 APPROVAL DATE: 11/24/92 REPLACEMENT OF CRDM 4A COOLER FAN MOTOR AT POWER OPERATION~summa This evaluation addressed the acceptability of replacing the 4A CRDM cooler fan motor while Unit 4 was in power operation (Mode 1).The evaluation addressed the use of the Polar Crane in Mode 1 operation, including identification of safe load paths, consequences of load drops on safety related equipment, and seismic considerations.

It further addressed the use of scaffolding and radiation shielding including adverse seismic interactions with safety related equipment, the effect of high energy line break jet impingements, and other potentials for adverse interactions.

Finally, it addressed the effects of the removal of the fan plenum and motor on the structural integrity of the CRDM cooler ductwork and associated CCW lines.Revision 1 of this evaluation provided additional clarification of the response to concerns related to the potential for sump screen blockage by debris.Safet Evaluation:

This evaluation concluded that this activity would have no adverse impact on the plant operations, and would not compromise the safety and licensing bases for Turkey Point Unit 4.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.

104 SAFETY EVALUATION i7PN PTN SECS 92 071 REVISION 0 UNIT APPROVAL DATE 3 11/20/92 SAFETY EVALUATION FOR ALLOlVING A MAN-BASKET TO REMAZN WITHIN CONTAINMENT DURING ALL MODES OF OPERATION~Smnn5 This evaluation addressed the acceptability of leaving a man-basket within the Unit 3 Containment Structure during all modes of operation.

The man-basket in combination with the Polar Crane was utilized during the Unit 3 Cycle 13 refueling outage for maintenance activities and for valve manipulations in preparation for the integrated leak rate test (ILRT).In order to remove the man-basket, the containment equipment hatch would be required to be opened.Due to schedular considerations Nuclear Engineering investigated the acceptability of allowing the basket to remain within containment during all modes of operation.

This safety evaluation concluded that the man-basket can remain within the containment structure during all modes of operation provided that all of the requirements stipulated within this evaluation were followed.Safet Evaluation:

The storage of a steel man-basket secured to structural steel on the 58 foot elevation within containment will not interact with any equipment that performs a safety function.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.

105 0

SAFETY EVALUATION i7PN PTN-SEMS-92 072 REVISION 0 UNIT TURNOVER DATE 3 11/27/92 SAFETY EVALUATION FOR LT-3 494 VENT PATH MODIFICATION

~SUlKR This safety evaluation was written to support the plant changes implemented under PC/M 92-176, whose implementation was completed and turned over to the plant by November 27, 1992.This PC/M was developed to address the modification of a steam generator level transmitter (LT-3-494) vent path by the removal of valve 3-20-802, which was replaced with a pipe cap.Valve 3-20-802 had been identified as requiring replacement during the ongoing refueling outage.During the analysis of this replacement valve, it was noted that the stress in the line containing the two series vent valves did not meet FSAR allowable stresses.To correct this condition, removal of the top most vent valve, 3-20-802, was required.This valve was replaced with a pipe cap, which served to to provide the same isolation function as the original valve.Safet Evaluation:

Valve 3-20-803 was the primary pressure boundary for the steam generator level transmitter LT-3-494 and the installed pipe cap provided the same backup pressure isolation as the original valve 3-20-802.The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations.

The hardware changes 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 changes identified in this PC/M and safety evaluation.

106

SAFETY EVALUATION JPN-PTN-SENS-93-007 REVISION 0 UNIT APPROVAL DATE 4 03/30/93 TEMPORARY REMOVAL OP STEAM GENERATOR 4C THRUST BEAM~summa This safety evaluation established requirements for the temporary removal and reinstallation of structural components to accommodate the Reactor Coolant Pump (RCP)motor replacement during refueling outages.To provide adequate clearance for rigging motors through the Unit 4 Containment equipment hatch and facilitate staging for the refueling outage, the Steam Generator 4C thrust beam, floor steel, handrail, grating and pipe supports for the 2-inch containment primary water service connections and 2-inch containment service air piping above the equipment hatch must be temporarily removed.Following the outage all components were replaced.Safet Evaluation:

No permanent change in the plant configuration was involved.The structural items removed were reinstalled to the same configuration and to the same design requirements as the original installation.

The effects on existing systems, structures, and components due to the temporary removal of these structural items were evaluated with respect to plant operational modes.The temporary changes identified and evaluated in this safety evaluation did not have any adverse effect on plant safety or plant operations.

The temporary plant modifications, 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 temporary changes identified in this safety evaluation.

107 0

SAFETY EVALUATION a7PN PTN SEMS-93 009 REVISION 0 UNITS APPROVAL DATE 3&4 03/16/93 MACHINING OF MOTOR OPERATED VALVE STEMS FOR INSTALLATION OF STRAIN GAUGES SPECIFICATION SPEC-M-009

~summar This evaluation provided the basis for the acceptability of using SPEC-M-009 in the maintenance process.FPL Specification, SPEC-M-009,"Machining of Motor Operated Valve Stems for Stain Gauge Installation" provided engineering guidance and details sufficient to allow field machining of threaded valve stem sections for installation of Teledyne miniature strain gauges.These strain gauges were provided in support of NRC Generic Letter, 89-10 concerning MOV actuator load monitoring.

By utilizing the specification in lieu of an engineering package greater flexibility in the implementation process resulted.This specification allowed all or part of the identified valve scope to be implemented, and additional valve scope could be added in the future by specification and corresponding calculation revisions, if desired.Safet Evaluation:

This evaluation concluded that the method of implementation and limitations imposed by SPEC-M-009 are consistent with all associated technical and licensing requirements.

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.

108 SAFETY EVALUATION aTPN PTN SEMS-93-010 REVISIONS 0&1 UNIT 4 APPROVAL DATES: Rev.0 03/25/93 Rev.1 04/08/93 INTAKE COOLING WATER VALVE REPLACEMENTS AND B HEADER CRAWL THROUGH INSPECTION

~summa The purpose of this safety evaluation was to demonstrate that there was no adverse effect on plant safety or operations associated with the replacement of eight Intake Cooling Water (ICW)valves and the crawl through inspection/repair of ICW piping.Eight Intake Cooling Water (ICW)isolation valves were replaced due to excessive leakage during the 1993 Unit 4 refueling outage.A crawl through inspection and repair of the Unit 4 B ICW header and the C ICW pump discharge piping was also performed during the same outage.Some of the valve replacements required one of the ICW headers to be removed from service.To ensure that Residual Heat Removal (RHR)and spent fuel pool cooling requirements continued to be met, ICW operations were controlled in accordance with the applicable Technical Specifications and system operating procedures.

Revision 1 of this safety evaluation added the replacement of valve 4-50-340 to the scope of this safety evaluation.

The additional scope was warranted based on the results of leak testing performed after the issuance of Revision 0.Safet Evaluation:

The ICW valve replacement and crawl through activities did'ot adversely affect the operation of equipment important to safety necessary to support any Mode of operations.

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.

109 SAFETY EVALUATION ZPN-PTN SEMS-93 011 REVISION 1 UNITS APPROVAL DATE 3&4 05/11/93 OMS SETPOINT DURING RCP OPERATION~mummar The purpose of this safety evaluation was to assess the ability of the Overpressure Mitigating System (OMS)to provide protection against the system design basis overpressurization events during RCP operation.

Recent industry findings on the methodology of determining Overpressure Mitigation System (OMS)setpoints prompted a review of the setpoints at Turkey Point.During the review, it was determined that the calculations for determining the setpoints did not consider pressure differences between the reactor vessel and the.pressure transmitters caused by RCP operation and elevation differences.

To ensure that in all cases no overpressure transient could occur, restrictions were imposed to either decrease the PORV stroke times or to limit RCP operations during cold, water solid operations.

These actions, in conjunction with ASME Code Case N-514 (which allows primary pressure to reach up to 1104 of the pressure/temperature limits during cold overpressure events), assured that the OMS was operable and capable of protecting the reactor vessel from damage from all postulated cold overpressure transients.

Safet Evaluation:

Reactor coolant pumps do not contribute to any accident mitigation analyses in Mode 5.A shorter PORV open stroke time does not adversely impact any previously postulated accident in Mode 5, and serves to mitigate those accidents addressed within the safety evaluation.

The actions and changes identified and evaluated in this safety evaluation did not have any adverse effect on plant safety or plant operations.

The actions and 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 and changes identified in this safety evaluation.

110 SAFETY EVALUATION aTPN-PTN-SEEP 93 0l.5 REVISION 0<<UNIT~4 APPROVAL DATE: 05/19/93 SAFETY EVALUATION ROR ACCEPTABLE UPPER AND LOWER TIME DELAY LIMITS FOR ECC 4A AND ECP 4A AGASTAT LOAD SE UENCING RELAYS~summa This safety evaluation provided acceptance criteria for ECC 4A and ECF 4A load block sequencer timing.During Engineered Safeguards Integrated Testing the two Agastat relays associated with sequencing the Emergency Containment Cooler (ECC)4A and Emergency Containment Filter (ECF)4A failed to meet the existing test acceptance criteria.The as-left setting for the ECF relay was 37.5 seconds.This safety evaluation established a basis for the as-left settings of the Agastat relays and provided acceptance criteria for future testing commensurate with equipment accuracies.

Safet Evaluation:

This safety evaluation demonstrated that the ECF and ECC fans will start and reach operating speeds within the time limits prescribed in the most limiting plant accident analyses.In addition, acceptance criteria for future testing was consistent with the most limiting design basis accident analyses.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.

111 SAPETY EVALUATION ZPN-PTN-SEMS93-017 REVISION 0 UNIT APPROVAL DATE 4 05/06/93 EVALUATION POR LOOSE OBJECTS IN THE SECONDARY SIDE OP STEAM GENERATOR C AT TURKEY POINT UNIT 4~8ammar This evaluation addressed the potential safety significance of operating Turkey Point Unit 4 Steam Generator'C', with loose objects (screws)present in the secondary side.These screws were described as three (3)round head screws, which attached an inspection camera light bracket to its camera housing.The bracket was located and retrieved from the tube lane (blowdown lane), however, thorough remote visual inspection of the tube lane did not reveal the screws.These screws were presumed to be in the tube lane and most probably below the blowdown pipe where visual contact could not be made.Their location was presumed to be between two (2)flow restrictor baffles located at Columns 72 and 79.Safet Evaluation:

Analysis showed that any potential tube wear from the screws would not occur beyond a depth equivalent to the current Technical Specification plugging limit of 404 wall loss.The screws were not expected to exit the steam generator and enter the Main Steam System and, therefore, will not impact any accident analysis that considers the Main Steam System.It was also determined that isolation of the Steam Generator Blowdown and Sampling System would not be adversely impacted by the screws.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.

112 8AFETY EVALUATION JPN-PTN-SEMS-93 018 REVISION 0 UNIT APPROVAL DATE 4 05/01/93 8AFETY EVALUATION FOR 8TEAM GENERATOR C SECONDARY SIDE FOREIGN OBJECTS~summa This evaluation addressed the effects of a foreign object identified on the tube sheet surface of the Unit 4'C'team Generator.

The object was a piece of wire approximately 3" long and 1/8" in diameter, and had been determined to be irretrievable.

Previous Eddy Current Test (ECT)data confirms that this object has remained lodged in the same location since the previous refueling outage.An ECT performed during this outage further shows that no damage has occurred to the tubes adjacent to the object due to it'presence.The purpose of this safety evaluation was to assess the acceptability of resuming Unit 4 operation with the foreign object remaining lodged in the C Steam Generator.

8afet Evaluation:

This safety evaluation determined that the object had been fixed in its present location for at least one full operating cycle and that no damage to the adjacent tubes had resulted.Based on this documented experience, future movement of the object was not expected.The generator would not be damaged by the foreign object during future operation.

However, continued monitoring of the object'would be performed to ensure that the conclusions of this safety evaluation remained valid during subsequent fuel cycles.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.

113

SAFETY EVALUATION i7PN PTN-SEMS 93 019 REVISION 0 UNIT APPROVAL DATE 4 05/01/93 SAFETY EVALUATION FOR STEAM GENERATOR A SECONDARY SIDE FOREIGN OBJECTS~summa r This evaluation addressed the potential safety impact of continued operation of the Turkey Point Unit 4 plant with a potentially mobile foreign object remaining in the secondary side of Steam Generator A.During a routine foreign object search and retrieval operation, a total of 4 foreign objects were detected.Three of the four identified objects were retrieved, and only one object remained.This object was described as a flat washer with a nut integral to the washer.In the worst case, foreign objects in the steam generator secondary side could cause significant tube wear, tube wear with primary to secondary leakage and possibly a potential tube rupture event.Safet Evaluation:

This evaluation demonstrated that operation of the steam generators with the identified foreign objects remaining in the steam generators would not have an adverse effect on the pressure boundary integrity of the steam generators.

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.

114

SECTION 3 RELOAD SAFETY EVALUATIONS 115

PLANT CHANGE MODIFICATION 91 108 UNIT TURN OVER DATE 3 12/18/92 TURKEY POINT UNIT 3 CYCLE 13 RELOAD SAFETY AND LICENS1NG CHECKLIST~Summa This engineering package provided the reload core design of the Turkey Point Unit 3 Cycle 13.This engineering design also extended the service life of the Hafnium Vessel Flux Depressor (HVFD)clusters to the end of Cycle 13.The primary design change to the core for Cyclh 13 was the replacement of 57 irradiated assemblies with 56 fresh Region 15 fuel assemblies and 1 irradiated assembly reinserted from Cycle 8.The fuel was arranged in a low leakage pattern with no significant differences between the Cycle 13 loading pattern and the Cycle 12 design.Cycle 13 also marked the elimination of secondary neutron sources in Turkey Point Unit 3.Region 15 used the same Debris Resistant Fuel Assemblies (DRFA)as Cycle 12, except for several minor design changes.Safet Evaluation:

The design of Turkey Point Unit, 3 Cycle 13 was evaluated by Westinghouse.

The Cycle 13 reload core design, including the reconstituted fuel assemblies, met all applicable design.criteria and all pertinent licensing basis.The minor design modifications to the fuel assembly and core components (WABA)did not affect the applicable design criteria for these components.

These changes had no impact on fuel rod performance, dimensional stability or core operating limits.The extension of the residence time of the HVFD rods likewise did not impact their performance or exceed their design criteria.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

116 PLANT CHANGE MODIFICATION 92-045 UNIT TURN OVER DATE 4 05/21/93 TURKEY POINT UNIT 4 CYCLE 14 RELOAD SAFETY AND LICENSING CHECKLIST~mamma This engineering package provided the reload core design of the Turkey Point Unit 4 Cycle 14.The primary design change to the core for Cycle 14 was the replacement of 52 irradiated assemblies with 52 fresh Region 16 fuel assemblies.

These fresh assemblies were Debris Resistant Fuel Asseqblies (DRFA)and all contain a 6-inch axial blanket of.71 w/o U (natural uranium)at both the top and bottom of the fuel stack.This was the first use of axial blankets in Unit 4.The fuel was arranged in a low leakage pattern with no significant differences between the Cycle 14 loading pattern and the Cycle 13 design.Region 16 used the same Debris Resistant Fuel Assemblies (DRFA)as Cycle 13, except for the following design changes which included: 1)the use of axial blankets;2)implementation of an anti-snag outer grid strap in the top and bottom inconel grids;and 3)a change to the Wet Annular Burnable Absorber (WABA)rodlet spacer length.The spacer within the WABA rodlet assembly was lengthened to shift the WABA absorber stack upward to align the center of the absorber stack with the fuel midplate.Safet Evaluation:

The design of Turkey Point Unit 4 Cycle 14 was evaluated by Westinghouse.

The Cycle 14 reload core design met all applicable design criteria and all pertinent licensing bases.The minor design modifications to the fuel assembly and core components (WABA)did not affect the applicable design criteria for these components.

These changes had no impact on fuel rod performance, dimensional stability or core operating limits.The extension of the residence time of the HVFD rods likewise did not impact their performance or exceed their design criteria.The modification in this Engineering Package did not have any adverse effect on plant safety or plant operations.

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.

117 SECTZON 4 ANNUAL REPORT OP POWER OPERATED RELZEP VALVE{PORV)ACTUATZONS 118 ANNUAL REPORT OF SAFETY AND RELIEF VALVE CHALLENGES By letter dated June 18, 1980 (L-80-186)

Florida Power and Light stated the intent to comply with the requirements of item II.K.3.3 of enclosure 3 to the commissioner's letter of May 7, 1980 (Five Additional TMI-2 Related Requirements to Operating Reactors).

The following is a list of power operated relief valve (PORV)challenges for Turkey Point Units 3 and 4 from July 1, 1992 to June 1, 1993.Unit 3 November 27, 1992 January 16, 1993 With the unit in Mode 6, PCV-3-456 actuated twice due to high RCS pressure which occurred as a result of starting the 3C Reactor Coolant Pump.A delay in stopping the charging while the pressurizer was being filled resulted in an RCS pressure increase which caused an actuation of PORV PCV-3-456.

UNIT 4 September 19, 1992 October 5, 1992 With Unit 4 in Mode 3, PORV PCV-4-456 opened during testing of valves MOV-4-750 and MOV-4-751.

With Unit 4 in Mode 5, PORV PCV-4-455C opened during a surveillance of the Overpressure Mitigating System, resulting in slight depressurization of the Reactor Coolant System.119

SECTION 5 STEAM GENERATOR TUBE INSPECTIONS FOR TURKEY POINT 120 0 0 Eddy Current Summary of Results Plant: Turkey Point 3 Examination Dates: 10/10/92 Through 10/18/92 Steam Generator Number 3E210A 3E210B 3E210C Total Tubes Inspected 3199 3200 3188 Total Ind.204 to 394 72 95 73 Total Ind.40<to 100<NONE Total Tubes Plugged as Preventive Maintenance NONE Total Tubes Plugged Location of Indications Steam Generator AVB Bars Cold Leg Hot Leg Drilled Support 1 through 6 Cold Leg Hot Leg Top of Tube Sheet to 1 Drilled Support 3E210A 3E210B 3E210C 37 44 59 13 22 14 17 NONE 12 Certification of Record We certify that the statements in this record are correct and the tubes inspected were tested in accordance with the requirements of Section XI of the ASME Code.FLORIDA POWER and LIGHT COMPANY Organization Date: Date: i2/8/~Prepared By: Reviewed By: S/G E dy Current Coordinator I pections Su ervisor 121 Form NZS-BB Owners'eport for Eddy Current Examination Results Page 2 of 2 Steam Generator Tubes Plugged Plant: Turkey Point 3 Steam Generator 3E210A Steam Generator 3E210B Steam Generator 3E210C Row Column Remarks Row Column Remarks Row Column Remarks 32 444 42 45 PTP 33 35 20 39 41 55 66 70 PTP PTP 60%41%PTP-Preventive tube plug based on memos attached to CNR 5 92-3-031 and 032.The following cumulative listings are attached for each steam genexator.

~Cumulative Distribution Summary~Pluggable Indications

~Indications20-394 122 COMPONENT: S/G A CUMMULATIVE DISTRIBUTION

SUMMARY

TURKEY POINT UNIT g 3 09/92 Page: 1 of 1 Date: 11/16/92 Time: 11:30 AM Examination Dates: 10/10/92 thru 10/18/92 Total Number of Tubes Inspected.....: 3199 Total Indications Between 204 and 394 Greater than or ecpxal to 404 Total Tubes Plugged as Preventive Maint Total Tubes Plugged 72 1 Location Of Indications 20%to 100%Hot Leg TSH-.5 to 01H-2.1 01H-2.0 to 06H+2.0 06H+2.1 to AV1-3.1 AV1-3.0 to AV4-3.0 Cold Leg 4 TSC-.5 to 01C-2.1 14 01C-2.0 to 06C+2.0: 13 21 06C+2.1 to AV4-3.1: 3 13 123 CUMULATIVE EXAMI HAT ION REPORT PTN-3 OUTAGE: 09/92 COMPOHENT: S/G A DESCRIPT ION: PLUGGABLE IND I CAT IOHS Page: 1 of 1 Date: 11/13/92 Time: 9:00 AM I I Extent IRowlCollLeglReqlTst/Hotel Reel I Probe'-I.--I---I-I I.I.I 21I 32I C ITEHITEH PSIAC012-03IA-720-M/ULC

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20K TO 39K Page Date Time 1of2:>>/13/92: 9:00 AM I Lo I Note/Reel I Extent LeglReqlTst/

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I IRowlcoll I-I I 21 41 I 91 41 sl I I I I 91 61 I 141 71 I 91>>l 91 1zl I 91 141 I 101 161 91 181 I 91 201 91 z11 t I 91 231 I 221 301 1231>>I 21 321 I 171 3zl I 181 371 I 191 371 I 211 381 I 291 4ol I I I I>>1441 I 171 451 I 381 4sl I I I I 411 461 I 241 47I I 371 471 1221>>I I zzl szl I I I I 3ol szl I zl 541 I 151 551 I 301 581'-l---l-c Io6clo6c C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH c ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH C ITEHITEH c Io6clo6c H ITECITEC H ITECITEC H ITECITEC H ITECITEC'ITEHIAV3 C ITEHITEH C ITEKITEH C ITEHITEH C ITEHITEH H ITECITEC C ITEHITEH C ITEHITEH H ITECITEC H ITECITEC C ITEHITEH H ITECITEC H ITECITEC H ITECITEc H ITECITEC c Io6clo6c H ITECITEC H ITECITEC H ITECITEC H ITECITEC--I-I"--Probe----I-I-I-IAC001-01IA-720-M/ULC lose Pslac004-01la-720-M/ULC Io6H IAC006-021A-720 M/ULC ITSH IAH030-07IMRPC720-3C/7ITSH Iac004-011A-720-M/ULc 106H PCIAC006-02IA 72 M/ULC ITSH IAC004-01IA-720 M/ULC 106H IAC005-01IA-720-M/ULC 106H PSIAC005-011A-720-M/ULC 106H PCIAC007-021A-720-M/ULC 102C IACOO 01IA 720 M/ULC 106H IAC005-01IA-720.M/ULC 102C IAC005-01IA-720.M/ULC 106H IAC005-01IA-720.M/ULC 106H IAC005-01IA-720-M/ULC 106H 106H SSIAC006-02IA-720-M/ULC PCIAC012-031A-720-M/ULC 104H SSIAC012-03IA-720-M/ULC 101H PSIAC001-01IA-720-M/ULC lose PCIAH028-06IA-720-M/ULC 101C PSIAH028.06IA-720-M/ULC 101H IAH029-07IA-720-M/ULC 101C IAH029-07IA-720-M/ULC 101C IAH032-08IMRPC680.5FH IAV2 PslAc014-04la-720-M/ULC 102H IAV1 PCIAC018-051A-720-M/ULC PCIAC018-051A-720-M/ULC IAV3 PslAc019-051A-720-M/ULc lav3 IAH029-07IA-720-M/ULC 102H PSIAC019-05IA-720-M/ULC IAV2 P IAC019 05IA 720 M/ULC IAV3 PSIAH033.081A.720.M/ULC 106H PSIAH017-031A-720-M/ULC IAV1 PSIAC019-05IA-720-M/ULC IAV3 PSIAH027-06IA-?20-M/ULC 106C IAH018-03IA-720-M/ULC 104H IAH030-07IMRPC720-3C/7104H P IAH018 03IA 720 M/ULC IAV3 PSIAC002-011A-720.M/ULC 101C ITsC PSIAH025-051A-720.M/ULC IAV1 IAH019.04IA-720-M/ULC IAH019 0 IA 72 M/ULC IAV2 IAH019-04IA-720.M/ULC IAV3 09/92 cation IVoltslDeglch I I IDiffl I-I---I--I-

.511491 11 301~611511 11 301 1.111571 11 261 1.411311 11 321.411501 11 311.211szl~711491 11 321 1.211601'll 221.611481 11 331.711401 11 391 F 811521 11 301.911461 11 351 811581 11 241 1.111511 11 301 1.011521 11 301.811481 11 341.611621 11 211.811601 11 231.511581 11 221.s11541oool 291.411511 11 311.711581 11 251 911601 11 231 2.211471 11 251 611601 11 231.81 IP 21 251.sl IP zl 231.sl IP zl zsl 1.011551 11 281 IP zl z61.sl II zl as/.611501 11 321 IP zl 291 44.1 2.4 3.6 3.9 2.5 2.7 2.5 2.4 2.4 21.5 2.3 23.9 2.4 2.3 2.2 2.3 28.8 12.9 43.1 10.6 24.4 3.1 3.1 12.5 18.8.0.0.0 3.9.0.0 5.4.0.0 3.5 40.9 42.4.0 3.3 8.3.0.0.0 IP zl 241.411441 11 371.911411 11 391.91>>61 11 391 Ii 21 231.411631 11 211 1.211491 11 341 1.sl II 21 3ol 1.ol II 21 271 1.sl li zl 3ol N/A Location I IvoltslDeglCh I X I I I-I--I--'I I I-I--I--'25 0 J CUMULATIVE EXAMINATION REPORT PTH-3 OUTAGE: 09/92 CDHPONENT DESCRIPTION S/G A: 20X TO 39X Page Date Time 2'of 2: 11/13/92: 9:00 AM I IRoMlcollLeg

+--I.-I---I I IH I 28I 59I H I I IH I 29I 59I H I I I>>I 41I 59I H I 38I 6sl c 9I 6rl H I zol 68I H I 19I 69I H I zrl rol H I 30I 71I H 1I 73I c I 32I rsl H I I IH 1I 81I c I 22I 81I H 9I Bzl H I 24I 82I H 1I 84I c I 19I 84I H I 9I 85I H 9I 86I H I 9I 90I H I 1zl 90I H'-I I---I Extent IReqlTst/H I-.I.ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITEHITEH ITECITEC ITEGITEC ITECITEC ITECITEC ITECITEC I06cl06c ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC I06cl06c ITECITEC ITECITEC ITECITEC ITECITEC I06cl06c ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC I-" I I I otel Reel I'robe I Lo I I-I-IAH019-04IA-720.M/ULC IAV4 IAH019-04IA-720.M/ULC IAV2 IAH019-04IA-720.M/ULC IAV3 IAH019-04IA-720.M/ULC IOSH IAH019-04IA-720.M/ULC lose PSIAH034-OBIA-720-M/ULC I06H SCIAC022-05IA-720-M/ULC IAV2 PSIAH009-02IA-720-SF/RM ITSC PSIAH027-06IA-720-M/ULC I04H PSIAH027-06IA-720-M/ULC I03C IAHO 1 IA 720 M/ULC l01H PCIAH 21 04IA 720 M/ULC l04C IAC003-01IA-720-M/ULC IBAC IAH 0 IA 720 M/ULC l04H IAH011-01IA.720.M/ULC I03H IAH023-05IA-720-M/ULC IAV3 SSIAH011-01IA-720.M/ULC I03C IAH01'I-01IA-720-M/ULC I02C IAc003-01IA-720-M/ULC IBAc IAH022 0 IA 720 M/ULC IOSH IAH012-02IA.720-M/ULC I06H SCIAH022-OSIA-720-M/ULC l06C SSIAH012-02IA-720.M/ULC I06H PSIAC003-01IA.720-M/ULC IBAC IAH015-03IA 720-M/ULC ITSH SSIAH012-02IA.720-M/ULC I06H IAH012-02IA.720.M/ULC I06H IAH012-02IA-720-M/ULC I06H PSIAH016-03IA-720-M/ULC I04H---I-I-I-cation.0.0.0 25.7 36.4 4.6.0 46.0 42.2 41.0 45.8 2.8 13.6 36.8 9.1 2.1 43.1 16.7 16.0 40.5 2.9 4.1 2.6 16.6.9 2.4 2.3 2.4-.5 09/92 I IvoltslDeglch I I I I I.rl li zl.rl li zl IP zl 1.2I16ol 1.3I162I 1I.5I148I 1I.9I149I 1I.3I159I 1I 1.0I159I 1I.3I1451 11.sl14rl.SI148I'll 1.6I14sl.8I143IP 2I.4I143I 11 1.0I144I 1I~5 I 155 I 1 I.3I>>8l>>l.9I146I 1.1I158I 1I 1.1I14s I I.zlisrl.4 I 159 I 1 I.9I144 I 1 I 1.7I162I 1.2I159I 1I~1I124II 1I I I I I N/A X IDiffl Location I IvoltslDeglCh I X I 33I 37I 34l 34I 35I zsl 38I 37I zrl 24I 35I zsl 36I zsl 22I 37I zol 3zl I I I I I I-I------------

I-----I-" I--I--'I----------I----I---I---I--+I zsl I I I I I 24I 24I 23 I 21 I 34I 23 I 3zl Nunber of RECORDS Selected from Current Outage Number of TUBES Selected from Current Outage: 72 62 126

COMPONENT: S/G B CUMMULATIVE DISTRIBUTION

SUMMARY

TURKEY POINT UNIT g 3 09/92 Page: 1 of 1 Date: 11/16/92 Time: 11:30 AM Examination Dates: 10/10/92 thru 10/18/92 Total Number of Tubes Inspected 3200 Total Indications Between 20%and 394 Greater than or equal'to 404 Total Tubes Plugged as Preventive Maint Total Tubes Plugged 95 0 Location Of Indications 20%to 100%Hot Leg TSH-.5 to 01H-2.1 01H-2.0 to 06H+2.0 06H+2.1 to AV1-3.1 AV1-3.0 to AV4-3.0 Cold Leg 12 TSC-.5 to 01C-2.1: 0 17 01C-2.0 to 06C+2.0: 22 13 06C+2.1 to AV4-3.1: 10 127 CUHULAT I VE EXAHI HAT ION REPORT PTH.3 OUTAGE: 09/92 COHPONEHT: S/G 8 DESCRIPTIOH

PLUGGABLE INDICATIONS Page: 1 of 1 Date: 11/13/92 Time: 9:00 AH I I Extent I 09/92 I IRoMICollLeglReqlTst/Hotel Reel I Probe I Location IVoltslDeglCh I+I I I I I I--I--------I-I-I I I 42I 45I C ITEHITEH*H IBC025.07IA-720.H/ULC IAV2.Ol 2.2I I35XI I--I---I".I I I-I-------------I-I I-I I H/A I X, IDiffl Location IVoltslgeglch I I I I-I-----------I-----I--I--I---+PTPI I I I I I I-I--I----------.

-I-----I---I--I---'unber of RECORDS Selected from Current Outage: 1 Number of TUBES Selected from Current Outage: 128

CUNULAT I VE EXAMINATION REPORT PTH-3 OUTAGE: 09/92 COMPONENT DESCRIPTIOH S/G 8: 20K TO 39K Page Date Time 1of3: 11/13/92: 9:00 AN Ol I Extent IReq(Tst/I-I ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH (TEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITEHITEH ITECITEC ITEHITEH ITECITEC ITECITEC ITECITEC ITEHITEH ITEHITEH ITECITEC ITEHITEH ITEHITEH I-I Roe(Col(Leg

---I--I---

111 21 c 1o(s(c as)6(c 101 71 c 51 81 c Ic 8(8(c 111 91 C z3(9(c I c I c 11(1o(c 23(1o(c 1O(11(C 111 111 c 181 121 C 191 121 c 29(>>I c 201 131 c 6(14(c 331 151 c 36(19(C z6(zo(c 371 201 c 33(21(c 331 231 c 4o(zs(c 331 261 c 401 261 C 40(z7(c 281 281 C 33(29(C 391 301 c 111 311 H 34(31(c 6(32(H 51 341 H 61 341 H 32(34(C Ic SI 351 H 6(36(c 441 361 c I.I.--I Note(Reel I Probe I-I-PS(BC 08 03(A 72 SF/RN IBC006 02(A 72 M/ULC PCIBC011-03IA-720-N/ULC IBC006-02(A-720-N/ULC (BC006-02(A-720-M/ULC IBC006-02IA-720-N/ULC (BC006-02(A-720-N/ULC PCIBC009-03IA-720-SF/RM (BC011-03(A-720.N/ULC IBC011-03IA-720-N/ULC IBC011 03(A-720-M/ULC PSIBC009-03IA-720-SF/RN (BC011-03(A-720.N/ULC PS(BC006-02(A-720.M/ULC SSIBC009.03IA-720-SF/RN Pc(BC009.03(A-720-SF/RN PSIBC009.03IA-720-SF/RN PS(BC011-03(A-720-N/ULC pc(Bc009-03(a-720-sF/RN SS(BC006.02(a-720.M/ULC PS(BC016-05(A-720.M/ULC Ps(BG016-05(a-7ZO-N/ULc PSIBC012-04IA.720.M/ULC PSIBC016-05IA-720.M/ULC SSIBC016-05(A 72 N/ULC SSIBC016.05IA-720.N/ULC IBC017-05IA-720.M/ULC PCIBC017-05IA-720 M/ULC PS(BC017-05(A-720.M/ULC PS(BC017-05(A-720.N/ULC PS(BC013-04(A-720.N/ULC PSIBC017-05IA-720.N/ULC SS(BC017-05(A-720 N/ULC PSIBH005-01IA-720.N/ULC PS(BC017-05(A-720.N/ULC (BH025-07(A-720-M/ULC IBH025-07IA-720.N/ULC PSIBH025-07IA-720.N/ULC SSIBC017-05IA-720-N/ULC SSIBC017-05IA-720-N/ULC IBH025-07(A-720-N/ULC PCIBC027-08IA-720.N/ULC SSIBC025-07IA-720-N/ULC I I Iozc Iozc 105H Iozc IgaH (04H 101H (ozc 106H lose lose Iozc (o6H Iozc (O2C 105H ITSH IAV1 (04H 102H IAV2 (avz (av4 ITSH 106H 106H (oSc 106H Iav3 (avz 103C 106}I 102C (ozc (avz ITSH (TSH ITSH Iav3 Iav4 ITSH ITSH IAY1 I 37.0 31.8 34.3 31.7 15.9 15.0~7 36.6 4.0 31.2 18.5 36.7 4.5 31.9 36.7 19.8~7.0 15.4 32.2.0.0.0~6 4.8 5.0 39.3 5.1.0.0 34.2 5.0 45.7 37.2.0 39.0 31.6 38.4.0.0 33.5 37.5.0 I Location I 09/92 I (Volts(Oeg(Cb I-I-I I.S(147(.811541 11.3(13SI.611551 11.7(156(1(1.811541 11.6(123(i 1(.411411 11 1.411431 11.711541 11 31140(11.611411 11 1.311441 11.3(14S(.711571 11.S(141(.6(137((I z(.311481 11.911441 11.S((I 2(~S(Ii 2(.s(Ii 2(1.S(<<S(1.011541 11.711361 11.511451 11 6(1411 11 (I 2(.s((I z(1.o(13S(.511391 11~51154(1(.311511 11~4(Ii 2(.811461 11.611511 11.4(137(.8(.61 IP 31.S(140(.7(136(1.O(-I-I I I I (Diff(I I 30(221 3S(z1(zo(22(30(32(221 361 361 33(30(zo(361 39(Z3(z9(311 21(z1(3o(22(38(301 33(Z1(23(3S(35(21(26(221 32(27(2Z(zo(38(Z1(I N/A Location I (Volts(oeg(ch I X I I I I I 129 CUHULATIVE EXAMINATION REPORT PTN-3 OUTAGE: 09/92 COMPONENT: S/G B DESCRIPTIOH

20K TO 39K Page: 2 of 3 Date: 11/13/92 Time: 9:00 AN I IR+I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I Extent ow(coIILeglReqlTst/

I I I I 141 371 H ITEC ITEC 42(37(C (TEHITEH 44(371 C ITEHITEH 341 381 C ITEHITEH I I c ITEHITEH 421 381 C ITEHITEH I I c ITEHITEH 11(39(H (TEC(TEC 391 391 C ITEHITEH 441 401 C ITEHITEH 14(421 H ITECITEC 44(42(C (TEH(TEH 61 441 C ITEHITSH 191 441 H ITECITEC 421 451 c ITEHITEH I I c ITEHITEH 341 461 C ITEHITEH (C (TEN(TEH 35(48(C (TEH(TEH I C (TEH(TEH 61 491 C ITEHITSH 261 491 H ITECITEC 451 491 c ITEHITEH 171 501 H ITECITEC 341 51(C ITEHITEH I I c ITEHITEH 341 53(C ITEHITEH (C ITEM(TEN I I c ITEHITEH 261 541 H ITEGITEC 29(55(H (TEC(TEC 421 551 C ITEHITEH I I c ITEHITEH 441 I c ITEHITEC 431 601 c ITEHITEc 351 661 c ITEHITEH 401 661 c ITEHITEH 9(69(c ITEKITEc 271 701 H ITEGITEC 291 701 M ITEGITEC 251 711 H ITECITEC I I H ITEGITEc 11(721 H ITECITEC--.I.--I--

I I I Note(Reel (Probe I I PSIBH005-01IA-720.H/ULC PSIBC025-07(A-720.H/ULC PS(BC025.07(A-720-H/ULC IBC019 06(A 720 M/ULC SSIBC019-06(A-720-N/ULC (BC025-07(A-720.H/ULC IBC025-07(A-720-H/ULC PSIBM006-02IA-720-H/ULC IBC019-06(A.720-N/ULC PSIBC025-07IA-720-M/ULC PS(BH006-02(A-720-H/ULC PSIBC025-07(A-720.H/ULC PSIBC027-08IA.720.H/ULC SSIBK006.02IA-720-M/ULC IBC025-07IA-720-H/ULC IBC025-07IA-720-H/ULC PCIBC020-06IA-720-M/ULC Pc(BC020-06(A 720 N/ULC PSIBC021-06IA-720.M/ULC PSIBC021-06IA 720-M/ULC PSIBC028-08(A-720-H/ULC (BH027-08(A-720.H/ULC IBC026-07(A-720 H/ULC PSIBH007-02IA.720-N/ULC PSIBC021-06IA-720-N/ULC PSIBC021-06IA.720-M/ULC PSIBC021.06IA-720-H/ULC PSIBC021-06IA-720.H/ULC PSIBC021-06IA 720-H/ULC PSIBH017-OSIA-720 M/ULC PSIBH017-OS(A-720-M/ULC PS(BC 6 0 IA 72 H/ULC (BC026-07(A-720.M/ULC sc(BM034-08(a-720-H/ULC PSIBK034-08IA-720.H/ULC SSIBC029.08IA-720-H/ULC PSIBC024 IA 720-H/ULC PSIBM034-08(A-720.M/ULC SCIBK019-05IA.720.M/ULC PS(BH019-05(A.720-M/ULC SSIBH019-OSIA 720-M/ULC PSIBH019-05IA-720-H/ULC PSIBH009-03IA.720-H/ULC I-I-102C ITSH lav4 (AV2 (av3 ITSH ITSH 101H IOSH Iavi 103M IAV1 ITSH (02K IAV2 (AV3 IAV2 IAV3 IAV2 (av3 (02C (02C (AV4 103H (av2 IAV3 IAV1 (av2 (AV3 103C (01C IAV2 IAV3 IAV4 (AV4 (06c lav4 IAV2 106H IOSH 106H (04C 102H I 46.5~8.0.0.0 1.4 3.2 12.0.8.0 21.1.0 39.3 32.0.0.0.0.0.0~0 17.4 15.7.0 25.0.0.0.0.0.0 40.3 18.1.0.0.0.0.5.0.0 4.8 33.1,3.0 37.5 36.6 Location 09/92 I I (Volts(Deglch (X (Diff(-I----I---I---I---I---.I--

.2(151(1(26(1.0(138(1(37(~41 I>>I 231 2.2(1.8(IP 3(30(1'11501 11 251~5(149(1(26(~5(144(1(32(1.51111IP 11 391.5((P 2(24(.511561 11 211.6(II 3(20(1.6(142(1(32(.81139(1(36(2.2(II 2(35(.9((r 2(28(~9(IP 2(281 1.5((P 2(31(.5((P 2(26(~711411 11 34(.4(135(.5(IP 2(22(.4(147(1(28(Ii 2(30(.4(IP 2(26(lr 2(27(Ii 2(26(.5((I 2(26(.4(147(1(28(.5(155(1(20(.4((I 2(21(1.0(IP 2(26(1.2((I 3(24(.8(IP 2(271~6(124(P 1(27(71 I>>I>>I.6((P 2(26(.411421 11 361.3(136(1(37(~6(137(1(39(.5(144(1(30(.611541 11 221 I-I---I-.I-N/A Location I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I-I I (volts(Deg(ch I X (I I I I I I I I I I I I I 130 0

CUHULATIVE EXAMINATION REPORT PTH-3 OUTACE: 09/92 COHPONEHT: S/6 B DESCRIPTIOH:

20K TO 39K Page: 3 of 3 Date: 11/13/92 Time: 9:00 AM---I---------

I PCIBH021-06IA-720-M/ULC PCIBM021-06IA-720.M/ULC PSIBH020.06IA.720.M/ULC PCIBM021-06IA 720 M/ULC PCIBH021-06IA-720.M/ULC IBH013.04IA-720.M/ULC PSIBH013-04IA-720-M/ULC PSIBH014-04IA-720-M/ULC PSIBH015-04IA-720.M/ULC I I I06C 5.4 I06c 5.6 I04C 32.2 I06C 5.5 I02c I02H 36.0 I01H 10.9 I02H 36.5 I02H 36.1 I I 341 74I H ITECITEC I 35I 74I H ITEGITEc I 28I 75I H ITECITEC I 34I 75I N ITECITEc I 35I 75I H ITECITEC I 11I 76I H ITEcITEC I 11I 78I H ITEGITEc I 11I 85I H ITEGITEC I 11I 89I H ITEcITEc+I I I I I I Extent I IRowICol ILegIReqITst/NoteI Reel I Probe I Location I I I I I 09/92 N/A I IVoltsIDegIch I X IDiffI Location IVoltsIDegICh I X I-I----I.--I-.I--I----I------------

I-----I--I--I--'.2I155I 1I 20I I I I I I.7I149I 1I 26I I I I I I I.4I151I 1I 26I I I I I I I.7I152I 1I 23I I I I I I I~5 I 140 I 1 I 34 I I I I I I I.4I141I 1I 38I I I I I I I 4I156I 1I 23l I I I I I I I.8I155 I 1I 22 I I I I I I I.5I138I 1I 34I I I I I I-I----I--I-.I-I I I-I---I---I--'wher of RECORDS Selected from Current Outage: 95 Number of TUBES Selected from Current Outage: 81 131 COMPONENT: S/G C CUMMULATIVE DISTRIBUTION

SUMMARY

TURKEY POINT UNIT I 3 09/92 Page: 1 of 1 Date: ll/16/92 Time: 11:30 AM Examination Dates: 10/10/92 thru 10/18/92 Total Number of Tubes Inspected.....: 3188 Total Indications Between 204 and 394 Greater than or equal to 404 Total Tubes Plugged as Preventive Maint Total Tubes Plugged 73 3 Location Of Indications 204 to 100%Hot Leg TSH-.5 to 01H-2.1 01H-2.0 to 06H+2.0 06H+2.1 to AV1-3.1 AV1-3.0 to AV4-3.0 Cold Leg 2 TSC-.5 to 01C-2.1: 3 6 01C-2.0 to 06C+2.0: 6 9 06C+2.1 to AV4-3.1: 13 37 132 CINULAT I VE EXANI HAT ION REPORT PTN-3 OUTAGE: 09/92 CONPOHENT: S/G C DESCRIPTION:

PLUGGABLE INDICATIONS Page: 1 of 1 Date: 11/13/92 Time: 9:00 PM I I Extent IRouIColILegIReqITst/NoteI Reel I Probe+.--I---I---I-I I I-----------

I 33I 39I C ITEHITEH*H ICC024-05IA-720.N/ULC I 3 I 1I C ITEHITEH N ICC024 0 IA 720 N/ULC 2I 55I C I06CI06C PCICC003-01IA-720-N/ULC I 20I 66I H ITECITEC PCICH015-03IA-720.N/ULC I 2I 70I C I06CI06C PCICC003-01IA-720.N/ULC I I I I-I"-I I 09/92 I I Location IVoltsIDegICh I I IDiffI I.".---------

I-----I---I-" I-I I IAv1.DI 2.3I I35XIPTPI IAY1.OI 2.4I I35XIPTPI ITsc 24.1 I.3I105I 1I 60I I I06c 2.4 I.6I134I 1I 41 I I I02c.7 I.5I107I 1I 45 I I--------I---.-I.-.I.-I-I I N/A I Location IVoltsIDegICh I 7 I I-I--I--I---'I I I I I I I I I I I I I I I I I I I I I I-I-----I--I----'umber of RECORDS Selected from Current Outage Hunber of TUBES Selected from Current Outage: 5 133

CUMULATIVE EXAMINATION REPORT PTN.3 OUTAGE: 09/92 COMPONENT DESCRIPTION S/G C: 20K TO 39K Page Date T iae 1of2: 11/13/92: 9:00 AM I IRog(+---I I 3(6(I 26(I>>I I 37(I 301 I 391 (30(I 4(I 431 I 4(I 35(I 3s(I I 8 I I I I I I I 34(I I I I I 351 I I I I I I I 33(I I I 3s(I I I I I 34((3s(I I I I I 231 I 3s(I I I 30(+---I-I Extent Col(Leg(Req(Tst/

--I---I-I 10(H (06C(06C 121 C ITEHITEH 151 C ITEHITEH 191 C ITEHITEH 28(C ITEKITEH 30(C ITEHITEH 301 C ITEHITEH 311 C ITEHITEH 31(C ITEN(TEN 331 H ITECITEC 33(C (TEH(06C 341 H ITECITEC 351 C ITEHITEH 361 C ITEHITEH I c ITEHITEH 391 C ITEHITEH I c ITEHITEH 391 C ITEHITEH I c ITEHITEH I C ITEHITEH 411 C ITEHITEH I c ITEHITEH I c ITEHITEH 411 C ITEHITEH I C ITEHITEH I c ITEHITEK I C ITEHITEH 431 C ITEHITEH I C ITEHITEH 431 C ITEHITEH I C ITEHITEH I C ITEHITEH 441 H ITECITEH 441 C ITEHITEH 441 C ITEHITEH I c ITEHITEH (C (TEH(TEH 45(C ITEHITEH 45(C ITEHITEH I C ITEHITEH I c ITEHITEH 461 C ITEHITsH 461 C ITEHITEH-I-.-I---I---

I Note/Reel'Probe I-I-PS ICH001-01 I a-700-SF/RM PSICC006-01IA-720.M/ULC SSICC017-03IA-720 M/ULC SCICC017 03IA-720 M/ULC SSICC023 05(A-720 M/ULC SSICC018 03(A 720 M/ULC S ICC023 0 IA 720 M/ULC PSICC018-03IA-720.M/ULC SSICC023 05(A 720 M/ULC ICH 9 0 (A 72 M/ULC SS(CH034-09(A-720.M/ULC SSICH029-07(A.720.M/ULC PSICC 23 IA 72 M/ULC ICC024-05IA-720.M/ULC (CC024-05(A-720.M/ULC PSICC026-06(A 72 M/ULC PSICC026-06IA.720.M/ULC ICC024-05IA-720.M/ULC ICC024-05(A-720-M/ULC ICC024-05IA-720.M/ULC ICC024-05IA-720-M/ULC (CC024-05(A-720.M/ULC ICC024-05(A-720.M/ULC (CC024-05(A-720 M/ULC (CC024-05(A-720.M/ULC ICC024-05(A-720-M/ULC ICC024-05IA-720.M/ULC PSICC025-05IA-720-M/ULC PSICC 5 0 IA 72 M/ULC PSICC025-05IA-720.M/ULC PSICC IA 72 M/ULC PS(CC025-05(A-720.M/ULC PS(CC031-06(A-720 M/ULC SSICC025-05IA-720.M/ULC PSICC025-OSIA-720-M/ULC PSICC 25 0 IA 72 M/ULC PS ICC025-05 (A-720.M/ULC (CC020-04(A-720.M/ULC PSICC025-05IA-720-M/ULC Ps(CCDZS-05(a-720-M/ULC PS(CC025-05(A-720.M/ULC PS(CC026.06(A-720.M/ULC ICC020-04IA-720.M/ULC

.---I-I I LocatIon I (DZH (ozc (DSH (AV4 (AV4 IAV4 IAV2 IAv3 (av3 ITSC IAV3 (TSC (av3 IAV2 IAV3 103C 103C (AV1 IAV2 Iav3 IAV1 Iav3 (av4 IAV1 IAV2 (AV4 Iavz IAV3 IAvz IAV3 IAV4 102H Iav3 IAV2 (av3 (AV4 Iav3 IAY2 IAV3 IAV4 ITSH IAV1 I-34.9 11.1 44.7.3.0.0 18.8.0.0 27.2.0 27.3.0.0.0 48.3 11.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0.0 50.4.0.0.0.0.1.0.0.0 28.4~1 09/92 I I N/A (VoIts(Deg(Ch (X IDIff(Location I I I I I I.911421 11 341.6(142(1(31(.811411 11 331 (P 2(23((I 3(z4(.s((P 3(24(1.311521 11 251 (P 2(zz((I 31 z4(.511421 11 331 II 2(23(.711471 11 271.sl II zl 23(.s((I 2(23(.s(II z(23(.811341 11 391.7(14s(1(30(z.3((I z(3s(.6(II 2(24(II 2(26(IP 2(zr(sl IP zl 231 1.0(IP 2(26(z.4((p 2(3s((p 2(zr(IP z(24((I Z(21(.s((I z(22(.s(II z(23(II 2(z6((r 2(29((p 2(27(~711481 11 291.S((I 3(23(IP 2(30(1.s(II 2(30(.8(IP 2(zs(.6(IP 2(24(II 2(33(II 2(24(.s((p 2(zz(.7(152(1(24(1.6((I 2(32(I I I I I (Volts(Deg(Ch (X (I I I I+134

CUMULATIVE EXAMINATION REPORT PTN-3 OUTAGE: 09/92 COMPONENT DESCRIPTION S/G C: 20X TO 39X Page: 2 of 2 Date: 11/13/92 Time: 9:00 AM I IRoM(Col(Leg

'--I--I--I I Ic I C 141(46(H I 301 481 K I I IH I 351 491 H I 43(53(H I 391 541 H I 22(SSI H I 26(SBI H I 331 581 H I 301 611 H I 32(62(H I 38(62(H I 24(63(H I I IH I 201 64(H I 301 64(H I 32(641 H I 381 651 H I I IH I I IH I 381 711 H (3SI 72(H (32(75(H 1(76(c I sl 871 H I 31881 H'I.I---I Extent IReqlTst/I-I ITEHITEH ITEHITEH ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC'TECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC (06c(06c ITEC(TEC ITECITEC 106c(06c I-I I Note(Reel I Probe I-I.(cco20-04(a-720-M/ULC ICC020.04IA-720-M/ULC ICH034-09IA-720-M/ULC PSICH018-04(A.720-SF/RK PSICH018-04(A.720-SF/RK PSICH030-07IA.720-M/ULC ICH033-08IA-720.M/ULC ICH031-07(A-720-M/ULC SSICH022-05(A-720.M/ULC PSICH022-05IA-720-M/ULC PCICH032-OBIA-720-M/ULC PSICH022-05IA-720.M/ULC PCICH032-08(A-720.M/ULC PCICH032-08(A 720 M/ULC SCICH023 0 IA 720 SF/RM SCICH023-Os(A-720-SF/RK PCICH009-02(A-720-SF/RM ICH014-03IA-720.M/ULC SCICH 23 0 IA 720 SF/RM ICH032-08IA-720.H/ULC PSICH032-08(A-720-H/ULC PSICH032-08(A-720.M/ULC PSICH032-08IA-720 M/ULC PSICH026-06IA-720.H/ULC PSICH026-06IA.720.M/ULC PSICH026-06IA-720 M/ULC ICC004-01IA.700-SF/RK PSICH017-04(A-720-SF/RK PSICH011-02(A-720-M/ULC PCICH007.-01IA-680-SF/RM I.I-Iav2 IAV3 106H IAV2 I AV3 (AV4 (06C (AV3 106K Iav2 (06C IAV2 Iav3 IAV3 Iav2 Iav3 106K 101C 106H (02H Iav2 IAV3 IAV4 IAV3 Iav2 106H (02c 104H (02H IBAH I---.1.1 5.8.0.0.0 3.2.0 2.2.0 37.3.0 9.1 11.9.0.0 14~1 50.3 2.3-.6.0.0.0.1 7.9 22.8 4.8 40.4 15.4 19.9 I Location I 09/92 (volts(Deg(Ch I I I I I.9(IP 21.7(14sl.8(Ii 2(1.6(Ii 2(.6(Ii 2(.41135(li 2(.8(13S(.6(Ii 2(.511491 11 Ii 2(411451 11.511451 11.s(li 21 Ii 2(.611S1(.7(156(.611591 11.6(106(P 1(Ii 2(Ii 2(IP 2(.7(li 21.511381 11.3(1491.611571 11.511561 11.4(1SDI I N/A I X (Diff(Location lvoltslDeglch I X I-I-I 321 26(I 321 28(341 381 20(39(221 I I I I I I I I I I I I I I I I I I I I 26(241 30(301 241 25(20(37(231 211 211 35(27(35(20(28(---I-Number of RECORDS Selected from Current Outage: 73 Number of TUBES Selected from Current Outage: 51 135

~t=PL Page t of1 FORM NIS.BB OWNERS'ATA REPORT FOR EDDY CURRENT EXAMINATION RESULTS As required by the provisions of the ASME CODE RULES EDDY CURRENT EXAMINATION RESULTS PLANT: Turkey Point Unit 4 EXAMINATION DATES: APRIL 24, 1993 thru APRIL 28, 1993 STEAM GENERATOR 4E210A 4E210B 4E210C TOTAL TUBES INSPECTED 3198 3206 3205 19 TOTAL TUBES 40%-100%NONE NONE NONE TUBES PLUGGED AS PREVENTNE MAINTENANCE NONE NONE NONE TUBES PLUGGED THIS OUTAGE NONE NONE NONE TOTAL PLUGGED TUBES IN S/G 16 STEAM GENERATOR AVB BARS LOCATION OF INDICATIONS (20%-100%)SUPPORT LOCATIONS 1 THROUGH 6 TOP OF TUBE SHEET TO¹t SUPPORT HOT LEG COLD LEG HOT LEG COLD LEG TOTAL INDICATIONS 20'/o-39%40%TO 100%%uo 4E21 OA NONE 4E210B 4E21 OC Remarks: 12 NONE 7 NONE NONE NONE 23 NONE NONE NONE DATE: DATE DATE CERTIFICATION OF RECORD We certify that the statements in this report are correct and the tubes inspected were tested in accordance with the requirements of Section XI of the ASME Code.Florida Power 8 Light Co.PREPARED BY:~~'+r~~~+4'~'DDY URRENT DINAT R REVIEWED BY: IN SU RVISOR APPROVED BY: S/PR RAM M ER an FPL Group company 136 CUHULAT I VE EXAHI HAT I ON REPORT PTN-4 OUTAGE: 04/93 COHPONENT: S/G A DESCRIPTION

20X TO 100X Page: 1 of 1 Date: 6/4/93 Time: 9:35 AN I I Extent I IRowIColILegIReqITst/NoteI Reel I Probe I Location+---I---I--I---I-------I--.-I-------"---I-I 281 141 C ITEHITEH IAC006-02IA-720-N/ULC 10'IH 42.6 I I I C ITEHITEH IAC006.02IA-720-N/ULC 102C 2.7 I 141 821 H ITECITEC PSIAH004-02IA-720.H/ULC 104C 9.5'---I---I---I---I--------I-I------I-04/93 I I IVoltsIDegICh I II IDiffI I--.-I---I---I---I--I-I 1.61 1451 11 321 I I.811461 11 311 I.811571 11 271 I--I---I---I--I-I+N/A I Location IVoltsIDegICh I X I----I-I--I---I---+

I I I I I I I I I I I I I I I I I I Hunber of RECORDS Selected from Current Outage: 3 Number of TUBES Selected from Current Outage: 2 137 t:OMPONENT: S/G B DESCRlPTlON

20K TO 100X CUMULATIVE EXAMI HAT ION REPORT PTN.4 OUTAGE: 04/93 Page: 1 of 1 Date: 6/4/93 Time: 9:35 AM I I Extent (Roe(Col(Leg(Req(Tst/Note(Reel I Probe'-I---I---I--

I I I 8(18(C (TEH(TEH IBC016.04(A-700.H/ULC 3(24(C (06C(06C IBC011-03(A-720-M/ULC I 221 371 H ITECITEC SSIBH026-07IA-720-M/ULC I 45(42(H (TECITEC IBH013-03(A-720.M/ULC (45(43(H (TEC(TEC IBH013-03(A-720.M/ULC I 221 481 H ITECITEC PCIBH023-06IA-720-M/ULC (45(48(H ITECITEC PSIBH01'1-03(A-720-M/ULC I 371 691 H ITECITEC SSIBH030.03IA-720.M/ULC I 141 821 H ITECITEC SSIBH004-02IA-720.M/ULC

'---I---l--I-" I I I I Location I ITSC 3.9 103C 25.5 102H.0 IAV2.0 JaVa.0 ITSH 6.3 IAV4.0 ITSH 21.6 102H 15.6 I 04/93 (Volts(Deg(Ch (X (Diff-I----I--I-.-I-I---.7(148(1(30(I.911501 11 281.4(143(1(32(.5((i 2(22(~6((i 2(23(I.7(1431 11 32(IP 2(24(.7(145(1(31(1.2(150(1(26(-I-I I I I N/A I Location (Volts(geg(Ch (X (I-------------

I----I--I--I---'I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I-------------I-----I---I---I---+

Nether of RECORDS Selected from Current Outage Number of TUBES Selected from Current Outage: 9 138 0

CUMULATIVE EXAMI HAT ION REPORT PTN-4 OUTAGE: 04/93 COMPOHENT: S/G C DESCRIPTION

20X.,TO 100X Page: 1 of 1 Date: 6/4/93 Time: 9:35 AM I (Row(Col(Leg

'-l---l (z8(z8(c I c I 37(3z(c (z6(37(c (3(SZ(C I 431 521 H I 401 531 c I 24(S6(H I 421 561 H I I IH I I IH I 331 611 c (24(62(H I 371 691 c I 321 701 H I 16(72(H I 3o(rz(H I 37(72(H (31(8O(H I zr(81(H I 261 821 H'.l.--l.I Extent (Req(Tst/N I I.'"-ITEHITEH ITEHITEH ITEHITEH ITEHITEH (o6c(o6c ITECITEC ITEHITEH ITECITEC ITECITEC ITECITEC ITECITEC ITEHITEH ITECITEC ITEHITEH ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC ITECITEC (TECITEC ITECITEC ITECITEC I.--I-ote(Reel I Probe I----I----.---ICC016-04IA-720-M/ULC PSICC016 04(A 72 M/ULC ICC018.04IA-720.M/ULC ICC014-03IA-720-M/ULC PSICC013-03(A-720.M/ULC SSICH022-05IA-720.M/ULC SCICC021-05IA-720-M/ULC PSICH023-06(A-720.SF/RM SSICH IA 720 M/ULC SSICH022.05IA-720-M/ULC SSICH022.05IA.720.M/ULC SSICC021-05IA-720.M/ULC PSICH024.06(A.720-M/ULC PC(CC022-06(A-720.H/ULC PSICH008.03(A 720.H/ULC ICH002-01IA-720.H/ULC SCICH006-02(A 720 M/ULC PSICH008.03IA-720.M/ULC ICH008 03IA-720-M/ULC PSICH007-02IA-720.M/ULC SICH022 05(A 720 M/ULC SCICH007-02IA-720.H/ULC SCICH007-02IA.720.H/ULC I I I Location I-105H 44.0 106H-.9 (o6c-.r 105C 31.5 101C 29.6 IDSH 12.0 (06C-.5 103C 38.7 (o6c..6 (osc (osc 32.3 13.2 (03C (O2C 26.1 35.3 106H IAV1 IDSH~1.0 43.0 (04C 21.1 (OSH 4S.O IAv3..-.z Iozc 106H IAV1 IAV1 I-50.6 2.3.0.0 I.911531 I 1.011491.71119IP 11.811571 11~61152(1(.611501 11.6(127(I 1(.611351 11 I 1.21120IP 11.511411 11.611431 11.311441 11.611621 11~511501 11.5(Ii z(.4(147(.311521 11.511391 11.51148IP 21.411481 11~4(142(~3(IP zl II 2(I-I I 3O(zo(zs(24(I 3o(zs(3o((33(>>I I zs(22(3o(261 I 361 I 231 I z8(zo(zo(".I.I 04/93 I I (Volts(oeg(Ch I X (Diff(I I I I I I+N/A I Location (Volts(oeg(Ch (I (I I---I---I-"+

I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I.I---I" I.-'unber of RECORDS Selected from Current Outage Number of TUBES Selected from Current Outage: 23 19 139 EtIJ r COMPONENT: S/G A CUMMULATIVE DISTRIBUTION

SUMMARY

TURKEY POINT UNIT g 4 04/93 Page: 1 of 1 Date: 06/14/93 Time: 1:30 PM Examination Dates: 04/24/93 thru 04/28/93 Total Number of Tubes Inspected.....: 3198 Total Indications Between 20'.and 39%Greater than or equal to 40'.,~...Total Tubes Plugged as Preventive Maint Total Tubes Plugged 0 16 Location Of Indications 20%to 100%E Hot Leg TSH-.5 to 01H-2.1 01H-2.0 to 06H+2.0 06H+2.1 to AV1-3.1 AVl-3.0 to AV4-3.0 Cold Leg 0 TSC-.5 to 01C-2.1: 0 1 01C-2.0 to 06C+2.0: 2 0 06C+2.1 to AV4-3.1: 0 COMPONENT: S/G B CUMMULATIVE DISTRIBUTION

SUMMARY

TURKEY POINT UNIT 5 4 04/93 Page: 1 of 1 Date: 06/14/93 Time: 1:30 PM Examination Dates: 04/24/93 thru 04/28/93 Total Number of Tubes Inspected~....: 3206 Total Indications-Between 20%and 39%Greater than or equal to 40%Total Tub'es Plugged as Preventive Maint Total Tubes Plugged Location Of Indications 20%to 100'.Hot Leg TSH-.5 to 01H-2.1 01H-2.0 to 06H+2.0 06H+2.1 to AV1-3.1 AV1-3.0 to AV4-3.0 Cold Leg 2 TSC-.5 to 01C-2.1 2 01C-2.0 to 06C+2.0 0 06C+2.1 to AV4-3.1 14,1 r

COMPONENT: S/G C CUMMULATIVE DISTRIBUTION

SUMMARY

TURKEY POINT UNIT g 4 04/93 Page: 1 of 1 Date: 06/14/93 Time: 1:30 PM Examination Dates: 04/24/93 thru 04/28/93 Total Number of Tubes Inspected 3205 Total Indications Between 20.and 39%Greater than or equal to 40~Total Tubes Plugged as Preventive Maint Total Tubes Plugged 23 0 Location Of Indications 20'.to 100'.Hot Leg Cold Leg TSH-,5 to 01H-2.1: 0 TSC-.5 to 01C-2.1: 0 01H-2.0 to 06H+2.0 06H+2.1 to AV1-3.1 AV1-3.0 to AV4-3.0 6 01C-2.0 to 06C+2.0: 12 4 06C+2.1 to AV4-3.1: 0 14?,