L-2002-198, 10 CFR 50.59 Report

10 CFR 50.59 Report

ML022910376
Person / Time
Site:	Turkey Point
Issue date:	10/07/2002
From:	Mcelwain J Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
-nr, FOIA/PA-2006-0214, L-2002-198
Download: ML022910376 (67)

Text

Ou 0 77n L-2002-198 10 CFR. 50.59 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Re: , Turkey Point Units 3 & 4 Docket Nos. 50-250 and 50-251 10 CFR 50.59 Report Florida Power and Light Company's summary report on Changes, Tests and Experiments Made Without Prior Commission Approval, for the period from October 24, 2000, through April 7, 2002, is attached. It also contains a summary of the Power Operated Relief Valve actuations, and the results of the Unit 3 steam generator tube inspection, which occurred during that time. This report also includes the reload safety evaluation summaries for Unit 3 Cycle 19, dated February 27, 2002, and Unit 4 Cycle 20, dated May 22, 2002.

Very truly yours, 6Zýlen Vice President Turkey Point Plant DRL Attachment cc: Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant an FPL Group company

ATTACHMENT to L-2002-198 Page I of 66 2000/2002 ANNUAL 10 CFR 50.59

SUMMARY

REPORT FLORIDA POWER & LIGHT COMPANY TURKEY POINT UNITS 3 & 4 1

ATTACHMENT to L-2002-198 Page 2 of 66 TURKEY POINT PLANT UNITS 3 AND 4 DOCKET NUMBERS 50-250 AND 50-251 CHANGES, TESTS AND EXPERIMENTS MADE AS ALLOWED BY 10 CFR 50.59 FOR THE PERIOD COVERING OCTOBER 24, 2000 THROUGH APRIL 7, 2002 2

ATTACHMENT to L-2002-198 Page 3 of 66 INTRODUCTION This report is divided into five (5) sections. The first section summarizes those changes made to the facility as described in the UFSAR that were performed by a Plant Change/Modification (PC/M). The second section summarizes those changes made to the facility or procedures as described in the UFSAR that were performed by a 10 CFR 50.59 evaluation. This includes those changes not performed by a PC/M, and any tests and experiments not described in the UFSAR that were performed during this reporting period. The third section provides a summary of the Unit 3 and Unit 4 fuel reload evaluations. The fourth section provides a list of power operated relief valve (PORV) actuations. This section is included as part of FPL's commitment to comply with the requirements of Item II.K.3.3 of NUREG 0737. The fifth and last section of this report provides a summary of the findings of any steam generator tube inspections. Only Unit 3 had steam generator tube inspections during this reporting period.

3

ATTACHMENT to L-2002-198 Page 4 o f 66 TABLE OF CONTENTS PAGE INTRODUCTION 3 TABLE of CONTENTS 4 SECTION 1 PLANT CHANGEJMODIFICATIONS96-096 PLANT / C-BUS RELIABILITY IMPROVEMENTS - UNIT 4 10 03/27/2001 97-052 ONE-HOUR THERMO-LAG UPGRADES FOR OUTDOOR 11 FIRE ZONES 02/08/2001 97-055 INTAKE STRUCTURE BAY WALLS CATHODIC PROTECTION 12

- UNIT 3 12/31/2000 97-057 TWENTY-FIVE MINUTE THERMO-LAG UPGRADES FOR 13 OUTDOOR FIRE ZONES 03/19/2001 99-042 INSTRUMENT AIR DRYER REPLACEMENT 14 05/01/2001 99-054 APPENDIX R DOCUMENTATION CHANGES IN SUPPORT OF FIRE 15 PROTECTION FUNCTIONAL INSPECTION ONOP / SSA MANUAL ACTIONS REVIEW 11/08/2000 00-002 TURBINE LUBE OIL CONDITIONER REPLACEMENT - UNIT 3 16 11/20/2001 00-009 STEAM GENERATOR FLEXIBLE TUBE STAKES 17 04/03/2001 00-021 UNIT 4 PERMANENT REMOVAL OF PRESSURIZER CUBICLE 18 MISSILE SHIELD PLUG 01/02/2001 00-022 REACTOR CAVITY SEAL ALTERNATIVE CONFIGURATION 19 10/10/2001 00-027 COLD OVERPRESSURE MITIGATION SYSTEM (COMS) 20 SETPOINT CHANGE 01/30/2002 4

ATTACHMENT to L-2002-198 Page 5 of 66 TABLE OF CONTENTS (Continued)

SECTION 1 PLANT CHANGE/MODIFICATIONS (Continued) PAGE 00-042 EMERGENCY DIESEL GENERATOR GOVERNOR CONTROL 21 CIRCUIT ENHANCEMENT - UNIT 4 07/11/2001 00-043 EMERGENCY DIESEL GENERATOR GOVERNOR CONTROL 22 CIRCUIT ENHANCEMENT - UNIT 3 04/12/2001 23 01-014 MOV-3-843A/B AND MOV-3-869 MODIFICATIONS - SI SYSTEM ENHANCEMENTS - UNIT 3 10/20/2001 24 01-059 AFW BUS STRIPPING RESET MODIFICATION - UNIT 4 04/04/2002 25 01-063 ABANDONMENT OF HYDROGEN RECOMBINER EXHAUST LINE TO CONTAINMENT & REPLACEMENT OF CHECK VALVE 4-40-205

- UNIT 4 04/04/2002 26 02-012 PERMANENT PLATFORMS / SCAFFOLDING / COMPONENTS IN CONTAINMENT 04/04/2002 SECTION 2 10 CFR 50.59 EVALUATIONS SEEJ-88-042 DE-ENERGIZATION OF UNIT 4 4160 VOLT SAFETY RELATED 28 BUSES TO ALLOW FOR MODIFICATIONS OR PERIODIC MAINTENANCE 03/11/2002 29 SENJ-89-084 SAFETY EVALUATION FOR REACTOR VESSEL MISSILE SHIELD REMOVAL DURING REDUCED RCS PRESSURE CONDITIONS 03/16/2001 30 SEEJ-89-085 DE-ENERGIZATION OF UNIT 3 4160 VOLT SAFETY RELATED BUSES 09/25/2001, 10/0512001 5

ATTACHMENT to L-2002-198 Page 6 o f 6 6 TABLE OF CONTENTS (Continued)

PAGE 10 CFR 50.59 EVALUATIONS SECTION 22 10 SECTION SECS-90-018 SPENT FUEL POOL KEYWAY GATE BOOT SEAL 31 REPLACEMENT 12/28/2000 SEMS-91-019 RHR HEAT REMOVAL SYSTEM IN-SERVICE TESTING 32 SAFETY EVALUATION 10/02/2001 33 SEMS-93-059 CONNECTION OF A TEMPORARY FILTER ASSEMBLY TO THE DIESEL FUEL OIL STORAGE TANKS 03/06/2001 34 SENP-95-007 UNIT 3 OPERABILITY OF RHR AND REFUELING SUPPORT EQUIPMENT DURING INTEGRATED SAFEGUARDS TESTING 10/15/2001 35 SENP-95-023 UNIT 4 OPERABILITY OF PLANT EQUIPMENT DURING INTEGRATED SAFEGUARDS TESTING 03/25/2002, 3/28/2002 36 SEMS-96-003 SAFETY EVALUATION FOR UNIT 4 STEAM GENERATORS' SECONDARY SIDE FOREIGN OBJECTS 12/05/2000 37 SEMS-96-014 A TEST OF THE USE OF SUB-MICRON ULTRAFINE FILTERS IN THE CVCS AND SFP 12/12/2000 38 SEMS-96-038 SAFETY EVALUATION FOR UNIT 3 STEAM GENERATORS' SECONDARY SIDE FOREIGN OBJECTS 02/14/2001 SEMS-96-040 SAFETY EVALUATION FOR THE TEMPORARY 39 INSTALLATION OF DRAIN HOSES AND PERFORMANCE OF HOT SPOT FLUSHES ON THE RHR SYSTEM 2/26/2001 SEMS-97-002 IMPACT OF CURE TIME ON THERMO-LAG FIRE BARRIER 40 PERFORMANCE 01/23/2001 SECS-98-0058 EVALUATION FOR STORAGE OF TOOLS AND EQUIPMENT 41 IN CONTAINMENT DURING ALL MODES OF OPERATION 10/17/2001, 10/25/2001 6

ATTACHMENT to L-2002-198 Page 7 of 66 TABLE OF CONTENTS (Continued)

SECTION 2 10 CFR 50.59 EVALUATIONS (Continued) PAGE SEFJ-00-026 UFSAR UPDATE TO OFFSITE RADIOLOGICAL DOSES DUE 42 TO CHANGES IN ASSUMPTIONS ASSOCIATED WITH STEAM GENERATOR TUBE RUPTURE AND STEAM LINE BREAK 01/18/2001 SENS-00-046 SAFETY EVALUATION FOR TEMPORARY LOWERING 43 OF UNIT 4 SPENT FUEL POOL WATER LEVEL FOR MAINTENANCE ACTIVITIES 05/10/2001 SENS-00-088 SAFETY EVALUATION FOR THROTTLING CCW MANUAL 44 VALVES 3/4-737A 11/21/2000 SEFJ-01-006 REACTOR COOLANT SYSTEM CHEMISTRY pH CONTROL 45 05/10/2001 46 SENS-01-024 SAFETY EVALUATION FOR OFFSITE DOSE CALCULATION MANUAL REVISION 9 03/22/2001 47 SEMS-01-025 ADDITION OF ALTERNATE AMINES TO THE SECONDARY SIDE SYSTEM 05/11/2001 48 SEFJ-01-026 UFSAR AND DBD CHANGE PACKAGES FOR THE REANALYSIS OF LOSS OF LOAD EVENT AND DBD CHANGE PACKAGE FOR THE ROD WITHDRAWAL AT POWER ANALYSIS 01/10/2002 49 SEMS-01-031 REPAIR OF HYDROGEN EXCESS FLOW CHECK VALVE, REPLACEMENT OF RV-4622 AND REPAIR OF LEAKING LINE TO GAS HOUSE 04/23/2001 TEMPORARY LOWERING OF UNIT 3 SFP LEVEL 50 SENS-01-057 08/28/2001, 09/13/2001 51 SECS-01-059 10 CFR 50.59 EVALUATION FOR STORAGE OF TWO NIS DETECTORS IN CONTAINMENT DURING ALL MODES OF OPERATION 08/17/2001 SENS-01-082 ISOLATION OF THE PRESSURIZER RELIEF TANK BRANCH 52 OF THE REACTOR COOLANT GAS VENT SYSTEM 11/21/2001 7

ATTACHMENT to L-2002-198 Page 8 of 66 TABLE OF CONTENTS (Continued)

SECTION 2 10 CFR 50.59 EVALUATIONS (Continued) PAGE SEMS-02-001 EARLY CORE OFFLOAD 53 03/14/2002, 03/25/2002 SECTION 3 RELOAD 10 CFR 50.59 EVALUATIONS01-036 TURKEY POINT UNIT 3 CYCLE 19 RELOAD DESIGN 55 02/27/2002 56 01-065 TURKEY POINT UNIT 4 CYCLE 20 RELOAD DESIGN 05/22/2002 SECTION 4 REPORT OF POWER OPERATED RELIEF VALVE (PORV) ACTUATIONS UNIT 3 58 UNIT 4 58 SECTION 5 STEAM GENERATOR TUBE INSPECTIONS FOR TURKEY POINT UNIT 3 60-66 8

ATTACHMENT to L-2002-198 Page 9 o f 6 6 SECTION 1 PLANT CHANGE / MODIFICATIONS 9

ATTACHMENT to L-2002-198 Page 10 of 66 PLANT CHANGE/MODIFICATION 96-096 Revision 1 UNIT: 4 TURN OVER DATE: 03/27/2001 PLANT/C-BUS RELIABILITY IMPROVEMENT MODIFICATIONS Summary:

This Engineering Package (EP) re-powered several Unit 4 non-safety electrical loads from non-vital sources to vital power sources. Each load was re-powered to improve the reliability and availability of the plant and to give operation personnel added control and flexibility when shutting down the unit in response to a unit trip resulting from a loss of offsite power (LOOP). For example, prior to the modification, failure of the C-bus resulted in loss of control power and indication for the 4C and 4D main steam reheater (MSR) stop valves and volume control tank (VCT) outlet valve. These valves were fed from the non-vital side of the 4B motor control center (MCC) which were powered from the C-bus. The extraneous operator actions required to respond to these valve failures diverted attention away from more essential recovery actions, and hampered the overall recovery process. To streamline the recovery process for C-bus failures and loss of offsite power events, these components were repowered to the vital side of the 4B MCC. In addition, the power feeds of the component cooling water (CCW) surge tank make-up valve (MOV-4-832) and the 4B primary water pump were switched from a non-vital power supply to a vital power supply, to enhance CCW system reliability during postulated accident conditions.

Revision I of this EP was issued to remove the requirements of re-powering the 4B steam generator feedwater pump from the 4C bus to the 4B bus. All other modifications described in the EP were implemented.

10 CFR 50.59 Evaluation:

The addition of the non-safety loads to the vital buses maintained bus independence and did not affect any of the vital bus ratings or bus protective relay settings. Calculations demonstrated that existing electrical distribution equipment will continue to operate within their design limits during steady state and transient operating conditions. Additionally, it was concluded that emergency diesel generator loading would not be impacted by the addition of these manual loads due to their low power consumption rating and intermittent nature of operation. An engineering review further demonstrated that the seismic qualification of the vital MCC panels were not adversely affected by the modifications due to the small weight changes involved. Based on the design package evaluation, the electrical modifications did not have any adverse effects on plant safety or operation or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation.

10

ATTACHMENT to L-2002-198 Page 11 of 66 PLANT CHANGE/MODIFICATION 97-052 UNIT: 3 &4 TURN OVER DATE: 02/08/2001 ONE-HOUR THERMO-LAG UPGRADES FOR OUTDOOR FIRE ZONES Summary:

Based on a fire hazards analysis, some outdoor cables either within 50 feet of a major combustible or within the area of highest temperatures of a postulated turbine lube oil fire require a one-hour fire bamer. Based on Industry scrutiny of Thermo-Lag raceway protection, it was found that many fire barriers did not provide the intended fire rating. Consistent with associated FPL commitments, this Engineering Package (EP) was issued to implement upgrades to satisfy a one-hour fire barrier requirement for those raceways containing essential circuits in outdoor fire zones, specifically in the area of the Unit 4 startup and main transformers, the Unit 3 main condenser and the Unit 3 and 4 condensate pump and condensate pit areas. This EP also addressed the affect of these fire barrier upgrades on ampacity derating. Modifications included both physical upgrades as well as qualification by evaluation.

10 CFR 50.59 Evaluation:

The modifications performed by this EP included additional overlays of Thermo-Lag materials and addition of oil drip shields where necessary. Qualification of Thermo-Lag systems upgraded or installed per this EP were based on fire endurance tests conducted within the industry and specifically by FPL. The application process and configuration of additional material to existing Thermo-Lag installations did not undermine the ampacity requirements of the electrical circuits or the structural integrity of the raceways. As such, these modifications did not affect the availability or function of any equipment whose malfunction is postulated in the UFSAR to initiate an accident or prevent an accident from occurring. The modifications did not undermine or require changes to normal plant power operations and thus there was no affect on expected frequency of occurrence of postulated events. The modifications did not change the availability or decrease the design basis performance capability of equipment important to safety to perform their respective safety related functions. Since the proposed changes did not compromise plant safety or require any change to technical specifications, prior NRC approval was not required for implementation.

11

ATTACHMENT to L-2002-198 Page 12 of 66 PLANT CHANGE/MODIFICATION 97-055 UNIT: 3 TURN OVER DATE: 12/31/2000 INTAKE STRUCTURE BAY WALLS CATHODIC PROTECTION - UNIT 3 Summary:

This Engineering Package (EP) provided for the installation of an impressed current cathodic protection system at the Intake Structure bay walls in all four (4) of the Unit 3 intake wells. This impressed current cathodic protection system was added as a result of the discovery of corrosion damage at several of the steel reinforcing bars embedded in the bay walls. The system was provided to protect this reinforcing steel from further corrosion. The cathodic protection system was powered from a non-safety related lighting panel since it is not required to be functional for the intake structure to perform its safety related function. To preclude any adverse seismic interaction with adjacent safety related equipment and components, the cathodic protection system was seismically designed.

10 CFR 50.59 Evaluation The modifications performed by this EP were implemented in such a manner as to preclude adversely impacting the structural integrity of the adjacent bays or the continued operability of the intake cooling water (ICW) and circulating water (CW) pumps housed within the bays. The configuration of the intake structure continued to meet all applicable loads in accordance with the requirements of Class I structures described in UFSAR Appendix 5A, including operating loads of the pumps and associated components. No SSCs required for accident mitigation were adversely affected by the installation of this cathodic protection system. Since the modifications did not impact safe operation of the plant or require changes to the plant technical specifications, prior NRC approval was not required for implementation.

12

ATTACHMENT to L-2002-198 Page 13 of 66 PLANT CHANGE/MODIFICATION 97-057 Revision 3 UNIT: 3 &4 TURN OVER DATE: 03/19/2001 TWENTY-FIVE MINUTE THERMO-LAG UPGRADES FOR OUTDOOR FIRE ZONES Summary:

Based on a fire hazards analysis, some outdoor cables required for safe shutdown, require a 25-minute fire barrier. This Engineenng Package (EP) implemented upgrades to provide a 25-minute fire barrier for those raceways containing essential circuits in outdoor fire zones and addressed affects of the fire barrier upgrade on ampacity derating. Fire barrier requirements are designed to satisfy NRC Generic Letter 86-10, Supplement 1. Consistent with associated FPL commitments and exemptions issued by NRC, this EP implemented upgrades to satisfy a 25-minute fire barrier requirement for those raceways containing essential circuits in outdoor fire zones located in the turbine building and at a distance greater than 50 feet from a major combustible. Modifications included both physical upgrades as well as qualification by evaluation.

10 CFR 50.59 Evaluation The application process and configuration of additional material to existing Thermo-Lag installations did not undermine the ampacity requirements of the electrical circuits or the structural integrity of the affected raceways. As such, these modifications did not affect the availability or function of any equipment whose malfunction is postulated in the UFSAR to initiate an accident or prevent an accident from occurring. The modifications did not undermine or require changes to normal plant power operations. Thus, there was no adverse affect on expected frequency of occurrence of postulated events. The modifications did not change the availability or decrease the design basis performance capability of equipment important to safety to perform their respective safety related functions. Since the modifications did not compromise plant safety or require any change to the plant technical specifications, pnor NRC approval was not required for implementation.

13

ATTACHMENT to L-2002-198 Page 14 of 66 PLANT CHANGE/MODIFICATION 99-042 UNIT: 3 &4 TURN OVER DATE: 05/01/2001 INSTRUMENT AIR DRYER REPLACEMENT Summary:

The purpose of the instrument air system is to provide motive power and control air for various safety related, quality related and non-nuclear safety related pneumatic components. This Engineering Package (EP) provided for the replacement of the instrument air (IA) dryer packages installed in each unit's instrument air system including desiccant chambers, automatic dryer controls, valves, silencers and associated piping and controls. This EP also provided for the addition of a secondary high efficiency filter upstream of the dryers to protect the desiccant from liquid moisture intrusion. The activity was required to replace aging, maintenance-intensive equipment with more efficient, reliable equipment having the same performance characteristics. The instrument air dryer equipment replaced or added by this EP is not required to be functional to ensure safe shutdown capability.

An UFSAR change package was provided as an attachment to this EP to reflect changes to dryer performance and did not alter any UFSAR conclusions.

10 CFR 50.59 Evaluation The modifications performed by this EP replaced existing IA dryers with new internally heated dryers with an improved design. The modification did not alter or adversely affect the power supply or reduce the performance capability of the IA system. The changes met or exceeded design, material and construction codes for IA system pressure and temperature design limits. Potential for loss of IA capability due to filter blockage was not increased and no common failure mode was created which would prevent supply from the opposite unit's IA system. The IA dryers are not credited in any design basis accident analysis. Thus, replacing the IA dryers and associated equipment had no affect on accident consequences. Since this modification did not compromise plant safety or require any change to the plant technical specifications, prior NRC approval was not required for implementation.

14

ATTACHMENT to L-2002-198 Page 15 of 66 PLANT CHANGE/MODIFICATION 99-054 UNIT: 3 &4 TURN OVER DATE: 11/08/2000 APPENDIX R DOCUMENTATION CHANGES IN SUPPORT OF FIRE PROTECTION FUNCTIONAL INSPECTION (FPFI)

ONOP / SSA MANUAL ACTIONS REVIEW Summary:

As part of the Fire Protection Functional Inspection (FPFI) Self Assessment project the Appendix R Safe Shutdown Analysis manual actions were reviewed against actions identified in specific plant procedures. This Engineering Package (EP) evaluated and/or resolved the inconsistencies and discrepancies identified between plant procedures and Appendix R engineering documents and provided the required inputs to Units 3 & 4 plant procedures and administrative changes to Appendix R engineering documents.

An UFSAR change package was provided as an attachment to this EP to update the Appendix R documentation associated with this manual actions review.

10 CFR 50.59 Evaluation This EP provided resolution of discrepancies discovered between plant procedures and Appendix R engineering documents. No physical changes were required. Appendix R engineering document changes were administrative only. These administrative "design changes" did not deviate from the established fire protection design criteria, design bases or regulatory requirements as described in the UFSAR. As such, there were no adverse affects on probability of occurrence or consequences of any accident evaluated in the UFSAR. These changes did not affect the probability of occurrence or consequences of any malfunctions of equipment important to safety previously evaluated in the UFSAR. The changes provided by this EP did not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR. The technical specification requirements and bases applicable to this modification were not affected and therefore there was no adverse affect on the margin of safety as defined in the plant technical specification bases. Since this modification did not impact the safe operation of the plant or require a change to the plant technical specifications, this EP was determined to not require prior NRC approval.

15

ATTACHMENT to L-2002-198 Page 16 of 66 PLANT CHANGE/MODIFICATION 00-002 UNIT: 3 TURN OVER DATE: 11/20/2001 TURBINE LUBE OIL CONDITIONER REPLACEMENT - UNIT 3 Summary:

This Engineering Package (EP) provided permanent facilities to replace the installed Unit 3 turbine lube oil filtration system with a KAYDON TURBO-TOC self-contained, skid-mounted oil conditioning system. The model installed by this EP has been in operation at other sites for some years and its performance has been proven superior to the existing filtration system. This EP also provided for raising a portion of the existing fire suppression ring protecting the lube oil conditioning area to allow access to the skid control panel. A UFSAR change package was included as an attachment to the EP to update the descriptive information about the Unit 3 turbine lube oil conditioner. This change package was consolidated with a similar package for Unit 4 for inclusion in the UFSAR update previously submitted to NRC.

10 CFR 50.59 Evaluation The modifications addressed by this EP did not impact operation, function, or design basis of any safety related equipment. The new filtration skid enhanced the turbine lube oil conditioning process.

No changes were made to any of the fire suppression system spray patterns or degree of component coverage. Additionally, the installation did not increase the probability of a lube oil spill. Since no technical specifications were affected, the changes implemented by this EP did not require prior NRC approval for implementation.

16

ATTACHMENT to L-2002-198 Page 17 of 66 PLANT CHANGE/MODIFICATION 00-009 Revision 1 UNIT: 3&4 TURN OVER DATE: 04/03/2001 STEAM GENERATOR FLEXIBLE TUBE STAKES Summary:

This Engineering Package (EP) provided for the installation of ABB-designed steam generator (S/G) flexible tube stakes and plugs in Turkey Point Units 3 and 4 S/Gs. These stakes are used where circumferential indications are found during eddy current testing and function to restrain severed tubes and dampen vibrations to mitigate additional wear on adjacent tubes. Revision 1 to this EP provided updated tube staking cnteria and requirements for Unit 4, provided additional design clarifications and allowed the use of "Sequoyah" tube stakes as an alternative to the ABB-designed tube stakes. A UFSAR change package was provided as an attachment to this modification package to update the descriptive information on the use of S/G flexible tube stakes.

10 CFR 50.59 Evaluation Cntena for plugging S/G tubes with circumferential indications are based on industry guidance and NRC recommendations. Stakes are installed in tubes no longer in service to remove heat from the RCS. These components function to protect the RCS pressure boundary, minimize primary to secondary leakage and prevent multiple tube damage. The equipment was analyzed to demonstrate it can withstand application of potential design loads and not interact with active tubes in the S/G.

Therefore, this modification did not increase the probability of occurrence or consequences of an accident previously analyzed in the UFSAR. Because the stake will be retained within both ends of a severed S/G tube and thus will not contact other tubes or S/G internals, the modification did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR. The stake is a S/G support component. It is installed in a tube with a circumferential crack located at or near the top of the tube sheet and the tube is then plugged, thereby isolating the stake. Since no new hazards are created by the installation of flexible tube stakes in the S/Gs, the actions and documentation changes identified in the EP did not adversely impact plant safety, or require changes to the plant technical specifications. Therefore, prior NRC approval for implementation of this modification was not required.

17

ATTACHMENT to L-2002-198 Page 18 of 66 PLANT CHANGE/MODIFICATION 00-021 UNIT: 4 TURN OVER DATE: 01/02/2001 UNIT 4 PERMANENT REMOVAL OF PRESSURIZER CUBICLE MISSILE SHIELD PLUG Summary:

This Engineering Package (EP) provided for the permanent removal and disposal of the pressurizer missile shield plug from the Unit 4 containment structure. Removal of this plug eliminates a heavy load lift in the containment, assists in refueling activities, facilitates inspection activities during normal operation and allows for increased air circulation within the pressurizer cubicle resulting in lower equipment operating temperatures. A UFSAR change package was provided as an attachment to this modification package to address the removal of the Unit 4 pressurizer cubicle missile shield.

10 CFR 50.59 Evaluation The original design function of the pressurizer cubicle missile shield was to afford protection from missiles potentially impacting safety related structures, systems and components (SSC) including the containment liner and pipe penetrations. This EP evaluated the credibility of missile generation from within the pressurizer cubicle based on the criteria for missiles provided in the UFSAR, and determined that there is no threat to the containment liner or other SSCs previously considered.

Additionally, there are no postulated missiles from outside the cubicle that could affect the equipment within the cubicle. The associated reduction in containment heat sink and increase in containment free volume created by the shield plug removal were determined to be insignificant. It was concluded by qualitative analysis that the change had a negligible affect on engineered safety features and containment performance during postulated accidents. Thus, this modification did not increase the probability of occurrence or consequences of an accident previously analyzed in the UFSAR.

Removal of the pressurizer cubicle missile shield plug did not increase either the probability of occurrence or consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR. No other potential targets or failure mechanisms were introduced by removal of this missile shield plug. Therefore, this modification did not create the possibility of either an accident or a malfunction of equipment important to safety of a different type than previously evaluated in the UFSAR. The pressurizer missile shield is not addressed in the plant technical specifications. Since this change did not adversely affect plant safety or operation and did not require a change in the plant technical specifications, prior NRC approval for implementation of this modification was not required.

18

ATTACHMENT to L-2002-198 Page 19 of 66 PLANT CHANGE/MODIFICATION 00-022 UNIT : 3&4 TURN OVER DATE: 10/10/2001 REACTOR CAVITY SEAL ALTERNATIVE CONFIGURATION Summary:

This Engineering Package (EP) provided an alternative reactor cavity seal configuration that can be used in lieu of the existing reactor cavity seal system for Units 3 and 4. The existing reactor cavity seal system consists of a continuous ring with a pre-loaded passive seal to prevent leakage in excess of design requirements. The passive seal consists of four EPDM polymer compression rings to provide the safety related means to limit reactor cavity leakage. Additionally, there is a non-safety related inflatable seal to reduce seal leakage to as low as practicable for ALARA and housekeeping considerations. The alternative seal configuration is a "T" shaped extruded EPDM seal sectioned into 5 units that rely on mechanical compression of the EPDM material. It does not require external pneumatic or hydraulic forces to maintain the seal. Seal compression is achieved using J-bolts that rotate under the RPV flange coupled with toggle nuts and a stainless steel cover plate that provides the pre-load. The cover plate caps the seal, providing protection from a dropped fuel assembly and transfers compression forces to the seal. The segmented aspect of the alternative design allows the dismantled seal to be removed from the containment following refueling and stored for later use in the other unit. A UFSAR change package was included in the EP to provide a description of the alternative configuration.

10 CFR 50.59 Evaluation The function and performance of the alternative segmented reactor cavity seal system is equivalent to that of the existing seal design. The alternative seal can withstand the design conditions considered for the existing seal and is considered a like-for-like design. The seal is designed to prevent a high volume loss of refueling water from the reactor cavity during fuel transfer. This design change will not affect the fuel handling accident analysis discussed in UFSAR Chapter 14.2. Therefore, use of this alternative segmented cavity seal does not increase the probability or consequences of occurrence of any accident or malfunction of equipment important to safety previously analyzed in the UFSAR.

The consequences of leakage of this seal are identical to that of the existing seal design and therefore will not create the possibility of an accident of a different type than previously evaluated in the UFSAR. No new hazards are created that can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the UFSAR. Since this change did not adversely affect plant safety, or require a change to the plant technical specifications, prior NRC approval was not required for implementation.

19

ATTACHMENT to L-2002-198 Page 20 of 66 PLANT CHANGE/MODIFICATION 00-027 UNIT 3&4 TURN OVER DATE: 01/30/2002 COLD OVERPRESSURE MITIGATION SYSTEM (COMS) SETPOINT CHANGE Summary:

This Engineering Package (EP) provided the design for the reactor coolant system (RCS) final Overpressure Mitigation System (OMS) setpoint to be implemented at Turkey Point for operation beyond 19 effective full power years (EFPY). The revised setpoint was 460 psig enabled at 285°F.

The OMS system is a Westinghouse version of Low Temperature Overpressure Protection (LTOP) that uses the power operated relief valves (PORV) in low pressure operating mode. The Turkey Point technical specifications were updated in Amendments 208/202 with new heatup and cooldown curves as well as new pressure-temperature (PT) limits applicable to 32 effective full power years (EFPY).

The new PT curves included the revision of the technical specification OMS setpoint maximum value from 415 +/- 15 psig enabled below 275°F to < 468 psig enabled below 275°F. Control room annunciators for OMS High Pressure Alert and OMS Control Activated were also increased to be consistent with the new OMS actuation setpoint. An UFSAR change package was provided as an attachment to this EP to reflect the new setpoints.

10 CFR 50.59 Evaluation The purpose of the OMS is to supplement the normal plant administrative controls to mitigate the potential for RCS bnttle fracture. The OMS is designed to provide the capability, during low pressure operation and water solid conditions, to automatically prevent the RCS metal components such as the reactor vessel from exceeding applicable limits established by 10 CFR 50, Appendix G. The UFSAR credits the OMS for RCS overpressure protection during low temperature operations. The setpoint change established by this EP did not impact the integrity of the RCS pressure boundary (RCPB) and therefore did not increase the potential for the occurrence of a loss of coolant accident. This EP did not make any physical changes to the facility design, material or construction standards. The setpoint change involved modification of LTOP limits and setpoints used to prevent accidents. No new hazards were created that can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed in the UFSAR. The technical specification requirements envelope the setpoint established in this EP. Since the modification did not adversely affect plant safety and did not require a change to the plant technical specifications, prior NRC approval was not required for implementation of the setpoint change.

20

ATTACHMENT to L-2002-198 Page 21 of 66 PLANT CHANGE/MODIFICATION 00-042 UNIT: 4 TURN OVER DATE: 07/11/2001 EMERGENCY DIESEL GENERATOR GOVERNOR CONTROL CIRCUIT ENHANCEMENT - UNIT 4 Summary:

This Engineering Package (EP) provided a design to rewire the governor control logic and added three new interposing relays within the existing Unit 4 vital 125-volt DC emergency diesel generator (EDG) control circuit. The design change was implemented to resolve relay contact problems experienced on Unit 3 that were attributed to low voltage conditions in the governor speed control circuit. The governor control circuit is required for the EDG to perform the safety related design function of automatically powenng loads during a loss of offsite power to achieve and maintain safe shutdown. The governor speed control unit utilized low voltage and current from an internal power supply for input control logic. To improve reliability of the governor unit, the entire control logic was rewired to 125-volt DC. New interposing relays were added to maintain the interface between the control logic and governor low voltage input. The new interposing relays have reed contacts that are wired inputs to the governor speed control unit. This EP did not change the way the EDG responds to any given input and enhances the design by utilizing contacts designed for low voltage to input to the governor speed control unit. An additional contact from the new idle/rated speed interposing relay was provided for the minimum fuel governor input. This additional contact was provided to enhance reliability as identified within the failure mode and effects analysis (FMEA) performed in the EP and does not change the operational response or control of the EDG.

10 CFR 50.59 Evaluation The EDG governor control circuits are required for the EDGs to perform the safety related design function of automatically powenng loads during a loss of offsite power to achieve and maintain safe shutdown. This EP was implemented to resolve relay contact problems experienced on Unit 3. The modifications added interposing relays and rewired the Unit 4 EDG governor control circuit. The EP reconfigures the circuit to permit the use of relay components that are better suited for low voltage applications. Elimination of this service-induced failure mechanism required that three new interposing relays be added to the control circuit. The addition of these interposing relays introduced new opportunities for an EDG failure. However, the failure rate of the new electrical relays is inherently very low, and the demand rate is very low. As demonstrated within the FMEA of the EP, the combination of these parameters resulted in a probability of failure that is lower than the current contact oxidation failure mechanism exhibited by the previous design. The circuit modifications lowered the likelihood of EDG failure. Therefore, the changes did not adversely affect plant safety or require a change to the plant technical specifications, and prior NRC approval was not required prior to implementation.

21

ATTACHMENT to L-2002-198 Page 22 of 66 PLANT CHANGE/MODIFICATION 00-043 UNIT : 3 TURN OVER DATE: 04/12/2001 EMERGENCY DIESEL GENERATOR GOVERNOR CONTROL CIRCUIT ENHANCEMENT - UNIT 3 Summary:

This Engineering Package (EP) provided a design to rewire the governor control logic and added three new interposing relays within the existing Unit 4 vital 125-volt DC emergency diesel generator (EDG) control circuit. The design change was implemented to resolve relay contact problems that were attributed to low voltage conditions in the governor speed control circuit. The governor control circuit is required for the EDG to perform the safety related design function of automatically powering loads during a loss of offsite power to achieve and maintain safe shutdown. The governor speed control unit utilized low voltage and current from an internal power supply for input control logic. To improve reliability of the governor unit, the entire control logic was rewired to 125-volt DC. New interposing relays were added to maintain the interface between the control logic and governor low voltage input. The new interposing relays have reed contacts that are wired inputs to the governor speed control unit. This EP did not change the way the EDG responds to any given input and enhances the design by utilizing contacts designed for low voltage to input to the governor speed control unit. An additional contact from the new idle/rated speed interposing relay was provided for the minimum fuel governor input. This additional contact was provided to enhance reliability as identified within the failure mode and effects analysis (FMEA) performed in the EP and does not change the operational response or control of the EDG.

10 CFR 50.59 Evaluation The EDG governor control circuits are required for the EDGs to perform the safety related design function of automatically powering loads during a loss of offsite power to achieve and maintain safe shutdown. This EP was implemented to resolve relay contact problems. The modifications added interposing relays and rewired the Unit 4 EDG governor control circuit. The EP reconfigures the circuit to permit the use of relay components that are better suited for low voltage applications.

Elimination of this service-induced failure mechanism required that three new interposing relays be added to the control circuit. The addition of these interposing relays introduced new opportunities for an EDG failure. However, the failure rate of the new electrical relays is inherently very low, and the demand rate is very low. As demonstrated within the FMEA of the EP, the combination of these parameters resulted in a probability of failure that is lower than the current contact oxidation failure mechanism exhibited by the previous design. The circuit modifications lowered the likelihood of EDG failure. Therefore, the changes did not adversely affect plant safety or require a change to the plant technical specifications, and prior NRC approval was not required prior to implementation.

22

Page 2 3 of 66 ATTACHMENT to L-2002-198 PLANT CHANGE/MODIFICATION 01-014 UNIT: 3 TURN OVER DATE: 10/20/2001 MOV-3-843A/B AND MOV-3-869 MODIFICATIONS SI SYSTEM ENHANCEMENTS - UNIT 3 Summary:

This Engineering Package (EP) made a series of modifications to the safety injection system (SIS) to reduce the likelihood of high head safety injection (HHSI) pump gas binding and to improve system and component reliability. The operation of the SIS is required to mitigate the consequences of the defined range of loss of coolant and secondary accidents by providing the emergency core cooling that is needed to protect the reactor core. The operation of motor operated valves MOV-3-843A, MOV-3 843B, and MOV-3-869 is necessary to allow the flow to the core. This EP revised the control logic of the three subject MOVs to close through the use of the limit switch in lieu of the torque switch to improve the leak tight seating characteristics of the valves. It also changed the actuator stems of the three MOVs to the "smartstem" design feature. This EP also relocated the equalizing line for MOV 3-843A and MOV-3-843B from a location upstream of the MOVs to a location downstream of the MOVs. The drilled disk in valve MOV-3-869 was replaced with a solid disk and a bonnet (inter-disk) equalizing line for MOV-3-869 was connected downstream of the MOV. The downstream side of these valves will be at a lower pressure than the upstream side during SIS actuation conditions. This arrangement provides a better vent/equalizing path. The EP also installed a one-inch check valve and test connection in each of the accumulator makeup lines, downstream of valves CV-3-851A/B/C. The only safety related function of the new check valves is to maintain RCS pressure boundary integrity.

The primary function was to prevent backflow of nitrogen saturated water toward the HHSI pumps.

10 CFR 50.59 Evaluation These modifications were evaluated and determined to not adversely affect the probability of occurrence or consequences of any accident or the malfunction of any equipment important to safety previously evaluated in the UFSAR. Similarly, these changes did not alter the function of the affected valves or introduce any new failure modes for valves or valve operators. Therefore, the modifications did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the UFSAR. Since no functional changes were made to any safety related structure, system or component, the modifications did not reduce the margin of safety as defined in the bases for any technical specification. As these modifications did not adversely affect safe operation of the plant or require a change to the plant technical specifications, prior NRC approval was not required for implementation.

23

ATTACHMENT to L-2002-198 Page 24 of 66 PLANT CHANGE/MODIFICATION 01-059 UNIT: 4 TURN OVER DATE: 04/04/2002 AFW BUS STRIPPING RESET MODIFICATION - UNIT 4 Summary:

This Engineering Package (EP) provided justification and instructions for revising the auxiliary feedwater (AFW) bus stripping actuating circuits such that the AFW auto start relays automatically reset after a sufficient time delay for the AFW pumps to have started. Previously, the AFW auto start relays did not reset when actuated from a bus stripping signal from the load sequencers until the start up transformer breaker was closed. The EP adds time delay and interposing relays in the circuitry that initiates the AFW auto start relays such that the auto start relays will reset after two minutes when these relays have been actuated from a bus stripping signal from the load sequencers. Resetting the autostart relays will not cause any valves to change position. The change improves the control room operators' ability to manage the operation of the AFW system and to more expeditiously shutdown AFW, when AFW operation was initiated from a stnpping signal.

10 CFR 50.59 Evaluation Bus stripping is an anticipatory start signal for loss of offsite power (LOOP) and is not the primary actuation signal for any design basis event described in the UFSAR. The LOOP event analysis is based on AFW start from a low-low steam generator water level signal. AFW pump start is required from a bus stripping signal by technical specifications in Modes 1, 2 and 3. This evaluation concluded that the modification did not impact the safe operation of the plant or require a change to the plant technical specifications, and thus did not require NRC approval pnor to implementation.

24

ATTACHMENT to L-2002-198 Page 25 of 66 PLANT CHANGE/MODIFICATION 01-063 UNIT: 4 TURN OVER DATE: 04/04/2002 ABANDONMENT OF HYDROGEN RECOMBINER EXHAUST LINE TO CONTAINMENT & REPLACEMENT OF CHECK VALVE 4-40-205 - UNIT 4 Summary:

Issuance of the NRC approved exemption to 10 CFR 50.44 eliminated the need for the hydrogen recombiner and PACV system at Turkey Point Units 3 and 4. Consistent with this, the PACV system was removed from the plant technical specifications by Amendments 217/211. This design change eliminated the portions of these systems that affected containment penetration 34, thereby reducing the amount of testing and maintenance required by this penetration. Penetration 34 provides service air supply to the containment. Service air is used for pneumatic tools for maintenance activities.

Service air is also used to pressurize containment for post-accident venting. Penetration 34 also provides a return path to containment from the hydrogen recombiner. This Engineering Package (EP) cut and capped the hydrogen recombiner return branch of the penetration and replaced the service air check valve 4-40-205 with a locked manual isolation valve. A UFSAR change package was provided with the EP to reflect changes made to containment penetration 34 and update the appropriate descriptions and design information.

10 CFR 50.59 Evaluation This EP abandoned in place the exhaust line from the hydrogen recombiner and reconfigured the containment isolation features for containment penetration 34, consistent with the NRC-approved exemption to 10 CFR 50.44 and issuance of Amendments 217/211 removing the PACV system from the Technical Specifications. The structures, systems and components affected by these changes are not accident or event initiators as described in the UFSAR. The modified valving arrangements on penetration 34 (replacing a swing check valve with a locked closed manual isolation valve and cutting and capping the hydrogen recombiner branch line) reduced the number of components necessary to maintain the containment function. The changes did not adversely affect the existing level of protection against the release of radioactivity to the outside atmosphere. The reconfigured penetration 34 satisfies the UFSAR single active failure criterion for containment isolation. The barriers credited for maintaining the isolation function consist of two manual isolation valves in series. Therefore, these modifications did not increase the frequency or likelihood of occurrence or result in more than a minimal increase in the consequences of an accident or malfunction of an SSC important to safety previously evaluated in the UFSAR. No new failure modes were created. Therefore, the change did not create the possibility for an accident or for the malfunction of any SSC important to safety of a different type or with a different result than any previously evaluated in the UJFSAR. The modification did not result in altering or exceeding a design basis limit for a fission product barrier as described in the UFSAR or in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses. Therefore, as these changes were performed based on receipt of the NRC exemption and approved plant technical specification changes, no further prior NRC approval was required for implementation.

25

ATTACHMENT to L-2002-198 Page 26 of 66 PLANT CHANGE/MODIFICATION 02-012 UNIT: 4 TURN OVER DATE: 04/04/2002 PERMANENT PLATFORMS / SCAFFOLDING / COMPONENTS IN CONTAINMENT Summary:

This Engineering Package (EP) provided justification and instructions for installing permanent structures (scaffolding platform frames, storage racks, fixed ladders, fencing and handrail) and components (storage barrels) in Unit 4 containment to replace temporary structures and components typically installed and removed during each outage. Scaffolding platform frames are used for performing maintenance on steam generators, reactor coolant pumps, feedwater nozzles, normal containment coolers and to connect temporary power. Storage racks are used for storage of scaffold poles that will remain in containment. The ladders are for use in accessing inspection ports on the steam generators. The fencing is used to prevent personnel from accessing the area around the reactor coolant drain tank (RCDT). The handrail is used to create a "bullpen" for personnel working on the steam generators. The barrels are used for storage of scaffolding knuckles used to build platforms during plant outages. This EP reduces dose and time spent in containment during plant outages by reducing transport, erection, and disassembly time for these structures and components.

10 CFR 50.59 Evaluation The potential adverse affects associated with seismic events, the potential impact on hydrogen generation, adverse affects on containment free volume and bulk material inventory affects related to heat sinks, potential fire hazards affects, containment sump interactions, jet impingement, post-LOCA flood levels, pressurization, air flow, thermal loads and secondary missile effects were considered in the design of these new structures. Restrictions imposed in this EP to ensure that the plant design bases with respect to seismic and high energy line break considerations were not compromised. There are no credible failure modes associated with the new structures and components that could adversely affect safety related structures, systems, or components. Therefore, since these changes do not adversely affect safe plant operation or require a change to the plant technical specification, NRC approval was not required prior to implementation.

26

ATTACHMENT to L-2002-198 Page 27 of 66 SECTION 2 10 CFR 50.59 EVALUATIONS 27

ATTACHMENT to L-2002-198 Page 28 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SEEJ-88-042 Revision 11 UNIT: 4 APPROVAL DATE: 3/11/2002 DE-ENERGIZATION OF UNIT 4 4160 VOLT SAFETY RELATED BUSES TO ALLOW FOR MODIFICATIONS OR PERIODIC MAINTENANCE Summary:

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 is de-energized and train "A" and "B" load centers are cross-connected. Also examined were technical and licensing concerns associated with de-energizing safety related equipment and effectively removing an emergency diesel generator (EDG) from service as the result of a Unit 4 4160 volt bus de energization. The de-energization of a Unit 4 4160 volt safety related bus, with Unit 4 in cold or refueling shutdown (Modes 5 and 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 related 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.

Revision 11 modified the evaluation to be consistent with changes to the 10 CFR 50.59 rule. The 10 CFR 50.59 Applicability and Screening form was added as an attachment and the eight questions now specified in the Rule were addressed. The revision assessed specific breaker position and loading changes associated with turbine lube oil (PC/M 00-01) and instrument air (PC/M 99-042) changes.

Use of the turbine turning gear drive motor circuit breaker was assessed to be acceptable only for intermittent operation in the manual mode. Use of the Main/Auxiliary transformer backfeed while replacing the Unit 4 Startup Transformer during the Unit 4 Cycle 20 refueling outage was also evaluated. While this option was always included, a provision was added to allow the main and auxiliary transformer cooling system to be loaded on an alternate MCC if the primary power source was not available. The load increase due to these changes was determined to be within the design rating of the load center transformer and the bus rating of 4160V bus 4A and 4B.

10 CFR 50.59 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 accident analyses addressed in the UFSAR or the results and conclusions of any previous 10 CFR50.59 evaluation. The actions or procedural changes identified and evaluated in this 10 CFR50.59 evaluation did not have any adverse affect on plant safety or plant operations and did not require changes to plant technical specifications.

Therefore, prior NRC approval was not required for implementation of the actions or precautions identified in this 10 CFR50.59 evaluation.

28

ATTACHMENT to L-2002-198 Page 29 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SENJ-89-084 Revision 2 UNIT: 3&4 APPROVAL DATE: 3/16/2001 SAFETY EVALUATION FOR REACTOR VESSEL MISSILE SHIELD REMOVAL DURING REDUCED RCS PRESSURE CONDITIONS Summary:

This 10 CFR 50.59 evaluation was developed to permit missile shield removal in Mode 4 for Rod Position Indication (RPI) maintenance. The purpose of this evaluation was to permanently update plant procedures to permit removal of the reactor vessel missile shield in Mode 4 or below when entering an outage with the reactor coolant system (RCS) pressure less than 500 psig. This evaluation also permitted the missile shield to remain off until the unit reached full temperature and pressure (up to and including Mode 3), when exiting a unit outage in which modifications were made to the RCS pressure boundary or where maintenance was performed on the control rod drive mechanisms (CRDMs) or the RPI system. The design basis of the CRDM missile shield is to block any missile generated by a control rod drive mechanism. This ensures the integrity of containment and ensures protection for systems or equipment required to maintain containment integrity. Control rod ejection accidents were analyzed in the FSAR for "at power" operation (Modes 1 and 2). Westinghouse also analyzed the effects of reactor coolant pressures/temperatures below operating values on reactor coolant system pipe ruptures; it was reported that calculations of critical flaw size for RCS pressures at or below 1000 psig showed that an RCS pipe rupture was not considered a credible event. This analysis concluded that material failure, and therefore, a rod ejection accident was not considered credible or should be postulated for Mode 4 operation.

Consistent with the Westinghouse analysis, Revision 2 permits removal of the missile shield when RCS pressure is reduced to less than or equal to 950 psig. It also requires missile shield replacement prior to exceeding 950 psig at the end of an outage if reactor vessel, reactor vessel head, or CRDM/RPI pressure boundary repairs are not made. The ability to defer missile shield replacement until prior to Mode 2 if reactor vessel, reactor vessel head, or CRDMIRPI pressure boundary repairs are made is retained.

10 CFR 50.59 Evaluation:

Movement of the reactor vessel missile shield complied with approved heavy loads handling procedures to minimize potential load drops. This lift would be similar to the movement of a missile shield following a refueling outage, which is explicitly permitted by the UFSAR. No new credible hazards were created. The actions or plant changes in procedures identified in this 10 CFR50.59 evaluation did not adversely affect plant safety or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.

29

ATTACHMENT to L-2002-198 Page 30 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SEES-89-085 Revision 15, Revision 16 UNIT: 3 APPROVAL DATE: 9/25/2001, Rev. 15 APPROVAL DATE: 10/05/2001, Rev. 16 DE-ENERGIZATION OF UNIT 3 4160 VOLT SAFETY RELATED BUSES Summary:

This evaluation developed 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 is de-energized and Train "A" and "B" load centers are cross-connected. Also examined were technical and licensing concerns associated with de-energizing safety related equipment and effectively removing an emergency diesel generator (EDG) from service as the result of a Unit 3 4160 volt bus outage. The de-energization of a Unit 3 4160 volt safety related bus, with Unit 3 in cold or refueling shutdown (Modes 5 and 6) or de-fueled and Unit 4 at power operation (Mode 1) or below, is sometimes necessary to permit 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 related 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.

Revision 15 modified the evaluation to be consistent with changes to the 10 CFR 50.59 rule. The 10 CFR 50.59 Applicability and Screening form was added as an attachment and the eight questions now specified in the Rule were addressed. The revision assessed specific breaker position and loading changes associated with turbine lube oil (PC/M 00-002) changes. Use of the Main/Auxiliary transformer backfeed while replacing the Unit 3 Startup Transformer during the Unit 3 Cycle 19 refueling outage was also evaluated. While this option was always included, a provision was added to allow the main and auxiliary transformer cooling system to be loaded on an alternate MCC if the primary power source was not available..

Revision 16 allowed the use of the turbine turning gear drive motor circuit breaker during modes 5 and 6, if required, only for intermittent operation in the manual mode and made the digital data processing system (DDPS) alternate power source available. These changes have no adverse affect on the load centers or the EDG loading.

10 CFR 50.59 Evaluation:

This 10 CFR 50.59 Evaluation addressed the technical and licensing requirements for the de energization of each Unit 3 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 accident analyses addressed in the UFSAR or the results and conclusions of any previous safety evaluations.

The actions or precautions identified and evaluated in this 10 CFR 50.59 Evaluation did not have any adverse effect on plant safety or plant operations and did require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or precautions identified in this 10 CFR 50.59 Evaluation.

30

ATTACHMENT to L-2002-198 Page 31 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SECS-90-018 Revision 1 UNIT: 3&4 APPROVAL DATE: 12/28/2000 SPENT FUEL POOL KEYWAY GATE BOOT SEAL REPLACEMENT Summary:

The inflatable seals on Units 3 & 4 Spent Fuel Pool (SFP) keyway gates are recommended for replacement every seven years to minimize the risk of leakage. The keyway gate and associated seal provide the required barrier between the SFP and the fuel transfer canal.

The replacement of the keyway seal is a direct one-for-one replacement and does not change the plant method of operation or configuration as described in the UFSAR. However, removal and reinstallation of the keyway gate is necessary for the replacement of the seal. The process to replace the seal constitutes a heavy load lift and requires a safety evaluation to assure compliance with the heavy load handling requirements of UFSAR and plant administrative procedures.

Revision 1 updated the evaluation to reflect the administrative controls, plant restrictions, and engineering analyses that have been implemented since the original evaluation was approved in 1990.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 Evaluation demonstrated that the removal and installation of the SFP keyway gate would not adversely impact plant safety or operation. The probability of occurrence of an accident previously evaluated in the UFSAR has not been increased since any accident involving the keyway gate is bounded by the fuel handling and cask drop accidents, which are addressed in the UFSAR.

Additionally, the defense in depth, safe load path, and heavy load lift requirements of UFSAR Appendix 51 are met by the provisions of providing two empty fuel racks from any edges of the gate, the use of a lateral restraint and rigging hardware with a safety factor of 10:1. It was concluded that the removal and installation of the keyway gate did not adversely affect plant safety and did not require a change to the plant technical specifications. Therefore, prior NRC approval was not required to remove and install the keyway gate.

31

ATTACHMENT to L-2002-198 Page 32 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SEMS-91-019 Revision 3 UNIT: 3 &4 APPROVAL DATE: 10/02/2001 RHR HEAT REMOVAL SYSTEM IN-SERVICE TESTING SAFETY EVALUATION Summary:

The original Turkey Point residual heat removal (RHR) system alignment utilized a common RHR pump minimum recirculation line for both pumps. Based on concerns regarding deadheading of safety related pumps at low flows while operating on a common recirculation line, Turkey Point installed individual mini-recirc lines on each RHR pump in 1987. The original recirculation line is currently used only for RHR pump standby mode testing. During a 1990 review of in-service testing procedures for the Component Cooling Water (CCW) and Intake Cooling Water (ICW) pumps, it was determined that entering the test alignments required the plant to enter a Limiting Condition for Operation (LCO) for the applicable system. This 10 CFR 50.59 evaluation examined the RHR pump in-service test procedure for similar concerns and provided two options to enhance the RHR pump in service test procedure. Option 1 closes both mini-recirc lines and uses the common (test) recirc line and a dedicated operator, ensuring that the RHR pumps are operable during in-service testing and that the RHR system is capable of meeting its design basis accident requirements while assuming a worst case single failure (i.e., an LCO is not required). Option 2 opens both mini-recirc lines and the common (test) recirc line as well as a dedicated operator. The potential for deadheading an RHR pump is precluded because the mini-recirculation lines for both pumps are open throughout the test.

However, this results in a net reduction in flow to the core during an accident and required further review. Additionally, the in-service testing criteria will require revision because of the higher test flows (i.e., slightly lower pump head developed).

Revision 3 revised the position on entering a technical specification LCO action statement when performing the RHR pump in-service test when Option 1 is utilized. The manual actions required by plant procedures during this test are appropriate and no discrepancies exist with the governing procedures and administrative controls used to perform the in-service test and system venting. The recommendation to enter the action statement for an inoperable RHR pump (preferably the pump being tested) is intended to heighten awareness that the RHR pump is in an off-normal configuration and must be returned to normal alignment at the onset of an accident. Revision 3 also updated the evaluation to be consistent with the revised 10 CFR 50.59 requirements and evaluation criteria.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation demonstrated that there was no safety issue during previous pump testing and the procedures and restrictions currently used for RHR pump in-service testing do not impact safe plant operation or adversely affect the technical specifications. The use of the alignment described in Option 1 does not increase the likelihood of a malfunction of equipment important to safety as previously evaluated in the UFSAR. The resulting configuration of the RHR system is capable of performing its intended safety function. Deadheading of an RHR pump during testing is precluded procedurally for non-accident situations and the impact of RHR pump deadheading is limited procedurally for accident conditions to maximize component availability. A dedicated operator ensures that corrective actions are completed before the potentially deadheaded pump could be damaged. It was concluded that the evaluated procedure changes did not require prior NRC approval or require a change to plant technical specifications.

32

ATTACHMENT to L-2002-198 Page 33 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SEMS-93-059 Revision 2 UNIT: 3 &4 APPROVAL DATE: 3/06/2001 CONNECTION OF A TEMPORARY FILTER ASSEMBLY TO THE DIESEL FUEL OIL STORAGE TANKS Summary:

This safety evaluation was developed to support the technical justification and basis for a temporary filtering system that was used to reduce the sediment concentrations in the diesel oil storage tanks (DOSTs) to as low as practical. Technical specifications require that the DOST be periodically sampled to verify that fuel viscosity, water and sediment levels are within acceptable limits. Sample results at the time indicated that sediment concentrations were acceptable, but close to the limit (< 10 mg/liter). These actions provided time for plant personnel to review long term solutions to the sediment problem. The temporary filtering system consisted of a pump, filters, valves and hoses to connect the system to existing drain and fill connections on the DOST. This temporary filtering system was installed and operated in accordance with an approved plant procedure.

Revision 1 revised the evaluation to reflect the applicability to Unit 4. This revision added a plant restriction for the minimum DOST level for Unit 4 filtering operations to allow for operator action time.

Revision 2 incorporated changes as a result of calculation revisions that changed the minimum DOST level for unit 4 filtering operations.

10 CFR 50.59 Evaluation:

The safety function of the diesel fuel oil storage tank has not been affected, since there is sufficient time for an operator to isolate the temporary filter assembly from the tank prior to reaching a technical specification limit. In addition, the connection of the filter assembly to the tank did not adversely impact the DOSTs. A dedicated operator was required to be stationed any time the filtering skid was connected to the DOST to restore the emergency diesel generator fuel system pressure in the event of a temporary hose/filter equipment failure. No other interaction with equipment important to safety was affected. The actions or plant changes (procedures and/or hardware) identified in this 10 CFR 50.59 evaluation did not adversely affect plant safety or operations, and did not require changes to the plant technical specifications. Therefore, pnor NRC approval was not required for implementation of the actions or changes identified within this evaluation.

33

ATTACHMENT to L-2002-198 Page 34 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SENP-95-007 Revision 5 UNIT: 3 APPROVAL DATE: 10/15/2001 OPERABILITY OF RHR AND REFUELING SUPPORT EQUIPMENT DURING INTEGRATED SAFEGUARDS TESTING Summary:

This safety evaluation reviewed the Unit 3 engineered safeguards integrated test (ESIT) procedures with respect to a generic Westinghouse concern related to the effectiveness of the steam generators (S/Gs) to remove decay heat during shutdown conditions. Westinghouse identified that there was a potential for gas formation within the steam generator U-tubes under certain reactor coolant system (RCS) pressure and level conditions in Mode 5 that could inhibit the ability to establish natural circulation cooling. To accommodate the potential unavailability of the S/Gs for decay heat removal under these conditions, plant technical specifications require that both trains of the residual heat removal system (RHR) be operable in Mode 5 when the RCS is in a "loops not filled" configuration.

Since safeguards testing was normally performed during Mode 5 with the RCS depressunzed and partially drained, this evaluation was developed to document that both trains of the RHR system would remain operable during the test period. The evaluation concluded that no restrictions on plant operations or additional operator actions, other than those already prescribed in the ESIT procedures, were required to ensure RHR operability.

Revision 5 expanded the scope to address operability of the refueling support equipment during performance of the ESIT in Mode 6 during core reload and updated the evaluation to reflect the new requirements and criteria associated with the revised 10 CFR 50.59. The evaluation concluded that performance of the ESIT in Modes 5 and 6 would not adversely affect plant operation and would not compromise the fuel handling accident analysis, provided the actions and restrictions identified in the evaluation are observed.

10 CFR 50.59 Evaluation:

This 10 CFR 50.59 evaluation examined the electrical, mechanical, and hydraulic configuration of the plant during performance of the ESIT in Modes 5 (loops not filled) and 6 (vessel level two feet below the flange). Actions and limitations were identified to ensure that both RHR loops would remain operable during the test sequence. Appropriate actions and limitations were also identified for the refueling support equipment to ensure core reload could be conducted during portions of the ESIT.

The evaluation also examined heat removal capability, tube vibration, thermal stress, and pressure boundary integrity of a component cooling water (CCW) heat exchanger if it is operated without intake cooling water (ICW) flow during the test. The actions and limitations identified and evaluated in this 10 CFR 50.59 evaluation did not have any adverse affect on plant safety or operations. No new failure modes were created. Since all licensing and design basis requirements would continue to be met during the ESIT and the proposed activity did not require changes to plant technical specifications, prior NRC approval was not required to initiate the test sequences.

34

ATTACHMENT to L-2002-198 Page 35 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SENP-95-023 Revision 5, Revision 6 UNIT: 4 APPROVAL DATE: 3/25/2002, Rev. 5 APPROVAL DATE: 3/28/2002, Rev. 6 OPERABILITY OF PLANT EQUIPMENT DURING INTEGRATED DURING INTEGRATED SAFEGUARDS TESTING Summary:

This safety evaluation reviewed the Unit 4 engineered safeguards integrated test (ESIT) procedures with respect to a generic Westinghouse concern related to the effectiveness of the steam generators (S/Gs) to remove decay heat during shutdown conditions. Westinghouse identified that there was a potential for gas formation within the steam generator U-tubes under certain reactor coolant system (RCS) pressure and level conditions in Mode 5 that could inhibit the ability to establish natural circulation cooling. To accommodate the potential unavailability of the S/Gs for decay heat removal under these conditions, plant technical specifications require that both trains of the residual heat removal system (RHR) be operable in Mode 5 when the RCS is in a "loops not filled" configuration.

Since safeguards testing was normally performed during Mode 5 with the RCS depressurized and partially drained, this evaluation was developed to document that both trains of the RHR system would remain operable during the test period. The evaluation concluded that no restrictions on plant operations or additional operator actions, other than those already prescribed in the ESIT procedures, were required to ensure RHR operability.

Revision 5 expanded the scope to address operability of the refueling support equipment during performance of the engineered safeguards integrated test in Mode 6 during core reload and updated the evaluation to reflect the new requirements and criteria associated with the revised 10 CFR 50.59.

The evaluation concluded that performance of the ESIT in Modes 5 and 6 will not adversely affect plant operation and will not compromise the fuel handling accident analysis, if the actions and restrictions identified in the evaluation are observed. Revision 6 expanded the scope to address operability of the refueling support equipment during performance of the engineered safeguards integrated test in Mode 6 during core off-load. It also updated the evaluation for consistency regarding power supply requirements for the refueling communication equipment and clarified assumptions made in Revision 4 of this evaluation regarding component cooling water (CCW) heat exchanger operation without shell-side flow.

10 CFR 50.59 Evaluation:

This 10 CFR 50.59 evaluation examined the electrical, mechanical, and hydraulic configuration of the plant during performance of the ESIT in Modes 5 (loops not filled) and 6 (vessel level two feet below the flange). Actions and limitations were identified to ensure that both RHR loops would remain operable during the test sequence. Appropriate actions and limitations were identified for the refueling support equipment to ensure core reload could be conducted during portions of the ESIT. The evaluation also examined heat removal capability, tube vibration, thermal stress, and pressure boundary integrity of a component cooling water (CCW) heat exchanger if it is operated without intake cooling water (ICW) flow during the test. The actions and limitations identified and evaluated in this 10 CFR 50.59 evaluation did not have any adverse affect on plant safety or operations. No new failure modes were created. Since all licensing and design basis requirements would continue to be met during the ESIT and the proposed activity did not require changes to plant technical specifications, prior NRC approval was not required to initiate the test sequences.

35

ATTACHMENT to L-2002-198 Page 36 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SEMS-96-003 Revision 4 UNIT: 4 APPROVAL DATE: 12/05/2000 SAFETY EVALUATION FOR UNIT 4 STEAM GENERATORS' SECONDARY SIDE FOREIGN OBJECTS Summary:

This evaluation addressed the potential safety significance of operating the Unit 4 steam generators (S/Gs) with irretrievable foreign objects present in the secondary side. Previously, individual safety evaluations addressed the acceptability of continued Unit 4 operation with foreign objects remaining in the S/Gs and associated systems. The purpose of this evaluation was to: (1) re-examine the analyses, results, requirements, and restrictions of previous evaluations while applying recent industry standards; (2) document the methodology for determining the interval between S/G eddy current tests as affected by estimated S/G tube wall wear times; and (3) provide a single Unit 4 safety evaluation to assess and document all of the Unit 4 S/G foreign object estimated wear times as adjusted by updated S/G eddy current data and steam generator Foreign Object Search and Retrievals (FOSAR) results.

FPL maintains a visual inspection program of the secondary side of S/Gs (in addition to the other inspection programs for S/Gs) to help prevent and detect the presence of loose parts.

Revision 4 incorporated results of the S/G inspections performed during the Cycle 19 refueling outage, which included both secondary side FOSAR inspections and eddy current test (ECT) examination of the full length of approximately 50% of active tubes, including 100% of hot leg tubes above the tube sheet. The inspections did not find any of the previous foreign objects but did identify two additional objects (one of which was not retrievable) and unassociated tube damage or degradation that necessitated tube plugging. The partial ECT performed during the outage was not adequate to reassess the wear time to minimum tube wall thickness for the existing foreign objects; however, this revision did reassess the wear time for object number 4 based on the FOSAR visual examination results and incorporated the new unretrieved object. This revision continues to identify the most restrictive requirement for future inspection as November 2003.

10 CFR 50.59 Evaluation:

Previous 10 CFR 50.59 evaluations documented for each S/G secondary side foreign object have considered the effects of the object upon tube integrity, chemistry, S/G instrumentation, the main steam system, and S/G blowdown and sampling systems. This evaluation established wear time to minimum tube wall thickness estimates based on conservative assumptions from Westinghouse WCAP-14258 and associated clarification correspondence. These wear times assume worst case conditions and actual wear times are likely to be much greater than the WCAP methodology would predict. Based on this assessment, this evaluation determined that currently identified foreign objects within the secondary side of the Unit 4 S/Gs do not adversely affect plant safety or operation and do not require changes to the plant technical specifications. Therefore, prior NRC approval was not required to implement the actions identified within this evaluation.

36

ATTACHMENT to L-2002-198 Page 37 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SEMS-96-014 Revision 3 UNIT: 3 &4 APPROVAL DATE: 12/12/2000 A TEST OF THE USE OF SUB-MICRON ULTRAFINE FILTERS IN THE CVCS AND SFP Summary:

This evaluation served to allow the temporary use of Ultrafine filter cartridges with absolute filtration ratings in the reactor coolant system (RCS), seal water injection, and seal water return filters in the chemical volume and control system to reduce plant radiation levels and to extend the life of reactor coolant pump seals. Implementation of the Ultrafine filter program was planned in three phases.

Phase I involved demonstration of proper fit and performance when installed side-by-side with a conventional filter having a similar absolute rating. Phase II of the testing program involved a gradual reduction in the absolute rating of the filters used. This would gradually filter out finer and finer particles as the overall RCS particulate inventory is reduced. This reduction would continue until the desired RCS cleanliness level is reached. Phase Ell involved the permanent use of these filters under formal plant design change documentation. Phase I of the program was evaluated in a previous safety evaluation. This evaluation was issue to address Phase II of the Ultrafine filter program.

Revision 3 implements Phase IlI of the program which evaluated and allows the use of Ultrafine filters on a permanent basis.

10 CFR 50.59 Evaluation:

This evaluation addressed the use of Ultrafine filter cartridges for the RCS, seal water return, seal water injection, and spent fuel pool filters. This evaluation concluded that these Ultrafine filters meets all current design criteria for the systems identified above. Failure modes were evaluated to ensure that the probability of occurrence and consequences of previously analyzed failures are not increased. The use of these filters does not change system design bases, functions, or operation of any safety related equipment, and does not adversely affect any safety related structures, systems or components. Therefore, the testing, implementation and plant actions identified in this 10 CFR 50.59 evaluation did not adversely affect plant safety or operation or require changes to the plant technical specifications. Thus, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.

37

ATTACHMENT to L-2002-198 Page 38 of 66 10 CFR 50.59 EVALUATION JPN-PTN-SEMS-96-038 Revision 4 UNIT: 3 APPROVAL DATE: 2/14/2001 SAFETY EVALUATION FOR UNIT 3 STEAM GENERATORS' SECONDARY SIDE FOREIGN OBJECTS Summary:

This evaluation addressed the potential safety significance of operating the Unit 3 steam generators (S/Gs) with foreign objects present in the secondary side. The foreign objects identified within the scope of this evaluation are those which are considered to be irretrievable. Previously, individual safety evaluations addressed the acceptability of continued Unit 3 operation while these foreign objects remained in the S/Gs and associated systems. The purpose of this evaluation was to: (1) re examine the analyses, results, requirements, and restrictions of previous evaluations while applying recent industry standards; (2) document the methodology for determining the interval between S/G eddy current tests as affected by estimated S/G tube wall wear times; and (3) provide a single Unit 3 10 CFR 50.59 Evaluation to assess and document all of the Unit 3 S/G foreign object estimated wear times as adjusted by updated S/G eddy current data and S/G Foreign Object Search and Retrievals (FOSAR) results.

Revision 4 incorporated comments on Revision 3 including a correction to the eddy current test (ECT) scope completed during the Unit 3 Cycle 18 refueling outage. Full length ECT was completed on only 50% of the S/G tubes. Therefore, revised wear times could not be calculated for all foreign objects. Wear times for the first three objects (3A S/G) were restored to previously reported values based on the 100% ECT conducted in October 1998 and adjusted to account for unit outages. The changes did not impact ECT inspection plans for upcoming outages.

10 CFR 50.59 Evaluation:

Previous 10 CFR 50.59 evaluations prepared for each S/G secondary side foreign object have considered the effects of the object upon tube integrity, chemistry, S/G instrumentation, the main steam system, and S/G blowdown and sampling systems. This evaluation established current wear time to minimum tube wall thickness estimates based on conservative assumptions from Westinghouse WCAP-14258 and associated Westinghouse clarification correspondence. These wear times assume worst case conditions and actual wear times are likely to be much greater than the Westinghouse methodology would predict. Based on this assessment, this evaluation determined that currently identified foreign objects within the secondary side of the Unit 3 S/Gs did not adversely affect plant safety or operation and did not require changes to the plant technical specifications.

Therefore, prior NRC approval was not required for continued operation of the plant with foreign objects present in the secondary side of the S/Gs, or endorsement of the programmatic actions identified within this evaluation.

38

ATTACHMENT to L-2002-198 Page 3 9 o f 66 10 CFR 50.59 EVALUATION JPN-PTN-SEMS-96-040 Revision 1 UNIT: 3 &4 APPROVAL DATE: 2/26/2001 SAFETY EVALUATION FOR THE TEMPORARY INSTALLATION OF DRAIN HOSES AND PERFORMANCE OF HOT-SPOT FLUSHES ON THE RHR SYSTEM Summary:

This 10 CFR 50.59 evaluation examined the procedure for flushing the residual heat removal (RHR) system with water from the refueling water storage tank to eliminate radioactive hot spots at various system drain locations. The affected drains were located in the suction piping from the refueling water storage tank and south containment recirculation sump, and at the RIR heat exchangers. The flushes were performed by installing a flush adapter and tygon hose to the discharge of each drain valve, routing the hose to the nearest suitable floor drain, and opening the drain valve for approximately 20 seconds. The affected drain valves were flushed one at a time while the RHR system remained operable in the normal standby valve lineup (Modes 1 - 3). A flushing flow rate of 45 gpm was expected through the drain piping based on the static head of the refueling water storage tank. As a precautionary measure, an additional ball valve was used with the flush adapter on the refueling water storage tank suction piping drains, to provide a backup isolation capability (in lieu of RWST isolation) in the event that the piping drain valve could not be re-closed when the flushing activity was complete.

Revision 1 of this evaluation provides flushing requirements for additional RHR piping system drain valves that are similar in configuration to the drain valves evaluated in Revision 0 of the evaluation.

10 CFR 50.59 Evaluation:

This evaluation addressed the temporary configuration of the system with the installed flushing adaptor, the impact on plant operation, and the various precautions imposed to ensure safe conduct of the maintenance activity. Stnct controls were imposed on the flushing process and contingency measures were developed to establish pressure boundary integrity for the open system should a drain valve fail to re-close, or actuation of the engineered safety features occur. The analyses, evaluations and implementation instructions supporting this activity ensured that no safety related systems, equipment or structure was adversely affected by the performance of the flushes on the RHR system as specified in the temporary procedures. Based on the precautions identified, the evaluation concluded that the maintenance activity could be performed in Modes 1 - 3, and that the activity did not adversely affect plant safety or operation or require changes to the plant technical specifications.

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

39

ATTACHMENT to L-2002-198 Page 40 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SEMS-97-002 UNIT: 3 &4 APPROVAL DATE: 1/23/2001 IMPACT OF CURE TIME ON THERMO-LAG FIRE BARRIER PERFORMANCE Summary:

Thermo-Lag fire barrier raceway protection systems are upgraded, installed or repaired in accordance with a Turkey Point specified installation and inspection guidelines. The specification requires a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cure time for Thermo-Lag trowel-graded materials. This evaluation was performed to determine if the requirement to maintain a continuous fire watch during the curing period could be eliminated.

The UFSAR states that while a fire barrier is inoperable, either a continuous fire watch must be established and maintained or, with verification of fire detection operability, an hourly watch may be established and maintained. A nonconforming fire barrier is considered inoperable until its final configuration is determined to be consistent with either a tested assembly or one of equivalent fire resistance. Thermo-Lag fire barriers may be upgraded and declared operable when installation to a rated configuration, 72-hour cure time and final inspection are completed. This evaluation shows that the inherent fire resistance performance capability of an installed Thermo-Lag fire barrier with uncured trowel-grade material provides a sufficient level of protection as to be functional so as to justify eliminating the continuous fire watch. In correspondence addressing NRC Bulletin 92-01, the NRC accepted Turkey Point compensatory measures where any fire zone containing Thermo-Lag 330-1 fire barrier matenal was provided with either one hour roving fire watches with automatic fire detection or one hour fire watches with closed circuit TV monitored on a continuous basis. These compensatory measures remained in place until corrective actions were taken to assure the operability of these fire barriers.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation demonstrated that the uncured Thermo-Lag trowel-grade material fire barrier condition provides a functionally equivalent level of fire protection as the cured Thermo-Lag fire barrier. As such, eliminating continuous fire watch coverage during cure time does not compromise the ability of any structure, system, or component to perform its respective safety related function. No new hazards are created by the conditions or actions evaluated in this 10 CFR 50.59 evaluation. Therefore, procedure changes to eliminate such fire watch coverage can be performed without prior NRC approval. However, final inspection after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is maintained for added assurance that the cured trowel-grade material is suitable.

40

ATTACHMENT to L-2002-198 Page 41 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SECS-98-058 Revision 2, Revision 3 UNIT: 3 &4 APPROVAL DATE: 10/17/2001, Rev. 2 APPROVAL DATE: 10/25/2001, Rev. 3 EVALUATION FOR STORAGE OF TOOLS AND EQUIPMENT IN CONTAINMENT DURING ALL MODES OF OPERATION Summary:

This evaluation addressed the acceptability of leaving a quantity of tools and equipment within the Unit 3 containment structure during all modes of plant operation. The items to be stored, and the storage locations within the Unit 3 containment, were specifically identified within the evaluation.

The purpose of leaving these tools and equipment within containment following refueling outages was to reduce the usage demand on the Unit 3 polar crane during refueling outages. This evaluation considered the potential for adverse seismic interactions with safety related equipment, the potential for additional hydrogen generation within containment during accidents, the impact on the containment free volume and heat sink analyses, the potential to obstruct flow to the containment sumps, and the impact on containment combustible loading. To ensure that the tools and equipment addressed in the evaluation were safely stored during plant operation, both generic and specific actions and restrictions were identified for implementation within the evaluation.

Revision 2 addressed the permanent storage of four (4) storage boxes containing scaffold tubes, twelve (12) 55-gallon drums containing scaffold knuckles and seventy-five (75) 13-foot long scaffold poles in containment. This revision concludes that these additional items can remain within the unit 3 containment structure during all modes of operation provided all stipulated requirements are followed. The revision also deleted reference to materials addressed in a previous revision as they have since been removed.

Revision 3 addressed the temporary storage of two (2) CRDM coil stack assemblies within containment until the next outage of sufficient duration to remove them. This revision concludes that these additional items can remain within the Unit 3 containment structure during all modes of operation provided all stipulated requirements are followed.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation concluded that the identified items can safely remain within containment during all modes of operation, provided that all of the restrictions and requirements identified within the evaluation were implemented following each outage. The evaluation further concluded that the identified restrictions and requirements would ensure that these activities would have no adverse effects on plant operation, and would not compromise the safety and licensing bases of the plant. Consequently, the requirements and restrictions identified in this 10 CFR 50.59 evaluation did not adversely affect plant safety or require changes to the plant technical specifications.

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

41

ATTACHMENT to L-2002-198 Page 42 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SEFJ-00-026 UNIT: 3 &4 APPROVAL DATE: 1/18/2001 UFSAR UPDATE TO OFFSITE RADIOLOGICAL DOSES DUE TO CHANGES IN ASSUMPTIONS ASSOCIATED WITH STEAM GENERATOR TUBE RUPTURE AND STEAM LINE BREAK Summary:

Westinghouse's Nuclear Safety Advisory Letter NSAL-00-004 reported that non-conservative assumptions had been used in the calculation of the accident-initiated iodine spiking rates in the reactor coolant system (RCS). These iodine spiking rates were used to calculate the offsite radiological doses of the Main Steam Line Break (MSLB) and the Steam Generator Tube Rupture (SGTR) design basis accidents for Turkey Point.

The offsite radiological doses resulting from the MSLB and SGTR accidents have been reanalyzed to address NSAL-00-004 with the following changes in input assumptions:

"* Letdown Flow: Increased to 132 gpm to replace the previous value of 60 gpm.

"* Letdown Cleanup: Increased to 100% to replace the previous value of 90%.

"* Leakage from RCS during Normal Operation: Increased to 11 gpm. This effect was not considered in the previous analyses.

These input assumptions are more conservative and result in a higher equilibrium value for the iodine release to the primary coolant and correspondingly faster iodine spike. The consequence of these more conservative assumptions is a slightly higher offsite thyroid dose for each accident. The standard review plan (SRP) and 10 CFR 100 acceptance criteria for thyroid dose is the licensing basis for these events and the revised dose values are well below the acceptance criteria.

A UFSAR change package was provided as an attachment to this 10 CFR 50.59 evaluation. It revised the dose values for the MSLB and SGTR events.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation demonstrated that substantial margin to the offsite thyroid dose acceptance criteria of 30 rem as given in the NRC SRP and 10 CFR 100 is maintained for both the SGTR and MSLB accidents. Changes to the offsite doses reported in the UFSAR change package, associated with the reanalysis of the MSLB and SGTR accidents, have been determined not to have an adverse effect on safe plant operation and not to require a change in the technical specifications.

Therefore, NRC approval is not required prior to the implementation of this FSAR change package.

42

ATTACHMENT to L-2002-198 Page 43 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SENS-00-046 Revision 2 UNIT: 4 APPROVAL DATE: 5/10/2001 SAFETY EVALUATION FOR TEMPORARY LOWERING OF UNIT 4 SPENT FUEL POOL WATER LEVEL FOR MAINTENANCE ACTIVITIES Summary:

This evaluation was developed to examine the effects of securing the spent fuel cooling pumps and reducing the pool level by about 1-foot in order to perform maintenance on valve 4-821 in the primary water system makeup line to the spent fuel pool (SFP). This evaluation addressed the effects of spent fuel handling accidents, spent fuel heatup rates, increased radiation levels resulting from lowered water (shielding) levels, and activation of system alarms. To reduce the potential for fuel handling accidents, all fuel movement and crane operation were suspended in accordance with Technical Specification 3/4.9.11. Considering the amount of decay time that has elapsed since the previous refueling (Cycle 18), pool heatup from 100 'F to 135 'F was estimated to take about 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, which would be a sufficient time to perform the required maintenance. Previous evaluations of reduced water levels have demonstrated that expected increases in radiation levels would be negligible. In order to preclude activation of the SFP alarms, pool temperature and level were required to be monitored on an hourly basis. A SFP temperature limit of 130 'F was established as an upper limit during the maintenance activity, at which time work would be secured and SFP cooling restored.

Revision 2 allowed the work to be performed during operating Cycle 19. The revision incorporated actual heatup data compiled in 2001 and updated the associated restrictions and required actions identified in the evaluation.

10 CFR 50.59 Evaluation:

This evaluation concluded that reducing the spent fuel pool level for maintenance on the primary water makeup valve would not adversely impact plant operation and would not compromise the spent fuel handling accident analyses, provided that the actions and restrictions identified in the evaluation were observed. Consequently, the reduced pool water level and other actions identified in this safety evaluation did not adversely affect plant safety or require changes to the plant technical specifications.

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

43

44 of 66 ATTACHMENT to L-2002-198 Page 10 CFR 50.59 EVALUATION PTN-ENG-SENS-00-088 UNIT: 3 &4 APPROVAL DATE: 11/21/2000 SAFETY EVALUATION FOR THROTTLING CCW MANUAL VALVES 3/4-737A Summary:

This 10 CFR 50.59 evaluation was prepared to analyze the impact on plant safety and operation associated with throttling normally open component cooling water (CCW) system manual valve 3/4 737A. Throttling this valve is considered a necessary design enhancement to prevent excessive CCW system process fluid loss in the event relief valve RV-3/4-715 should lift and fail to reseat as witnessed during a previously reported loss of offsite power (LOOP) event on Unit 4 in October 2000.

The manual valve 3/4-737A is now maintained throttled during normal operation to limit CCW flow to the excess letdown heat exchanger to a maximum of 50 gpm which is the maximum system leak rate evaluated in the UFSAR. To provide additional assurance of CCW system integrity when excess letdown is in service, procedures now require a designated operator to be available to manually close valve 314-737A on receipt of a containment isolation phase "A" signal. Procedures also now require slow re-initiation of CCW to the excess letdown heat exchanger in the event CCW is isolated.

10 CFR 50.59 Evaluation:

This 10 CFR 50.59 evaluation addressed the impact on plant safety associated with throttling normally open CCW system manual valve 3/4-737A. Throttling this valve is considered a necessary design enhancement to prevent excessive CCW system process fluid loss in the event relief valve RV 3/4-715 should lift and fail to reseat as witnessed during a previously reported LOOP event on Unit 4 in October 2000. Excess letdown provides no safety related function and is not credited in shutting down the reactor or maintaining the reactor in a safe shutdown condition. It is not relied on to mitigate the consequences of accidents identified in the UFSAR. The evaluation concluded that throttling valve 3/4-737 such that CCW flow rate will not exceed 50 gpm while excess letdown is in its normal standby lineup, and with conditions similar to the LOOP event reported previously, will preclude excessive inventory loss while maintaining system integrity. This is consistent with the primary water system design capability to provide a fill rate to the CCW head tank greater than 50 gpm.

The actions and precautions identified in this 10 CFR 50.59 evaluation did not adversely affect plant safety or operation or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the identified actions and requisite precautions.

44

Page 4 5 of 66 ATTACHMENT to L-2002-198 10 CFR 50.59 EVALUATION PTN-ENG-SEFJ-01-006 UNIT: 3 &4 APPROVAL DATE: 5/10/2001 REACTOR COOLANT SYSTEM CHEMISTRY pH CONTROL Summary:

This 10 CFR 50.59 evaluation was prepared to examine operation of Turkey Point Units 3 and 4 with higher reactor coolant system (RCS) pH. Control of pressurized water reactor (PWR) primary water chemistry is essential to maximize fuel and materials integrity and minimize plant radiation fields in component/access areas where maintenance may be required. To achieve these goals, critical parameters for control include both oxygen and pH. Hydrogen is used to scavenge dissolved oxygen in the coolant during normal operation while pH is maintained within a prescribed range via lithium (Li) in the form of lithium hydroxide. Higher pH operation minimizes the rate of crud deposition, while the possibility of accelerating cladding corrosion in regions of the core where subcooled boiling occurs increases as local lithium concentrations increase. Higher lithium concentrations can also lead to an increased likelihood of Alloy 600 cracking. Therefore, the chemistry regime used must achieve a balance to assure primary system pressure boundary integrity, fuel cladding integrity and minimization of out-of-core radiation fields.

This evaluation supports the corporate objectives established in Nuclear Policy NP-916, Nuclear Plant Chemistry Parameters. This policy emphasizes the importance of establishing and maintaining appropriate water chemistry conditions in the nuclear power plants and supports a chemistry program of "minimizing in-plant radiation fields and releases of radioactivity to the environment." The policy states that "both nuclear plants shall implement a chemistry program that reflects the primary and secondary chemistry recommendations and guidelines developed by the NSSS vendors and recognized industry groups (i.e., EPRI, INPO)..." In this evaluation, industry experience with operation at high RCS pH has been considered in evaluating the safety significance of the proposed chemistry program at Turkey Point in the context of the NP-916 requirements. As such, the proposed operation at a higher pH has been evaluated by Westinghouse (the NSSS and fuel vendor) and reflects the primary chemistry recommendations and guidelines developed by recognized industry groups.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation demonstrated that no new hazards were created by the modified pH control program. This evaluation concluded that higher RCS pH operation may be effectively and acceptably implemented at Turkey Point. The RCS chemistry modified boron-lithium pH control program is acceptable with regard to the general corrosion of cracking of primary side materials including those in the steam generators. The evaluation determined that with proper cycle-specific core design, the fuel rod design criteria for Zircoloy-4 cladding fuel is met. Since the RCS pressure boundary integrity and system function are not adversely affected, the probability of occurrence of an accident evaluated in the UFSAR will be no greater than the original design basis of the plant. RCS operation at a higher pH does not adversely affect plant safety or operation or require a modification of the plant technical Specifications. Therefore, NRC approval is not required prior to the implementation of this RCS chemistry change.

45

ATTACHMENT to L-2002-198 Page 46 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SENS-01-024 UNIT: 3&4 APPROVAL DATE: 3/22/2001 SAFETY EVALUATION FOR OFFSITE DOSE CALCULATION MANUAL REVISION 9 Summary:

The Offsite Dose Calculation Manual (ODCM) describes acceptable methods for calculating radioactivity concentrations in the environment and potential offsite doses associated with liquid and gaseous effluents from Turkey Point nuclear units. The calculations are performed to satisfy technical specifications and to ensure that allowable doses to the public are not exceeded. The ODCM identifies sampling locations for offsite releases and for the environmental monitoring program. This evaluation reviewed editorial changes to the ODCM as well as clarifications associated with the single analysis basis on which radionuclide quantities are calculated, units of measure for radionuclides in fish, the erroneous identification of a lower limit of detection, and the erroneous description of a sampling point. One sampling location was changed as a result of the previous location becoming inaccessible. The new location was selected on the bases of accessibility, proximity to the old site and compliance with NUREG-0472 requirements. A supplemental sampling point was changed for sampling milk for the ingestion pathway. This location is not one of the required sampling points and is routinely updated based on availability of the milk sample. Another clarification was also made on the accuracy of the identified location of another sampling point.

Based on a Global Positioning System (GPS) used during sampling, the sampling location designated as W could be W or WNW depending on where vegetation is available for gathering. The change was made to ensure correct identification of the sampling point.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation demonstrated that the ODCM revision maintained the level of radioactive effluent control required by federal regulations unchanged (i.e., 10 CFR 20.1302, 10 CFR 50.36a, 10 CFR 50, Appendix I and 40 CFR 190), and did not adversely affect the accuracy or reliability of effluent, dose or setpoint calculations. The technical specifications do not state any specific requirements associated with sampling locations or calculation methods. The ODCM changes were not associated with plant systems and therefore could not create any new failure modes within the power plant. Analytical methods remained unchanged and therefore the dose calculations were not adversely affected by the changes. The evaluation determined that the ODCM changes did not adversely affect safe operation of the plant and did not require a change to technical specifications. Therefore, this activity did not require prior NRC approval.

46

ATTACHMENT to L-2002-198 Page 47 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SEMS-01-025 UNIT: 3&4 APPROVAL DATE: 5/11/2001 ADDITION OF ALTERNATE AMINES TO THE SECONDARY SIDE SYSTEM Summary:

This evaluation provided the technical justification to use various alternate amines in the secondary system fluid for chemistry pH and erosion-corrosion control. It addressed the expected effects on the plant secondary system piping and components when combinations of alternate amines are used. The chemicals considered were Methoxypropylamine (MIPA), Dimethylamine (DMA), and Morpholine.

The evaluation addressed the use of these alternate amines alone and in combination with each other with respect to the balance of secondary side chemical control agents used at Turkey Point, specifically Ethanolamine (ETA), hydrazine, carbohydrazide, and amonium hydroxide. The proposed chemistry regiment is expected to improve pH control throughout the secondary system and reduce iron transport rates, which in turn should reduce sludge deposition and piping corrosion rates in the secondary system.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation demonstrated that no new hazards were created by the alternate amine treatment scheme evaluated and there was no reduction in system piping or component reliability.

Hence, it was concluded that operation of the secondary system with various combinations of amines and conventional chemical control agents would not affect steam generator performance or integrity such that probability of occurrence or consequences of an accident are altered. Since the actions and changes identified in this evaluation did not adversely affect plant safety or require changes to the plant technical specifications, prior NRC approval was not required to implement the actions or changes identified within this evaluation.

47

ATTACHMENT to L-2002-198 Page 48 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SEFJ-01-026 UNIT: 3&4 APPROVAL DATE: 1/10/2002 UFSAR AND DBD CHANGE PACKAGES FOR THE REANALYSIS OF THE LOSS OF LOAD EVENT AND DBD CHANGE PACKAGE FOR THE ROD WITHDRAWAL AT POWER ANALYSIS Summary:

The Loss of Load (LOL) design basis event was reanalyzed to change some analysis inputs. These changes were required to address input inconsistencies in the previous FSAR analysis. The original LOL analysis in the Turkey Point thermal power uprate submittal (performed in 1996 under license amendments 191 and 185 for Units 3 and 4 respectively) did not include uncertainties in the initial pressurizer water level. This error was corrected and the UIFSAR was updated accordingly.

Subsequently, a root cause analysis was performed to assess the impact of the error. As a result of the root cause analysis several modeling inconsistencies in the LOL analysis were identified. When these inconsistencies were corrected, the conservatism in one of the input assumptions had to be relaxed in order to meet the event's acceptance criteria. The affected input is the assumption for the shift in the pressurizer safety valve opening pressure setpoint due to the presence of water in the loop seal. The assumption of +1% for this shift in the previous analysis has been relaxed to +0.8% in the new analysis.

This shift is added to the required +2% tolerance in the opening analysis pressure setpoint. The results of the new analysis with the change in the pressure shift show that all of the applicable safety acceptance criteria, specifically the overpressure criteria, continue to be met. The changes to the LOL analysis are all input changes and are not changes to the methodology used to analyze this event which remains unchanged.

In addition, a UFSAR change package and design basis document (DBD) change package were provided as attachments to this evaluation to document the associated input changes to the LOL analysis. A DBD change package was also provided to clarify the meaning of existing statements about rod withdrawal at power accident analysis.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation documents changes to the analysis input assumptions that are not related to changes in the plant configuration. The reanalysis of the LOL event shows compliance with all of the appropriate safety analysis acceptance criteria. The pressurizer safety valve opening setpoint pressure shift is a parameter that relates to a structure, system or component important to safety. The change to this parameter did not involve a change to the actual performance of the component, but involved a change in the analysis assumption of the component. The new analysis assumption is more consistent with the actual plant configuration than the assumption used before which was generic and intended to envelope the performance of a number of valve designs. The input changes and the corresponding FSAR and DBD change packages have been evaluated and determined not to adversely affect plant safety or require a change in the plant technical specifications. Therefore, NRC approval was not required prior to the implementation of the UFSAR and DBD changes.

48

ATTACHMENT to L-2002-198 Page 49 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SEMS-01-031 Revision 1 UNIT: 3&4 APPROVAL DATE: 4/23/2001 REPAIR OF HYDROGEN EXCESS FLOW CHECK VALVE, REPLACEMENT OF RV-4622 AND REPAIR OF LEAKING LINE TO GAS HOUSE Summary:

This evaluation reviewed the temporary alignment and system arrangement with alternate hydrogen supply to the Volume Control Tanks (VCT) while system repairs were being made to relief valve RV 4633, an excess flow check valve, and a pipe on the hydrogen gas system. The alternative hydrogen supply consisted of a single hydrogen bottle with an integral pressure regulator and excess flow check valve installed in each of the charging pump rooms. The use of excess flow check valves was required to limit flammable gas flow rates into the auxiliary building or other confined spaces in the event of a hydrogen supply line rupture or damage. As such, the excess flow check valve was considered a fire protection feature.

Revision 1 deleted the requirement to have a dedicated operator stationed at the excess flow check valve (following installation) while it is open.

10 CFR 50.59 Evaluation:

The alternate hydrogen supply system was reviewed against UFSAR requirements for reactor coolant system (RCS) hydrogen concentration, VCT pressure control, and fire protection. It was concluded that the alternate hydrogen control/supply system met all UFSAR functional requirements. The evaluation considered the effects of regulator failures and interconnecting pipe failures. It was determined that the excess flow check valves and auxiliary building ventilation system provided adequate level of protection against hydrogen ignition during postulated pressure boundary failures. The evaluation determined that the temporary alignment and system arrangement using a single hydrogen bottle in each charging pump rooms as an alternate hydrogen supply to the VCT did not adversely affect plant safety or require a change to the plant technical specifications. Therefore, prior NRC approval was not required to use the alternate source of temporary hydrogen to the VCTs.

49

ATTACHMENT to L-2002-198 Page 50 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SENS-01-057 Revision 0, Revision 1 UNIT: 3 APPROVAL DATE: 8/28/2001, Rev. 0 APPROVAL DATE: 9/13/2001, Rev. 1 TEMPORARY LOWERING OF UNIT 3 SFP LEVEL Summary:

This evaluation was developed to examine the effects of securing the spent fuel cooling pumps and reducing the pool level by about one foot in order to perform maintenance on valve 3-798B in the filtration return header to the spent fuel pool (SFP). This evaluation addressed the effects of spent fuel handling accidents, spent fuel heatup rate, increased radiation levels resulting from lowered water (shielding) levels, and activation of system alarms. To reduce the potential for fuel handling accidents, all fuel movement and crane operation were suspended in accordance with Technical Specification 3/4.9.11. Considering the decay heat load in the SFP, it was estimated to take about 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> to heat the pool from 108 'F to 130 *F. This duration was determined to be sufficient to perform the required maintenance. Previous evaluations of reduced water levels have demonstrated that expected increases in radiation levels would be negligible. In order to preclude activation of the SFP alarms, pool temperature and level were required to be monitored on an hourly basis. A SFP temperature limit of 130 'F was established as an upper limit during the maintenance activity, at which time work would be secured and SFP cooling restored.

Revision 1 revised the drain down level to adjust for an apparent variance of 2 - 3 inches between the as-constructed elevation of the vacuum breaker hole and the elevation originally used in this evaluation. The drain down elevation change of 2 to 3 inches has no adverse affect on the conclusions reached in this evaluation because there is still approximately 24 feet of borated water above the fuel elements.

10 CFR 50.59 Evaluation:

This evaluation concluded that reducing the spent fuel pool level for maintenance on the isolation valve in the filtration return header would not adversely impact plant operation and would not compromise the spent fuel handling accident analyses, provided that the actions and restrictions identified in the evaluation were observed. Consequently, the reduced pool water level and other actions identified in this 10 CFR 50.59 evaluation did not adversely affect plant safety or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.

50

ATTACHMENT to L-2002-198 Page 51 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SECS-01-059 UNIT: 3 APPROVAL DATE: 8/17/2001 10 CFR 50.59 EVALUATION FOR STORAGE OF TWO NIS DETECTORS IN CONTAINMENT DURING ALL MODES OF OPERATION Summary:

This evaluation addressed the acceptability of temporarily storing two NIS detectors within the Unit 3 containment during all modes of operation. The evaluation considered the potential for adverse seismic interaction with safety related equipment, hydrogen generation, containment free volume and heat sink analysis, containment subcompartment analysis, combustible material, interactions with the containment sump, and the potential for adverse interaction due to high energy line break jet impingment. The evaluation concluded that two NIS detectors could remain temporarily within the unit 3 containment during all modes of operation provided that all of the requirements stipulated were followed. The two detectors were subsequently removed from the Unit 3 containment during the Unit 3 Cycle 19 refueling outage in October 2001.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation concluded that the NIS detectors can safely remain within containment during all modes of operation, provided that all of the restrictions and requirements identified within the evaluation were implemented. The evaluation further concluded that the identified restrictions and requirements would ensure that these activities would have no adverse effects on plant operation, and would not compromise the safety and licensing bases for Unit 3. Consequently, the requirements and restrictions identified in this 10 CFR 50.59 evaluation did not adversely affect plant safety or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the requirements or restrictions identified within this evaluation.

51

ATTACHMENT to L-2002-198 Page 52 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SENS-01-082 UNIT: 3&4 APPROVAL DATE: 11/21/2001 ISOLATION OF THE PRESSURIZER RELIEF TANK BRANCH OF THE REACTOR COOLANT GAS VENT SYSTEM Summary:

This 10 CFR 50.59 evaluation was developed to permit isolation of the pressurizer relief tank (PRT) branch of the reactor coolant gas vent system (RCGVS) in Mode 1. The RCGVS at Turkey Point allows fluid in the reactor coolant system (RCS) to be vented either to the PRT or the containment atmosphere.

Isolation of the PRT branch was desired because one of the system isolation valves was leaking past its seat causing a continuous stream of water to leak into the PRT. This constant leakage of hot RCS fluid into the PRT posed a significant burden to plant operators, necessitating periodic draining and venting of the PRT. The affected portion of the RCGVS was isolated by closing a manual isolation valve inside containment. The evaluation addressed the ability of the modified RCGVS to satisfy its design function, and the effectiveness of the manual isolation valve and upstream piping to maintain RCS pressure conditions.

10 CFR 50.59 Evaluation:

The proposed configuration change was reviewed against all regulatory and design requirements. It was concluded that isolating a manual valve in the PRT branch of the RCGVS would not adversely affect any system design requirements. The modified arrangement was determined to maintain compliance with plant technical specifications, post-TMiI commitments, and UFSAR loading conditions. No new credible hazards were created. The actions and procedure changes identified in this evaluation did not adversely affect plant safety or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.

52

ATTACHMENT to L-2002-198 Page 53 of 66 10 CFR 50.59 EVALUATION PTN-ENG-SEMS-02-001 Revision 0, Revsion 1 UNIT: 3&4 APPROVAL DATE: 3/14/02, Rev. 0 APPROVAL DATE: 3/25/02, Rev. 1 EARLY CORE OFFLOAD Summary:

The plant technical specifications prohibit the movement of irradiated fuel in the reactor core until the reactor has been subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. This evaluation determined that it is feasible from a spent fuel pool heat addition standpoint to begin transferring fuel immediately after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of core subcriticality provided fuel transfer is suspended when pool bulk temperature reaches 1400F.

Suspending fuel transfer activities when the pool bulk temperature reaches 140°F prevents the bulk temperature from exceeding the regulatory commitment of limiting pool temperature to 150oF. The evaluation further determined that restricting the fuel transfer rate to six assemblies per hour or less would provide reasonable assurance that fuel transfer could continue uninterrupted without reaching heatup limits. However, the core average fuel heatup rate of 50.5 Btu/second restricts offload start to no earlier than 116 hours0.00134 days <br />0.0322 hours <br />1.917989e-4 weeks <br />4.4138e-5 months <br /> with no power coastdown and 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> with coastdown if average power is maintained below 50% power for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to shutdown. Based on the new analysis, procedure changes were recommended to limit the minimum offload start time to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> with no coastdown or to 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> if coastdown is performed prior to shutdown.

A UFSAR change package was provided as an attachment to this evaluation to add the analysis results to Appendix 14D.

Revision 1 provided criteria for evaluating other combinations of pre-shutdown operation (i.e., other than 48-hour coastdown) and modifying procedural controls accordingly.

10 CFR 50.59 Evaluation:

The 10 CFR 50.59 evaluation demonstrated that core offload could safely begin no earlier than 116 hours0.00134 days <br />0.0322 hours <br />1.917989e-4 weeks <br />4.4138e-5 months <br /> following subcriticality with no power coastdown or 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> with coastdown if average power is maintained below 50% power for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to shutdown. The fuel transfer restrictions ensures that fuel cladding temperature remains within analyzed limits and do not compromise fuel clad integrity. The earlier offload start times and reduced offload rate do not introduce new failure modes because the equipment and practices employed in fuel handing are unchanged. The evaluation concluded that the change in offload time and offload rate did not affect plant safety or operation or require a change to the plant technical specifications. Therefore, prior NRC approval was not required to implement the revised offload times or offload rates.

53

ATTACHMENT to L-2002-198 Page 54 o f 6 6 SECTION 3 RELOAD SAFETY EVALUATIONS 54

ATTACHMENT to L-2002-198 Page 55 of 66 PLANT CHANGE/MODIFICATION 01-036 UNIT: 3 TURN OVER DATE: 02/27/2002 TURKEY POINT UNIT 3 CYCLE 19 RELOAD DESIGN Summary:

This Engineering Package provided the reload core design for the Turkey Point Unit 3 Cycle 19 reload.

The primary design change to the core for Cycle 19 was the replacement of 56 irradiated assemblies with 56 fresh Debris Resistant Fuel Assemblies (DRFA). Similar to past reloads, these fresh Debris Resistant Fuel Assemblies (DRFA), with a nominal fuel enrichment of either 4.0 w/o or 4.4 w/o, all contain a nominal 6-inch axial blanket of natural U02 pellets at both the top and bottom of the fuel stack. No Wet Annular Burnable Absorbers (WABA) were used in this reload consistent with the current core design practice. The maximum fuel enrichments for the DRFA used in Cycle 19, including the 0.05 w/o fabrication uncertainty, is 4.45 w/o which is no greater than the Technical Specification limit 4.50 w/o.

Mechanical design changes made to the new fuel assemblies include reduction of the axial blanket pellet length (from 0.545 in. to 0.500 in.), change of the top nozzle spring screw from Inconel 600 to bead blasted Inconel 718, top nozzle design change from a welded plug design to an integral plug design, and top nozzle replacement of one fuel assembly. None of these design changes had any adverse impact on fuel performance.

Cross core fuel bundle shuffles were utilized in the Cycle 19 loading pattern to minimize potential power asymmetries. The fuel was arranged in a low leakage pattern with no significant differences between the Cycle 18 and Cycle 19 patterns.

10 CFR 50.59 Evaluation:

The Unit 3 Cycle 19 reload core design was evaluated by FPL and by the fuel supplier, Westinghouse Electric Corporation. The Cycle 19 reload core design met all applicable design criteria, appropriate licensing bases, and the requirements of plant technical specifications. The mechanical design modifications to fuel assemblies in this reload did not affect applicable design criteria and did not increase the radiological consequences of any accident previously evaluated in the UFSAR. These changes had no impact on fuel rod performance, dimensional stability or core operating limits. The Cycle 19 core reload did not have any adverse effect on plant safety or plant operations or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation.

55

ATTACHMENT to L-2002-198 I Page 56 of 66 PLANT CHANGE/MODIFICATION 01-065 UNrr: 4 TURN OVER DATE: 05/22/2002 TURKEY POINT UNIT 4 CYCLE 20 RELOAD DESIGN Summary:

This Engineering Package provided the reload core design for the Turkey Point Unit 4 Cycle 20 reload.

The primary design change to the core for Cycle 20 was the replacement of 56 irradiated assemblies with 56 fresh Debris Resistant Fuel Assemblies (DRFA). Similar to past reloads, these fresh Debris Resistant Fuel Assemblies (DRFA) all contain a nominal 6-inch axial blanket of natural U02 pellets at both the top and bottom of the fuel stack. No Wet Annular Burnable Absorbers (WABA) were used in this reload consistent with the current core design practice. The maximum fuel enrichments for the DRFA used in Cycle 20, including the 0.05 w/o fabrication uncertainty, is 4.45 w/o which is no greater than the Technical Specification limit 4.50 w/o.

Mechanical design changes made to the new fuel assemblies include reduction of the axial blanket pellet length (from 0.545 in. to 0.500 in.), change of the top nozzle spring screw from Inconel 600 to bead blasted Inconel 718, and a top nozzle design change from a welded plug design to an integral plug design and top nozzle replacement of one fuel assembly. None of these design changes had any adverse impact on fuel performance.

Cross core fuel bundle shuffles were utilized in the Cycle 20 loading pattern to minimize potential power asymmetries. The fuel was arranged in a low leakage pattern with no significant differences between the Cycle 19 and Cycle 20 patterns.

10 CFR 50.59 Evaluation:

The Unit 4 Cycle 20 reload core design was evaluated by FPL and by the fuel supplier, Westinghouse Electric Corporation. The Cycle 20 reload core design met all applicable design criteria, appropriate licensing bases, and the requirements of plant technical specifications. The minor design modifications to fuel assemblies in this reload did not affect applicable design criteria and did not increase the radiological consequences of any accident previously evaluated in the UFSAR. These changes had no impact on fuel rod performance, dimensional stability or core operating limits. The Cycle 20 core reload did not have any adverse effect on plant safety or plant operations or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation.

56

ATTACHMENT to L-2002-198 Page 57 of 66 SECTION 4 REPORT OF POWER OPERATED RELIEF VALVE (PORV) ACTUATIONS 57

ATTACHMENT to L-2002-198 Page 5 8 o f 66 ANNUAL REPORT OF SAFETY AND RELIEF VALVE CHALLENGES By letter dated June 18, 1980 (L-80-186) Florida Power and Light stated their intent to comply with the requirements of Item Il.K.3.3 of Enclosure 3 to the Commissioner's letter of May 7, 1980 (Five Additional TMI-2 Related Requirements for Operating Reactors). Pursuant to these requirements, a summary of the power operated relief valve (PORV) actuations that have occurred at Turkey Point during this reporting period is provided below:

Unit 3 No PORV actuations have occurred on Turkey Point Unit 3 between October 24, 2000 and April 7, 2002.

Unit 4 No PORV actuations have occurred on Turkey Point Unit 4 between October 24, 2000 and April 7, 2002.

58

ATTACHMENT to L-2002-198 Page 59 of 66 SECTION 5 STEAM GENERATOR TUBE INSPECTIONS FOR TURKEY POINT 59

ATTACHMENT to L-2002-198 Page 60 of 66 CSI-NDE-01-064 Attachment 1 Page 1 of 1 FORM NIS-BB OWNERS' DATA REPORT FOR EDDY CURRENT EXAMINATION RESULTS As required by the provisions of the ASME CODE RULES EDDY CURRENT EXAMINATION RESULTS PLANT: Turkey Point Unit 3 EXAMINATION DATE: October 9, 2001 through October 13, 2001 TOTAL TOTAL TOTAL TUBES TUBES TOTAL STEAM TUBES TUBES TUBES PREVENTIVELY PLUGGED PLUGGED GENERATOR INSPECTED 20%-39% >40%, PIT & PLUGGED (PTP) THIS TUBES VOL OUTAGE IN S/G 3E210A (Bobbin) 3169 4(11 0 0 0 See RPC 3E210B (Bobbin) 3158 5(11 0 _ _ IM)1 See RPC 3E210C (Bobbin) 3163 17(1) 0 0 0 See RPC 3E210A (RPC) 1739(s) 0 0 1(3) 1 46 9

3E210B (RPC) 1820 (8) 0 1(4) (5)(6) 10 67 3E210C (RPc) 1685s) 0 207 2 53 LOCATION OF INDICATIONS (20% - 100%, PIT & VOL)

Tube Tube Freespan Top of Top of Total Total STEAM AVB Supports Supports 8H thru 6C Tubesheet Tubesheet Indications Indications GENERATOR Bars I thru 6 1 thru 6 UBEND to #1 to #1 20%-39% >40%. PIT C/L H/L Support C/IL Support H/L &VOL 3E210A(Bobbin) 4(11 0 0 0 0 0 4 0 3E210B (Bobbin) 11 (ll' 0 0 0 0 0 11 0 3E210C (Bobbin) 23(1j) 0 0 0 0 0 23 0 3E210 A (RPC) 0 n/a 1 0 n/a 0 1 0 3E210B (RPC) 1 4 5 0 n/a 0 9 1(4) 3E210C (RPc) 0 n/a 2 0 n/a 0 2 0

,, I I Remarks:

(1) Mechanical wear damage at anti-vibration bars (AVB) was depth sized using qualified bobbin coil sizing technique.

(2) One tube was preventively plugged due to mechanical wear (34% through wall) at an anti vibration bar (AV2) In the u-bend.

(3) One tube In 3A was preventatively plugged due to minor (< 40% by RP) wear at the first hot leg broached support plate.

(4) One tube was plugged due to loose part related wear Indication (LPI) at the 3d cold leg support that exceeded the 40% plug limit based on RP plus point techniques.

(5) One tube In 3B was preventatively plugged due to a restriction In the u-bend to a plus point examination. (Row I Column 3)

(6) The remaining.8 4 ear Indications In S/G 3B are located at broached support plates and were depth sized at < 40% by plus point and were preventively plugged.

(7) Two tubes were preventatively plugged due to mechanical wear at the 2nd hot leg support measuring less than 40% by Plus Point technique (8) Includes tubes in the dent, low row U-bend and hot leg TTS expansion transition programs.

60

ATTACHMENT to L-2002-198 Page 61 of 66 PTN-3 S/G "A" OUTAGE: 10/01 13-Oct-01

Description:

20% TO 39% Indications Page I of I Extent Row Col Volts Deg Chan Indn Percent Location Utll Utli2 Cal Probe Tested Dataset Date 28 59 066 P2 22 AV2 PS AC004 A-720-MAILC TEHrEC IS-DEG 10/i 31 44 0.52 P2 20 AV3 PC ACD22 A-720-MIJLC TEHTEC IS-IMP 10/10/01 33 1S 047 P2 20 AV3 AC016 A-720-MWUL TEHTEC IS-DEG 10/10/01 37 47 098 P2 30 AVW AC020 A-720-MAXC TEHTEC IS-DEG IWI0/1 Total Indications: 4 Total Tubes: 4 61

ATTACHMENT to L-2002-198 Page 62 of 66 PTN-3 SIG "B" OUTAGE: 10/01 13-Oct-01

Description:

20% to 39 %Indications. Page I of I Extent Row Col Volts Deg Chan Indn Percent Location Utoll UtflZ Cal Probe Tested Dataset Date 30 42 0.71 P2 24 AV4 -0.24 PS DC029 A20-M/ULC T EC IS-AMP 10/0/01 30 42 0.14 P2 27 AV3 0.13 PS BC029 A-720-MUL TEH IS-IMP 1(0/10/0 30 42 0.58 P2 20 AV2 -0.03 PS BC029 A-720-M/UIC TEHTEC IS-IMP 10/101 32 34 0.6 P2 2i AV4 .008 PS 80029 A-720-M/UILC TITEC IS-DEG 10/10.VI 32 34 109 P2 31 AV3 033 RC BC029 A.?20-MUL. TEHTEC IS-DEO 10/10/01

.32 34 08 P2 26 AV2 -043 PS BC029 A-720-MWAU TEHEC IS-DEG 10/10/01 32 34 0.86 P2 27 AVI PS BC029 A-720-MMUL TEHTEC IS-DEO 10/101 34 51 1.45 P2 34 AV2 BCOI8 A-720-MUC TEHTBC IS-DEG 10/10101 34 53 072 P2 25 AV2 BC019 A-720-MIULC TEHiC IS-DEa 10/10/01 34 53 066 P2 24 AVI BC019 A.'0-MWULC TEHTEC IS-DEG 10/10/01 35 48 0.51 P2 20 AV3 BC019 A-720-M/ULC TEHIEC IS-IMP 10/10/01 Total Indications: 11 Total Tubes: 5 62

ATTACHMENT to L-2002-198 Page 63 of 66 MT-3 SIG licit 13-Oct-01

Description:

20 %to 39 %Indications~ Past' Iof I Row Col! Volts Deg Cbanijndn, Percei~t L'oca'ton utfil -umH cal Prdb~e 'TaJted .' Dataset' Dite 23 45 0OAS P2 22 AV3 Ct= 2--A-720-MXUL TBCITII 1S.D'EG 10/20/01 24 63 0.37 P2 25 AV3 CDO)3 A-720-MJUL TEIITEC IS-DEO 10/10 2S 62 046 P2 23 AV3 .CCD03 A-720-WUL TEHTEC IS-DEC 10/9/01 26 58 0.66 P2 25 AVZ' -0.34 CCO04 A.720.M&L TE-IiEC IS-DEG I10I90 28 48 063 P2 28 AV2 043 PS .CCO14 A 0-MMULC 'IIITEC IS-DiG t0/10~10 30 31 0.46 P2 22 AV3 'CH022 A-720-dIUL TECIE11 SDEG M 101001V 30 31 0.39 P2 20 "1 -0.31 - C12022 A-720-WALC TECTEH 13-DiG 10110/01 30 31 0.3 , P2 23 AV2 CH022 A-70MtUL 2WrBH 13-DiG WIWt/I 30 61 059 P2 26 AV2 -. CCOO3 -A-A7i0-AU=, TW=TE IS.DEQ 10/10 33 '31 0.3 P2 24 AV3 !0.0D3 PS CM14- A-720-MILMC TEHTEC 16-DiG WIWI/0 34 31 0.52 - P2ý 21 AVI CCOIS A-720441UL TwlITC IS-DiO 10(10/0 34 31 0.73 ?2 26 AV3 CC015 A-720-ULC T7fl1IE 1S.DEO 10130f01 34 41 0.71 -P2 - 25 AV2, CCO1S A-720-14/JW TENTE -SDEG 10110/03 34 41 0.71 P2, S 27 AV3 . COD15 A-72D44/UL TEHTEC IS-DEG 10/10/0 34 41 .0.81 . P2 21 AV4 CCO1 A-7204&=UL 'TEIIT IS-DEG 30110/01 34 41 0.71 . P2 26 AVI PS CCCIS A-720-MUAC TEHTEC iS-DBG 20/10/01 34 44 0.41 P2 21 AV3 -0.11 PS CCO14 A-72044ULC TE1IC IS-DEG 1ý0/2001 35 35 046 P2 22 AV3 US0 PS CC3D14 A.72081UL TETEC IS-IMP 10110/01 35 36 0.52 P2 20 AVS CC15 A-720-M/11L TEHTEC 25DiG 10(10/01 3S 49 0.44 P2 P -22 AV4 -0.19 PS CCO12 A.?244UNI TEIITEC IS-DEG 10/10/01 37 29 0.42 P2 -21 AV4 PS COD14 A-720-WULC TEIITE IS-IMP 10/10/01 38 a5 0.49 P2 20 AV4' CCO13 A-M?0.WUL TER=IE IS-DBO 10110/01

-38 71 0.53 ýP2 25 AV3 0.13 PS CC012 A-720-WtIC Ti1EHM IS-DiG 10110/01 Total Indications: 23 Total Tubes: 17 63

ATTACHMENT to L-2002-198 Page 64 of 66 PTN-3 S/G "lA"l OUTAGE: 10/01 Pluggable Indications 10/13/01 11:08:00 AM Page 1 of 1 ROW COL CAL VOLTS DEG CH  % IND UtiI2 SUPPORT INCHES 32 15 AH041 0 0 0 PTP 01H -0.45 TOTAL INDICATIONS: 1 TOTAL TUBES: I 64

ATTACHMENT to L-2002-198 Page 65 of 66 PTN-3 S/G '!B" OUTAGE: 10/01 Pluggable Indications 10/13/01 11:07:15 AM Page 1 of 1 ROW COL CAL VOLTS DEG CH  % IND Utll2 SUPPORT INCHES 1 3 BH046 0 0 0 PTP 0 15 76 BHO41 0 0 0 PTP 031H -0.7 26 77 BH049 0 0 0 PTP 02H -0.48 27 41 BC049 0 0 0 PTm 03C 0.59 27 42 BC049 0 0 0 PTP 03C 0.59 28 41 BC036 0 0 0 PT? 03C 0.69 28 41 BC036 1.22 84 8 0 LYI 03C 0.69 28 41 BC049 0 0 0 PTP 03C 0.61 30 17 BC050 0 0 0 PT? 02C 0.56 32 19 BH021 0 0 0 PTP 02H -0.61 32 66 BHO41 0 0 P1 0 PTP 02H -0.84 34 51 BC018 0 0 0 PTP AV2 0 38 69 BH041 0 0 Pi 0 PT? 02H 0.99 TOTAL INDICATIONS: 13 TOTAL TUBES: 11 65

ATTACHMENT to L-2002-198 Page 6 6 o f 66 PTN-3 S/G "C" OUTAGE: 10101 Pluggable Indications 10113/01 1108:44 AM Page 1 of 1 ROW COL CAL VOLTS DEG CH % IND UtI12 SUPPORT INCHES 19 85 CH027 0 0 0 P7? 02H -0.78 32 64 CH048 0 0 0 FPl 02H -0.59

-TOTAL INDICATIONS: 2 TOTAL TUBES: 2 66