ML20217B987
| ML20217B987 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 04/08/1999 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20217B984 | List: |
| References | |
| NUDOCS 9910130135 | |
| Download: ML20217B987 (93) | |
Text
p. -
1998/1999 10 CFR 50.59
SUMMARY
REPORT FLORIDA POWER & LIGHT COMPANY TURKEY POINT UNITS 3 & 4
$ TOME E ago a
)
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 l OCTOBER 14,1997 THROUGH APRIL 8,1999 ,
l l
1 J
INTRODUCTION This report is submitted in accordance with 10 CFR 50.59(b), which requires that:
i) changes in the facility as described in the SAR
' ii) changes in procedures as described in the SAR, and iii) tests and experiments not described in the SAR, which are conducted with'out prior Commission approval, be _ reported to the Commission for the same period as required by 50.71(e) for the Turkey Point-FSAR update. This report is intended to meet this requirement for the period covering October 14,1997, through April 8,1999.
This report is divided into five (5) sections. The first section summarizes those changes made to the facility as described in the SAR 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 SAR that were performed by a Safety Evaluation. This includes those changes not performed by a PC/M, and any tests and experiments not described in the SAR that were performed during this reporting period. The third section provides a summary of the Unit 3 and Unit 4 fuel reload safety
~
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 a steam generator tube
- inspection during this reporting period.
2 1
l f
TABLEOF CONTENTS' e
I SECTION 1 PLANT CHANGE / MODIFICATIONS PAGE 95-040 ABANDONMENT OF VARIOUS BORON RECYCLE 11 SYSTEM AND LIQUID WATE DISPOSAL SYSTEM COMPONENTS 01/30/98 i
-95,126 DELETION OF INDOOR ELECTRICAL RACEWAY 12 i g FIRE-PROOFING REQUIREMENTS
!- 10/27/98 96-014 THERMO-LAG OVERLAY UPGRADES FOR INDOOR 13 FIRE ZONES.
01/15/99 i 96-059 R11/12 FIRMWARE DEFECT AND HIGH TEMPERATURE 14 PROBLEMS - UNIT 3 12/31/97
~
96-060 R11/12 FIRMWARE DEFECT AND HIGH TEMPERATURE 15 PROBLEMS - UNIT 4 12/19/97 96-068- FUEL TRANSFER TUBE BLIND FLANGE BOLT REDUCTION 16 11/25/98 96-083 INSTALLATION OF GREASE FILLER AND FOAM AT 17 CONDENSER WATERBOX EXPANSION JOINT PIT FOR CORROSION PROTECTION 06/10/98 96-092 UNIT 3 ADDITION OF CCW HEAD TANK 18 -l 10/25/98 i 96-093 UNIT 4 ADDITION OF CCW HEAD TANK 19 03/18/99 4 97-002 REPLACEMENT OF UNIT 3 ICW VALVE 3-50-360 20 11/07/97
.3 J
1 SECTION 1' PLANT CHANGE / MODIFICATIONS (Continued) PAGE 97-003 THERMAL OVERPRESSURIZATION OF ISOLATED 21
. PIPING 10/26/98 97-021 SAFETY INJECTION PIPE VENTING MODIFICATION 22 10/03/98 97-024 FIRE BARRIERS UPGRADES 23 04/09/98 97-031 FIRE PIPING UPGRADES 24 04/06/99 97-033 ELIMINATION OF ELECTRICAL TRIP TO AUXILIARY 25 FEEDWATER TURBINES 03/24/99 97-039 PLANT RELIABILITY IMPROVEMENT MODIFICATION 26 (C-BUS) 10/10/98 98-005 AUXILIARY FEEDWATER CONTROL VALVE INSTRUMENT 27 AIR SUPPLY FILTERS 04/07/99 98-016 MOV-3-744A AND MOV-3-744B REPACKING AND 28 EQUALIZING LINE INSTALLATION 10/23/98 98-028 REPLACEMENT OF UNIT 3 DIESEL OIL TRANSFER 29 PUMP PIPING 10/13/98 98-037 MSIV CONTROL CIRCUlT LOGIC CHANGE 30 1 04/03/99 i 981040 REACTOR COOLANT PUMP 3B MOTOR 31 REFURBISHMENT / UPGRADE 10/21/98 98-049 MOV ENHANCEMENTS - LIMITORQUE TECHNICAL 32 UPDATE 98-01 04/04/99 4
l l
SECTION 1 PLANT CHANGE / MODIFICATIONS (Continued) PAGE
- 99-001 AFW TRAIN 2 STEAM HEADER MODIFICATION 33 04/07/99 SECTION 2 SAFETY EVALUATIONS M-87-136 SAFETY EVALUATION FOR CCW BASKET 35 STRAINER BACKWASHING 11/25/98 SEEJ-88-042 SAFETY EVALUATION FOR DE-ENERGlZATION OF UNIT 36 4 4160 VOLT SAFETY RELATED BUSSES 03/02/99; 03/22/99 SEEJ-89-085 SAFETY EVALUATION FOR DE-ENERGlZATION OF UNIT 37 ;
3 4160 VOLT SAFETY RELATED BUSSES {
07/30/98 '
SENP-95-007 SAFETY EVALUATION FOR OPERABILITY OF RHR 38 DURINMG INTEGRATED SAFEGUARDS TESTING ,
10/02/98 SEMS-95-023 - SAFETY EVALUATION FOR STEAM SAFETY VALVE 39 PERFORMANCE OF MAIN SETPOINT VERIFICATION TEST IN MODE 1 09/18/98;01/27/99 SEMS-96-003 SAFETY EVALUATION FOR UNIT 4 STEAM 40 GENERATORS' SECONDARY SIDE FOREIGN OBJECTS 12/04/97 SENS-96-011 SAFETY EVALUATION FOR USE OF MANUAL 41 ACTIONS TO ISOLATE THE TPCW HEAT EXCHANGERS 01/12/99 SEMS-96-038 SAFETY EVALUATION FOR UNIT 3 STEAM 42 GENERATORS- SECONDARY SIDE FOREIGN OBJECTS 01/27/99
. SENS-96-079 SAFETY EVALU ATION FOR NEl INITIATIVE 43 FOR LICENSING BASIS CONFORMANCE 03/19/98 l
J
SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SEFJ-97-028 SAFETY EVALUATION FOR BEST ESTIMATE LARGE 44 BREAK LOCA (BELOCA) FSAR AND DBD UPDATES FOR 1 TURKEY POINT UNITS 3&4 i 03/19/98 1 i
SECS-97-061 SAFETY EVALUATION FOR CONTAINMENT POLAR 45 ;
CRANE MAINTENANCE INSPECTION PROCEDURE 03/04/99 l
SENS-97-087 SAFETY EVALUATION FOR TRANSFER TO COLD 46 LEG RECIRCULATION EOP CHANGES 11/18/97 SEES-97-094 SAFETY EVALUATION FOR TEMPORARY INSTALLATION 47 i OF REMOTE MONITOR FOR "C" RCP OIL LEVEL I VERIFICATION 12/19/97; 08/18/98 SEMS-97-096 SAFETY EVALUATION FOR FIRE RATED PENETRATION 48 i SEALS 03/19/98; 02/16/99 SEES-98-024 SAFETY EVALUATION FOR EDG LOADING 49 CONSIDERATIONS DUE TO MANUAL LOADING OF ELECTRIC DRIVEN FIRE PUMP 04/02/98- ;
SENS-98-031 SAFETY EVALUATION FOR ON-LINE FILTERING OF 50 THE CCW SYSTEM 06/09/98 SENS-98-033 SAFETY EVALUATION FOR THE DETERMINATION 51 AND ASSESSMENT OF COMBUSTIBLE OILS IN CONTAINMENT BUILDINGS 06/09/98 SEYS-98-036 SAFETY EVALUATION FOR DIAGNOSTIC TESTING OF 52 THE ATMOSPHERIC DUMP VALVE CV *-1607 09/17/98 SEMS-98-039 SAFETY EVALUATION FOR HHSI PUMP 53 ALTERNATE SHUTDOWN FOR CHARGING PUMPS 09/17/98 6
J
l 1
SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SEMS-98-045 SAFETY EVALUATION FOR CONTROL BUILDING 54 BUILT-UP ROOFING COMBUSTIBILITY 10/21/98 SENS-98-047 SAFETY EVALUATION FOR AFW PUMP LOW-FLOW 55 OPERATING RESTRICTIONS 08/06/98 SEES-98-049 SAFETY EVALUATION FOR EDG AND SAFETY RELATED 56 BUS LOADING DUE TO INCREASED POWER RATING OF INSTRUMENT AIR DRYER TOWERS 09/15/98 SENS-98-050 SAFETY EVALUATION FOR OFFSITE DOSE 57 CALCULATION MANUAL REVISIONS RELATED TO THE INTERLABORATORY COMPARISON PROGRAM 02/18/99 SEMS-98-051 SAFETY EVALUATION FOR TEMPORARY LOWERING OF 58 l UNIT 4 SPENT FUEL POOL (SFP) LEVEL TO SUPPORT l MAINTENANCE ACTIVITIES ON THE SFP COOLING DEMINERALIZER RETURN VALVE 4-7988 l 09/15/98 '
SENS-98-053 SAFETY EVALUATION FOR DIFFERENTIAL PRESSURE 59 TESTING OF THE 3B FEEDWATER PUMP MOTOR OPERATED VALVE MOV-3-1421 09/29/98 SENS-98-054 SAFETY EVALUATION FOR DIFFERENTIAL PRESSURE 60 TESTING OF MOTOR OPERATED VALVE MOV-3-856A 10/13/98 SECS-98-058 SAFETY EVALUATION FOR STORAGE OF TOOLS 61 AND EQUIPMENT IN CONTAINMENT DURING ALL MODES OF OPERATION 10/21/98 SENS-98-063 SAFETY EVALUATION FOR FREEZE SEAL FOR 62 REACTOR COOLANT PUMP SEAL INJECTION NOZZLE INLET FLANGES 10/15/98; 03/24/99 7
a
r:
l' SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SEYS-98-069 SAFETY EVALUATION FOR FIRE PROTECTION 63 SURVEILLANCE REDUCTION TASK 04/03/99 SENS-98-071 SAFETY EVALUATION FOR FSAR ACCURACY 64 REVIEW CHANGES FOR CHAPTER 1 02/18/99 SENS-98-072 SAFETY EVALUATION FOR INSTALLATION AND 65 OPERATION OF FEEDWATER ULTRASONIC FLOW MONITORING EQUIPMENT 02/18/99 SEFJ-99-001 SAFETY EVALUATION FOR TO TEMPERATURE / 66 POWER COASTDOWN FOR TURKEY POINT UNIT 4 CYCLE 17 02/25/99 SEFJ-99-002 SAFETY EVALUATION FOR IMPLEMENTATION OF 67 THE SINGLE-POINT INCORE / EXCORE CAllBRATION 04/01/99 SEMS-99-002 SAFETY EVALUATION FOR FSAR ACCURACY 68 REVIEW CHANGES FOR CHAPTER 6 03/12/99 SENS-99-008 SAFETY EVALUATION FOR CONDUCTING RCS 69 FILL AND VENT ACTIVITIES DURING ENGINEERED !
SAFEGUARDS INTEGRATED TESTING 03/29/99 ,
SEMS-99-010 SAFETY EVALUATION FOR RCS CHEMICAL 70 ;
DEGASSING 03/02/99 l
SECS-99-012 SAFETY EVALUATION FOR USE OF ACOUSTIC 71 EMISSION TECHNOLOGY AS AN ALTERNATE METHOD FOR NDE OF SPECIAL LIFTING DEVICES 03/12/99 SEES-99-017 SAFETY EVALUATION FOR TEMPERATURE 72 MONITORING EQUIPMENT IN CONTAINMENT DURING ALL MODES OF OPERATION 03/18/99 8
i SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SENS-99-018 SAFETY EVALUATION FOR USE OF FREEZE SEAL 73 IN SUPPORT OF MAINTENANCE ON RELIEF VALVE, RV-4-791 E 03/12/99 SENS-99-024 SAFETY EVALUATION FOR LEAK INSPECTION OF RHR 74 (PIGGY-BACK) RECIRCULATION FLOW PATHS 03/12/99 i SECS-99-025 SAFETY EVALUATION FOR STORAGE OF TOOLS AND 75 l EQUIPMENT IN CONTAINMENT DURING ALL MODES OF OPERATION 03/31/99;04/06/99 j SECL-99-034 SAFETY EVALUATION FOR FOREIGN MATERIAL IN 76 THE REACTOR VESSEL l 04/03/99 SENS-99-035 SAFETY EVALUATION FOR ALTERNATIVE RHR 77 i FLOW PATH AND OTHER PLANT CONDITIONS TO SUPPORT MOV-4-865B REPAIR l ACTIVITIES 03/20/99 SECTION 3 RELOAD SAFETY EVALUATIONS l
98-014 TURKEY POINT UNIT 3 CYCLE 17 RELOAD DESIGN 79 01/25/99 98-015 TURKEY POINT UNIT 4 CYCLE 18 RELOAD DESIGN 80 07/19/99 l
SECTION 4 REPORT OF POWER OPERATED RELIEF VALVE (PORV) ACTUATIONS {
UNIT 3 82 UNIT 4 82 SECTION 5 STEAM GENERATOR TUBE INSPECTIONS FOR TURKEY POINT UNIT 3 84 - 92 9
J
7-l l
1 I
i SECTION 1 PLANT CHANGE / MODIFICATIONS .
l 1
1 1
1 l l l 1 j
l I
l l~
1 l l
l l
l l
l l 10 t _ )
[ .- ,
PLANT CHANGE / MODIFICATION 95-040 I UNIT : 3&4 TURN OVER DATE : 01/30/98 ABANDONMENT OF VARIOUS BORON RECYCLE
- AND LIQUID WASTE DISPOSAL SYSTEM COMPONENTS Summary:
This Engineering Package provided the design documentation required to abandon the concentrates' holding tank,' waste condensate demineralizer, polishing domineralizer, and base and cation demineralizers that were used in the plant boron recycle system.
These components were installed as part of the original plant design to conserve boric acid usage and minimize liquid radioactive waste processing. The boron recycle system was designed to receive borated radioactive effluent from numerous sources
' including excess letdown during startup, reactor coolant loop drains, pressurizer relief tank,.and clean radioactive drains. The output of the process was separation and reclamation of boric acid and primary water. By the mid-1980s, the ion exchange, gas stripping, and evaporation method of processing liquid wastes had proven too i costly to operate and maintain. Presently, all radioactive liquids requiring cleanup are j
. processed by the waste disposal portable demineralizer skid located in the Radwaste Building,' and released to the circulating water system. Spent resins are transferred
- to shipping containers, dewatered, and sent off-site for final processing and disposal.
All electrical . andL mechanical equipment associated with the above tanks and-
. deminsralizers~ were abandoned in-place. Power feeds were isolated by lifting leads and de-energizing breakers. The various system boundary valves were lock-wired closed. A new section of piping was installed between the-waste holdup tank and the common discharge header of the gas stripper feed pumps to shorten the bypass flow path around the abandoned equipment. A new isolation valve was also installed by this Engineering Package to control the volume of water discharged to the holdup tank.-
Safety Evaluation:
l
' The boron recycle tanks and demineralizers did not serve any safety related functions and were not requiied to support safe shutdown of the plant. The configuration changes imposed by the abandonment of the boron recycle equipment had no j adverse . impact on - plant safety or plant operations. As demonstrated in the !
Engineering Package, the associated modifications did not constitute an unreviewed safety question or require changes to the plant Technical specifications. Therefore, prior NRC approval was not required for implementation.
11
y l
l l
PLANT CHANGE / MODIFICATION 95-126 UNIT : 3&4 TURN OVER DATE : 10/27/98 DELETION OF INDOOR ELECTRICAL RACEWAY FIRE-PROOFING REQUIREMENTS Summary:
This Engineering Package revised the Turkey Point Units 3 and 4 Appendix R Safe Shutdown Analysis (SSA), Appendix R Essential Equipment List (EEL), Appendix R Essential Cable List (ECL), and Raceway Fire Protection Wrap drawings, to modify the fire barrier requirements for certain raceways in 19 indoor fire areas around the plant.
The revisions implemented by this Engineering Package removed the requirement to maintain Thermo-Lag fire barriers on certain raceways providing safe shutdown capability. The documentation changes were justified by imposing additional manual actions and/or demonstrating the availabi!ity of redundant equipment that is protected from the effects of a fire. Changes to the above documents were also made to combine Fire Zone 41 (boric acid tank and pump room) with Fire Zone 55 (Unit 3 charging pump room) in to Fire Area O. The requirement to combine Fire Zones 41 and 55 into a common fire area was predicated on the fact that a single fire could encompass both equipment rooms.
Safety Evaluation:
The document changes implemented by this Engineering Package were enveloped by established fire protection design criteria and regulatory requirements. In those cases where fire barrier requirements were removed by this Engineering Package, compensatory measures were identified or a justification provided which ensured continued availability of the safe shutdown function. The new proceduralized manual actions were evaluated to ensure that adequate time existed to perform them, and that adequate emergency (Appendix R) lighting and access and egress paths existed for successful completion. None of the normal and safe shutdown functions of equipment affected by this modification were altered. Based on the evaluation criteria provided in this Engineering Package, the changes did not constitute an unreviewed safety question, or require changes to the plant technical specifications.
Therefore, prior NRC approval was not required for implementation of these modifications.
12
.. _____ _ _ _ _ _ _ )
PLANT CHANGE / MODIFICATION 96-014 UNIT : 3&4 TURN OVER DATE : 01/15/99 THERMO-LAG OVERLAY UPGRADES FOR INDOOR FIRE ZONES Summary:
This Engineering Package provided.the design change documentation necessary to upgrade the' Thermo-Lag fire barrier material- on those indoor electrical raceways (outside containment) that were' identified in PC/M 95-126 as requiring protection.
The existing' Thermo-Lag material on these raceways has been shown by industry testing to provide a reduced fire rating. The modifications implemented by this EP were limited to the application of. additional Thermo-Lag overlays on existing fire barriers for circuits requiring protection, as well as incidental coverage of nearby components.
Safety Evaluation:
The modifications performed by this Engineering Package restored the required fire rating to those . protected raceways located indoors (outside containment), and resolved the combustibility and degradation issues surrounding Thermo-Lag fire barrier systems. An engineering review demonstrated that the seismic qualification "of the modified conduits and terminal / pull boxes would not be adversely impacted. !
by the additional overlay of Thermo-lag material, due to the minor weight changes !
involvsd. It also demonstrated that the modified raceways would continue to ;
satisfy circuit ampacity requirements. Since the installation of additional Thermo- l Lag material on affected raceway components ~ did not change the operation, function, or design basis of _ any structure, system, or component important to safety, these: modifications did not constitute an unreviewed safety question or
- require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation.
13 s
'\
- PLANT CHANGE / MODIFICATION 96-059 UNIT : 3 TURN OVER DATE : 12/31/97 R11/12 FIRMWARE DEFECT
. ANO HIGH TEMPERATURE PROBLEMS ,
{
l Summary: i l
This Engineering. Package modified: the micro-processor unit of the containment
. gaseous and particulate radiation' monitoring skid to correct an error in the resident
~ firmware.- .The modification consisted of replacing the existing EPROMS which j contained Revision 2 of the manufacturer's firmware with EPROMS that contained j '
Revision 4 of the firmware. The revised firmware corrected software defects that were identified by Sorrento Electronics Inc. in a 10 CFR 21 notification.
' Cooling fans were also added to the micro-processor cabinet and mass flowmeter control cabinet to resolve concerns regarding high temperatures and premature component failures. The cooling fans were designed to maintain the ambient
. temperature within the two enclosures below the values used. in the equipment j
- qualification aging analysis. The fans were powered from an existing 120 V ac power supply-within the skid assembly which obviated the need to install new power cables and conduit. ~ All wiring was routed' internal to the cabinets or within existing i skid wireways. Fuses for the fans were sized to coordinate with the main skid power supply fuse. -it was concluded that emergency diesel generator loading would not be impacted by the addition of the fans due to their low power consumption rating. !
l l
Safety Evaluation:
The revised firmware did~ not alter the function, reliability, or sensitivity of the radiation monitors. The ability of the monitors to initiate a containment ventilation isolation signal or control room ventilation isolation signal remained unchanged with the new firmware. In addition, the changes did not affect any setpoints or operating characteristics of equipment required to prevent or mitigate the consequences of design basis accidents. Since no new failure' modes were created by the revised firmware and ventilation fan additions, the modification did not constitute an unreviewed safety question or require changes to the plant technical specifications.
' Therefore, prior NRC approval was not required for implementation.
14 a
PLANT CHANGE / MODIFICATION 96-060 UNIT : 4 TURN OVER DATE : 12/19/97 R11/12 FIRMWARE DEFECT AND HIGH TEMPERATURE PROBLEMS Sumr my:
This Engineering Package modified the micro-processor unit of the containment gaseous and particulate radiation monitoring skid to correct an error in the resident firmware. The modification consisted of replacing the existing EPROMS which contained Revision 2 of the manufacturer's firmware with EPROMS that contained Revision 4 of the firmware. The revised firmware corrected software defects that were identified by Sorrento Electronics Inc. in a 10 CFR 21 notification.
Cooling fans were also added to the micro-processor cabinet and mass flowmeter control cabinet to resolve concerns re0arding high temperatures and premature component failures. The cooling fans were designed to maintain the ambient temperature within the two enclosures below the values used in the equipment qualification aging analysis. The fans were powered from an existing 120 V ac power supply within the skid assembly which obviated the need to install new power cables and conduit. All wiring was routed internal to the cabinets or within existing skid wireways. Fuses for the fans were sized to coordinate with the main skid power supply fuse. It was concluded that emergency diesel generator loading would not be impacted by the addition of the fans'due to their low power consumption rating.
Safety Evaluation:
The revised firmware did not alter the function, reliability, or sensitivity of the radiation monitors. The ability of the monitors to initiate a NNainment ventilation isolation signal or control room ventilation isolation signal rcauined unchanged with l the new firmware. In addition, the changes did not affect an, setpoints or operating '
characteristics of equipment required to prevent or mitigate the consequences of design basis accidents. Since no new failure modes were created by the revised firmware and ventilation fan additions, the modification did not constitute an i unreviewed safety question or require changes to the plant technical specifications.
Therefore, prior NRC approval was not required for implementation.
1 15 j
i PLANT CHANGE / MODIFICATION 96-068 UNIT : 3 TURN OVER DATE : 11/25/98 FUEL TRANSTER TUBE ByMD FLANGE BOLT REDUCTION Summary:
This Engineering Package modified the fuel transfer tube blind flange bolt pattern to simplify flange assembly and disassembly during refueling outages. The fuel transfer tube is used to transfer reactor fuel between the containment building and the spent fuel pool during the core off-load and re-load windows of an outage. It is sealed closed with two_ concentric gaskets during normal plant operation to satisfy containment isolation requirement. The double gaskets serve as redundant barriers to containment leakage during accidents. The concentric design also permits verification that a proper seal has been established. The existing flange and gasket design utilized 20 bolts to achieve the required gasket sealing. This Engineering Package changed the gasket design such that proper gasket loading could be achieved with a minimum of 8 bolts. The bolt size, material and preload values remained unchanged.
This Engineering Package also provided the design documentation necessary to I permit the local leak rate test connection to be routed from the existing location ;
adjacent to the flange to a new location that would be accessible from the operating i floor of the containment building.
Safety Evaluation:
The modifications addressed by this Engineering Package maintained the existing level of protection against the release of radioactivity to the environment. No changes were made to the containment isolation design bases and no new failure l modes were introduced by the new flange bolt and gasket design. Additionally, no changes were made that would increase the consequences of an accident previously
> evaluated in the FSAR. Since no functional or performance characteristics were ,
altered by the design changes, the modifications approved by this Engineering l Package did not involve an unreviewed safety question or require changes to plant technical specifications. Consequently, prior NRC approval was not required for <
implementation I
16 i
p PLANT CHANGE / MODIFICATION 96-083 UNIT : 3&4 TURN OVER DATE : 6/10/98 INSTALLATION OF GREASE FILLER AND FOAM AT CONDENSER WATERBOX
~ EXPANSION JOINT PIT FOR CORROSION PROTECTION Summary; This Engineering Package added water repellent grease, polyurethane foam, and sta!ntess steel sheet metal lagging around the main condenser water box outlet expansion joints to protect the metal flanges and flange bolts from corrosion. These metallic components are located in an outdoor condenser pit and tend to corrode over time because standing water _ collects in the joint area' between the inlet and outlet flanges of the expansion joint. To remedy this situation, a film of grease was applied to the metallic parts of the. flanges and flange bolting to act as a protective barrier against moisture intrusion. PolyuretNne foam was added in between the expansion joint flanges to protect the grease coating and prevent water accumulation in the gap space. Stainless steel lagging was installed over the expansion joint to protect the foam from damage during future maintenance activities around the condenser waterboxes. Since the grease and polyurethane foam were manufactured from j combustible materials, the lagging also maintains a physical barrier between the foam material and any transient combustibles or ignition sources that may be introduced in !
the immediate vicinity of the expansion joints. The addition of these components did i not hamper the movement of the expansion joint or impact operation of the l circulating water system. l I
Safety Evaluation:
The main plant condenser and circulating water syst m do not perform any safety related functions and are not subject to limiting conditions for operation in the plant technical specificatiens. The increase in combustible loading caused by the grease i and polyurethane material was evaluated and found not to create a significant impact on the plant fire hazard analysis. It was further demonstrated that those areas adjacent to the condenser pit which contain significant combustibles were :
equipped with appropriate fire protection features such that the introduction of flammable liquids within the condenser pit is not considered to be credible. Since the expansion joint modifications did not change the operation, function or design bases of any structure, system or component important to safety, the changes implemented by this Engineering Package did not involve an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation.
17 J
1 PLANT CHANGE / MODIFICATION 96-092 l UNIT : 3 TURN OVER DATE : 10/25/98 UNIT 3 ADDITION OF CCW HEAD TANK Summary:
This Engineering Package installed a static head tank above the existing component cooling water (CCW) system surge tank to increase the operating pressure of the system, and resolve potential voiding concerns identified in Generic Letter 96-06.
Functions such as insurge capacity, normal operating band, system vent, etc. were transferred from the surge tank to the new head tank. The head tank was sized smaller than the existing surge tank but the existing water volume and operating range were retained in the modified design. A variety of miscellaneous changes were implemented to accommodate both the increase in normal system operating pressure (static head increase) and the reduced insurge capacity. These included instrument changes, relief valve setpoint changes, logic change for RCV-3-609, increase in the excess letdown heat exchanger design rating, and replacement of the MOV 716A&B spring packs to accommodate the required increase in closing torque. A self-regulated pressure control valve (PCV-3-832) was also installed in the normal fill line to the CCW system to provide a controlled makeup rate to the CCW head tank.
No physical modifications were necessary to adapt the existing CCW surge tank to water solid operation. A low level alarm, however, was added to the surge tank to l assist Operations personnel with level control during periods of reduced CCW j inventory.
1 Safety Evaluation:
The modifications addressed by this Engineering Package eliminated a potential failure mode for the CCW system and restored its capability to operate as a subcooled system in accordance with the original design intent. The increase in CCW system normal and transient operating pressures were analyzed and found to be within applicable design code requirements. In addition, all of the components installed by this Engineering Package were designed to accommodate the applicable FSAR seismic and hurricane wind loads. The logic change for RCV-3-609 did not adversely affect any system safety function. Modifications performed on the exces; letdown heat exchanger and system relief valves ensured that system inventory and unction would not be adversely affected by anticipated overpiessure events. Based on the evaluation criteria contained in the Engineering Package, these modifications did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of f the CCW system changes.
l 18
u PLANT CHANGE / MODIFICATION 96-093 UNIT : 4 TURN OVER DATE : 03/18/99 UNIT 4 ADDITION OF CCW HEAD TANK Summary:
.This Engineering Package installed a static head tank above the existing component cooling. water (CCW) system surge tank to increase the cperating pressure of the L system, and resolve potential voiding concerns identified in Generic Letter 96-06.
Functions such as insurge capacity, normal operating band, system vent, etc. were transferred from the surge tank tc, the new head tank. The head tank was sized smaller than the existing surge tank but the existing water volume and operating o range were retained in the modified design. A variety of miscellaneous changes were implemented to accommodate both the increase in normal system operating pressure
. (static head increase) and the reduced insurge capacity. These included instrument changes, relief valve setpoint changes, logic change for RCV-4-609, increase in the excess letdown heat exchanger design rating, and replacement of the MOV .716A&B spring packs to accommodate the required increase in closing torque.
This revision installed a self-regulating pressure control valve (PCV-4-832):in the normal CCW system. fill line to control the makeup rate to the new head tank. This change is intended to reduce the potential for overfilling the tank, o'r actuating the system relief valves, if a failure of the motor-operated fill isolation valve occurs. .
'\
Safety Evaluation:
-l The modifications addressed by this revision of the Engineering Package enhanced I the ability to add makeup water to the CCW system during normal fill activities. The i
change- did. not' impact the operational or performance characteristics of any j component in the normal fill or alternate fill flowpaths. Additionally, the change dir' I not create'any new failure modes or adversely affect any system safety functions. !
The modified design continues'to meet all applicable FSAR loading conditions. Based i on the. evaluation contained 1in the Engineering Package,. design change did not
~
l constitute an unreviewed safety question or require changes to the plant technical :
specifications. ;Therefore, prior NRC approval was not required for implementation. j
~ 19
t PLANT CHANGE / MODIFICATION 97-002 I
l UNIT : 3 TURN OVER DATE : 11/07/97 i
REPLACEMENT OF UNIT 3 ICW VALVE 3 50-360 Summary: ;
The Engineering Package modified the configuration of the intake cooling water (ICW) system piping downstream of the component cooling water (CCW) heat exchangers to circumvent the use of valve 3-50-360 for header isolation. Valve 3-50-360 has i historically required an excessive amount of torqua to operate its handwheel and could not be repaired without isolating all ICW flow to the CCW heat exchangers.
Manual operation of valve 3-50-360 is periodically required to provide a downstream isolation function for the 3C CCW heat exchanger during heat exchanger cleaning operations, it is also used to throttle ICW flow through the heat exchanger during quarterly ICW inservice tests.
To avoid an ICW system outage, this Engineering Package installed an additional isolation valve upstream of valve 3-50-360 to provide the heat exchanger isolation function. The new isolation valve (3-50-382) was installed at the location previously occupied by the abandoned continuous tube cleaning strainer. Installing a new isolation valve at this location alleviated the 3C CCW heat exchanger isolation problem without having to shutdown the entire ICW system. Minor support modifications were required to accommodate the dimensional dJferences between
.the new valve and the abandoned strainer assembly. Valve 3-50-360 was set to the ,
full open position and retained as passive part of the ICW pressure boundary. The i valve 3 50-360 handwheel was removed to prevent inadvertent isolation of ICW ;
flow. Flow control during inservice testing was assigned to valve 3-50-351 located l upstream of the 3C CCW heat exchanger. I 1
Safety Evaluation:
The new isolation valve (and accompanying spool piece) added by this Engineering <
Package were seismically installed and did not adversely affect the operation, '
function, or design bases of.the ICW system. Additionally, no new failure modes were created by the passive piping and valve changes. Since the response of the ICW system during design basis accidents remained unchanged with the new piping design, the modification did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was
' not required for implementation.
20 J
r PLANT CHANGE / MODIFICATION 97-003 UNIT : 3 TURN OVER DATE : 10/26/98 THERMAL OVERPRESSURIZATION OF ISOLATED PIPING Summary:
This ' Engineering Package modified several sections of water-filled piping inside containment to eliminate the potential for isolated segments to become thermally overpressurized from heating during accident conditions. This potential failure mode was identified in Generic Letter 96-06 as a condition that could jeopardize the ability of systems to perform their safety related functions, and putentially lead to a loss of containment integrity.
Those piping segments potentially susceptible to thermally induced overpressurization were modified by either a) installing a thermal relief valve within the isolated boundary, b) modifying the failure position of an affected boundary valve from fail-close to fail-open, or c) drillirig a 1/8" diameter hole in an affected boundary valve disc.
The new thermal relief valves were generally set to relieve at pressures 10% higher than the lowest rated component within the isolated bounds. The relief valves were installed on the inside contoinment portion of affected piping associated with containment penetrations. The_ discharge of each relief valve was piped to a collection tank ins.de containment.
Safety Evaluation:
The modifications addressed by this Engineering Package eliminated a potential failure mode for various water-filled piping systems inside containment. The provision of thermal relief paths for the affected piping segments did not alter any of the critical functional characteristics of the piping system. The affected piping segments were evaluated in accordance with seismic criteria contained in the FSAR to ensure that the installation did not create the possibility of any adverse interactions with safety related structures, systems and components. Those modifications made to system boundary valves maintained the existing containment penetration design bases and did not introduce any new failure modes not previously analyzed. The FSAR commitment that containment isolation be established assuming an independent single active failure remains intact with the modified design. Accordingly, the implemented changes did not involve an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation.
21 2
q l F
1
- PLANT CHANGE / MODIFICATION 97-021 UNIT : 3 TURN OVER DATE : 10/03/98 SAFETY INJECTION PIPE VENTING MODIFICATION
- Summary:
This: Engineering: Package was. developed to eliminate a personnel safety hazard associated with venting the Unit 3 high head safety injection (HHSI) piping. The venting procedure historically required that a %-inch blind flange be removed at valve 940V and a venting flange with a hose attachment be installed in its place to vent the system. The exchange had to be performed on a ladder since valve 3-940V is located approximately 10 feet off .the floor. Due to- the weight of the flange
- assemblies, the work posed a safety hazard to personnel on the ground - especially when a flange was passed between the operator on the ladder and the operator on the ground. To alleviate the safety concern, the vent piping was extended from valve 3-940V to a point - approximately 4 feet off the- . floor. The extension was accomplished using' %-inch welded stainless steel tubing and a new terminal vent
- isolation valve.
Safety Evaluation:
The changes implemented: by 'this Engineering Package enhanced the ability to j periodically vent the HHSI system piping. The vent path change did not adversely i affect'the operation,. function, or' design bases of the HHSI system. Additionally, ;
-.no'new failure modes were created by the passive piping and valve changes. . The l Engineering Package evaluated the new configuration and determined :that the '
modified vent piping satisfied all of applicable Updated FSAR loading conditions, i including seismic loads. Since the response of the HHSI system during design basis accidents remained unchanged with:the new vent piping design, the modification j did not constitute an unreviewed safety question or require changes to the plant '
technical ' specifications. Therefore, prior NRC approval was' not required for implementation.
~ '
.r 22-L s
T PLANT CHANGE / MODIFICATION 97-024 UNIT : 3&4
~
TURN OVER DATE : 04/09/98 FIRE BARRIERS UPGRADES Summary:
The' qualification L of Thermo-Lag fire barrier systems is an industry issue. In-addressing this issue, a major objective of the Turkey Point Thermo-Lag project is to reduce the applications of Thermo-Lag in the Fire Protection Program. ~ This is accomplished by relying on other methods of fire protection and only providing new or ' upgraded Thermo-Lag installations where alternatives are not practical.
Accordingly, this Engineering Package provided the necessary design documentation to: a) plug ventilation openings in the west wall of.the Electrical Equipment Room to
-c ' eliminate the need to protect redundant safe shutdown raceways on either side of the wall, b) plug niches in the Computer Room and MCC 3B and 4B Rooms to eliminate reliance on Thermo-Lag covered steel plates as fire barriers, c) install curbs to limit the size of an oil pool fire postulated to result from a turbine lube oil spill, d) replace the steel floor grating in Unit 4 condenser and condensate pump pits with checkered plate steel, e) add additional area drains at grade level beneath the electrical generators, and d)' eliminate . reliance on Thermo-Lag ~ for reach-rod assemblies.
Although replacement of grating with checkered plate in the Unit 4 Condensate Pump .
n Pit does not reduce Thermo-Lag reliance per se, it enhanced the wet pipe sprinkler i system response time and was included in this Engineering Package as an additional '
structural modification.
i Safety Evaluation: )
The modifications' addressed by -this Engineering Package reduced the scope of !
- Thermo-Lag fire barrier material application and maintenance, and brought the plant into compliance with the existing fire hazards analysis. The changes did not alter the operation, function, or design basis of. any structure, system, or component
. considered important to safety. Additionally, the various structural modifications did !
not introduce any new functional or spatial interactions with equipment considered ;
important to safety. ' Since no new equipment or Operator actions were invoked by !
the changes, the modifications' did not constitute an unreviewed safety question, or require a change to the' plant technical specifications.
1 i ,
23 a
PLANT CHANGE / MODIFICATION 97-031 UNIT : 3&4 TURN OVER DATE : 04/06/99 FIRE PIPING UPGRADES Summary:
This Engineering Package modified the hydraulic configuration of the main fire loop to accommodate the increase in demand created by an expanded turbine building sprinkler system design. The modifications included adding a cross-tie between the east and west fire mains and separating the auxiliary transformer and hydrogen seal oil system fixed water spray systems from the main transformer fixed water spray system. The cross-tie was added to increase the available water supply flow and pressure at the fire main points of service. Sepa ating the main transformer fixed water spray system supply from the supplies to the auxiliary transformer and hydrogen seal oil units was intended to reduce the fixed spray demand. The combination of these two changes provided a substantial increase in the water pressure and flow rate available at the fire main to support the increase in demand.
The demineralized water storage tank hose connection station was also modified to increase the flow delivery capacity of the fire pump.
The primary purpose of these modifications was to support a reduction in the application and maintenance of Thermo-Lag fire barrier material, and to provide an adequate source of water to the planned turbine building sprinkler system to accommodate a postulated turbine lube oil fire that results from a gross failure at the low-pressure turbine and/or generator bearings. The postulated fire scenario and commitments for protection coverage are described in the exemption request submitted via letter L-97-181.
Safety Evaluation: i l
The. modifications addressed by this Engineering Package did not impact the l operation, function, or design basis of any safety related equipment. Separating the fixed water spray systems and adding a fire main cross-tie enhanced the ability to deliver the required water flow and pressure to the turbine building service points. No changes were made to any of the spray system patterns or existing !
degree of component coverage. Additionally, the fire piping upgrades did not )
increase the probability of an internal flooding event. Since no new equipment or Operator actions were invoked by the changes, the modifications did not constitute an 'unreviewed safety question, or require a change to the plant technical specifications.
24
PLANT CHANGE / MODIFICATION 97-033 UNIT : 3&4 TURN OVER DATE .: 03/24/99 ELIMINATION OF ELECTRICAL TRIP TO AUXILIARY FEEDWATER TURBINES Summary:
This Engineering Package disabled the electronic overspeed protection feature installed on the turbine-driven auxiliary feedwater (AFW) pumps to improve system performance. The turbine electronic overspeed protection feature was added in the early 1980's during replacement of the turbine drivers. The intent of this feature was to increase the reliability of the steam driven pumps by providing an overspeed protection circuit which could terminate minor turbine overspeed transients and automatically reset and restart the turbine controls to resume AFW flow delivery.
Although this capability was provided to enhance the operation of the AFW pump, improved reliability was not realized. Several spurious trips of the "A" pump were attributed to the electronic overspeed protection circuitry. Testing of the electronic overspeed device indicated that the circuitry was susceptible to producing spurious trips when exposed to fluctuations in the de power supply from inductive noise.
Disabling the electronic overspeed trip eliminated this failure mechanism.
Safety Evaluation:
A review of the FSAR safety analyses and anticipated transients without scram events demonstrated- that no credit is taken for the auto-restart feature of the electronic overspeed protection circuitry. In all cases, the AFW is monitored at some constant flow rate with controlled changes in flow rates at specified time intervals.
Any deterioration in the flow delivered by a pump from that assumed in the analyses is considered a failure and modeled accordingly, No credit is taken for that pump or any intermittent flow that may be delivered. Sufficient redundancy of pumps, valves, i
and pipq systems exist to ensure that the AFW system can accomplish its safety functions under any single failure condition, including the failure of a pump on mechanical or electronic overspeed. Since the modification did not alter the function, design basis, or performance characteristics of the AFW system, the modification did not have any adverse effects on plant safety or operation. Consequently, disabling the electronic overspeed circuitry did not involve an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation.
25
3 > ,
PLANT CHANGE / MODIFICATION 97-039 UNIT : 3 TURN OVER DATE : 10/10/98 PLANT RELIABILITY IMPROVEMENT -
MODIFICATION (C-BUS)
Summaryi This. Engineering' Package provided the necessary design documentation to repower several non-safety related loads currently supplied by a non-vital source from a vital {
. power supply. The power _ supply change.was made to improve plant reliability and J availability, and improve the plant-Operators ability to respond to a unit trip resulting from a loss of the C 4160 V bus. The loads that were repowered included the 3C moisture separator reheater -(MSR) steam stop valve MOV-3-1433, the 3D MSR '
steam -stop valve MOV-3-1434, the component cooling water (CCW) surge' tank make-up valve MOV-3-832, and the volume control tank (VCT) outlet valve LCV <
115C. !
Each load was switched from' a non-vital breaker to a new vital breaker within'the existing MCC. Due to space limitations, however. the starter cubicles for the two MSR stop valve loads and the VCT outlet valve load could not be physically relocated to the vital MCC panel. Instead, an existing spare breaker bucket _in the vital MCC panel was modified to accommodate the additional loads. The design utilized a single vital breaker to power three sets of three slugs in the non-vital MCC which in turn fed the individual MOV loads. The slugs were added in series with the downstream
- circuit breakers to provide a means for_ disconnecting each individual circuit. The new power and control cables were routed internal to the MCCs within existing wireways.
~
Safety 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. Consequently, the modifications did not involve an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation.
26
V PLANT CHANGE / MODIFICATION 98-005 UNIT : 3&4 TURN OVER DATE : 04/07/99 AUXILIARY FEEDWATER CONTROL VALVE INSTRUMENT AIR SUPPLY FILTERS Summary:
This engineering package provided the necessary design documentation to modify the instrument air (IA) supply piping to the auxiliary feedwater (AFW) flow control valves to improve valve reliability and availability. The modifications included reconfiguring the lA supply filters from a series arrangement to a parallel arrangement, replacing the galvanized IA piping upstream and downstream of the filters with stainless steel tubing, and installing redundant check valves in the IA piping at the point where it interfaces with the emergency nitrogen backup supply piping. The modifications were primarily intended to eliminate debris intrusion as a potential failure mode for the code boundary check valves installed between the quality related lA system and the safety related AFW nitrogen backup system.
The reconfiguration of the lA supply filters was intended to permit periodic filter replacement without disrupting filtered IA to the AFW flow control valves. The reconfiguration necessitated the installation of new isolation valves, tubing, and pipe supports. In an effort to reduce the available sources of IA debris, the existing galvanized piping immediately upstream of the filters and all of the piping downstream of the filters was replaced with stainless steel tubing. The addition of ;
redundant check valves in series with the existing code boundary check valves was l intended to increase the isolation reliability of the code pressure boundary. The '
design of the new check valves was selected to be different than the existing check valves to eliminate the potential for common mode failures. ;
Safety Evaluation: 1 The changes implemented by this Engineering Package enhanced the ability of the AFW nitrogen backup system to perform its safety function during accident conditions. The changes did not adversely affect the function, operation, or design bases of the lA system or the AFW nitrogen backup system. Operation of the flow control valves remained unchanged by the piping modifications. A calculation performed in support of the changes demonstrated that the modified piping design continues to satisfy all FSAR loading mnditions. Based on the evaluation criteria provided in this Engineering Package, the changes did not constitute an unreviewed safety question, or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation.
27
PLANT CHANGE / MODIFICATION 98-016 UNIT : 3 TURN OVER DATE : 10/23/98 MOV-3-744A AND MOV-3-744B REPACKING AND EQUAllZING LINE INSTALLATION Summary: l This Engineering Package installed bonnet equalizing lines on the residual heat i removal (RHR) system discharge isolation valves, MOV-3-744A and MOV-3-744B.
These equalizing lines allow any potential fluid trapped in the bonnet cavities to be vented back to the reactor coolant system (RCS) during depressurization events, and eliminates the potential for high pressure fluid remaining in the bonnets to hydraulically lock the valves in the closed position. i installation of the equalizing lines required that the existing packing leak-off lines be '
cut, capped, and abandoned in place. The lower set of packing rings also had to be removed frorn the valve stuffing boxes to establish the necessary vent path between the bonnet and the packing leakoff port. The equalizing lines converge downstream ,
of the valves and form a common vent path to the RCS. Connection to the RCS is at !
an existing vent valve location. The. installation required approximately 10 feet of i 3/8" stainless steel tubing, associated tube clamps, and supports. I A check valve was installed in each equalizing branch line and a single manual i isolation valve was installed in the common tubing run. The manual isolation valve is i required to be locked in the open position during plant operation. l Safety Evaluation:
The modifications addressed by this Engineering Package eliminated a potential failure l mode for the RHR system. No new failure modes were created by the modified stuffing box arrangement or the equalizing line installation. The new tubing and i
. supports were evaluated in accordance with seismic criteria contained in the FSAR to I ensure that the. installation did not create the possibility of any adverse interactions with safety related structures, systems and components. The components installed by this Engineering Package (i.e.,10 feet of stainless steel tubing, associated tube clamps, supports, and valves) were found to have a negligible impact on the existing emergency core cooling system (ECCS) heat sink analysis, and hydrogen generation i analysis. Based on the evaluation criteria contained in the Engineering Package, the '
modifications did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation.
28
V u
' PLANT CHANGE / MODIFICATION 98-028
- G UNIT : 3 TURN OVER DATE : '10/13/98 REPLACEMENT OF UNIT 3 DIESEL OIL
- TRANSFER PUMP PIPING Summary:
~
This Engineering Package replaced the underground. diesel fuel oil piping on Unit 3 with above ground piping due. to the- discovery of a through wall leak near the location where one of the piping segments penetrated the soil. The scope of the
. replacement included the underground piping between the diesel oil storage tank and the _ suction of the diesel' oil transfer pumps. It also-included the underground segments located at the discharge of the pumps to the diesel oil day tanks. Routing .
L the new piping. above ground was intended to reduce the potential for similar corrosion failures and facilitate _ future pipe inspections. Pipe sleeves were installed in the earthen dike surrounding the diesel oil storage tank to preclude subjecting the new piping to subterranean conditions. The modified pipe runs' utilized equivalent pipe lengths and fittings such that there.was no appreciable change in the flow characteristics of the pipes. The existing underground piping was drained, flushed,
- capped, and abandoned in place.
In addition to the pipe replacement activities, new isolation valves were installed in the branch supply lines .from the diesel oil storage tank to permit independent isolation of downstream system pumps and components for convenient maintenance and servic'eability.
- Safety Evaluation:
I The modifications performed by this Engineering Package did not change the function, operation, or design basis of the diesel fuel oil transfer system. Equivalent hydraulic. performance _. characteristics were retained with the modified piping arrangement. Additionally, the use of above ground piping (instead of below ground piping)'did not alter the design of the facility with regard to missile protection. The ability to met the applicable Updated FSAR seismic and wind loading conditions was ;
demonstrated by calculation. Based _on the evaluation criteria contained in the '
' Engineering- Package, the modifications did ; not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior ;
NRC approval was not required for implementation.
29 4
s ,
j!
e ,
PLANT CHANGE / MODIFICATION 98-037 UNIT : 4 TURN OVER DATE : 04/03/99
'MSIV CONTROL CIRCUlT LOGIC CHANGE Summary: -
This Engineering Package modified the Unit 4 main steam isolation valves (MSIVs) to prevent inadvertent closure of the valves due to a failure of an auxiliary control relay
'in'the actuation logic. An auxiliary control relay is provided in each MSIV circuit to seal-in a manual' or automatic' MSIV closure signal. It is designed to maintain the MSIV in a closed position until the operator resets the seal-in ' condition by taking the respective NISIV control switch to the open position. The auxiliary relay was originally designed to be normally energized and to de-energize to provide the seal-in function.'.With this arrangement, any failure of the relay would cause the associated MSIV to close. Inadvertent closure of an MSIV at power will result in a reactor trip condition. This Engineering Package reversed the auxiliary control relay logic so that it will be de-energized during normal plant operation and energized when required to accomplish its intended function.
Failure of a MSIV auxiliary control relay has previously occurred on Unit 3 causing inadvertent closure of the affected MSIV and a Unit 3 reactor trip event. This design -
change was implemented to prevent a similar trip condition from occurring on Unit 4.
1 Safety Evaluation: ~
The modification performed by this Engineering Package was evaluated to ensure that .
no new failure modes were created. It was concluded that the circuit changes did i not alter the method of isolating the MSIVs during an accident, or retard the valve i
. closing speed. The . evaluation demonstrated that sufficient redundancy and j electrical separation were retained in the modified design to ensure that the MSIVs ;
will close as required under single active failure conditions. Since the design basis ;
for main steam isolation.-was not affected by the relay logic change, the circuit '
modifications did not constitute an unreviewed safety question or require changes Lto the plant technical specifications. Therefore, prior NRC approval was not
. required for implementation.
30
PLANT CHANGE / MODIFICATION 98-040
' UNIT : 3 TURN OVER DATE : 10/21/98 REACTOR COOLANT PUMP 3B MOTOR REFURBISHMENT / UPGRADE Summary:
This engineering package provided for the refurbishment and upgrade of the 3B 1 reactor coolant pump (RCP)~ motor. The design bases established in the Updated l FSAR were reviewed and determined to be unaffected because the modifications !
met all FSAR criteria stipulated for the original design. In addition, the modifications did not impact any plant technical specifications. The original installed motor was replaced with a spare motor which was refurbished at the Westinghouse Electro-Mechanical Division facility. This refurbishment consisted of inspection 'and maintenance activities performed to the existing design specifications. In addition, a multiport drain sump and labyrinth entry vent port were installed concurrent with the refurbishment to ensure consistency with the latest RCP technology, and to realize additional reliability and availability. The intent of this modification is to essentially eliminate the anomalous oil level alarms caused by dynamic fluid effects and improve oil pressure sensing characteristics of the RCP motor. A new setpoint was also established for the lower oil reservoir high level alarm.
Safety Evaluation:
The only safety related function performed by the RCP motors is to maintain a i sufficient amount of inertir (through its flywheel) to satisfy the coastdown flow requirements assumed in the plant safety analyses for protection against departure from. nucleate boiling. The design bases established in the Updated FSAR were reviewed and determined not to be affected because the modifications met all Updated FSAR criteria stgulated for the original design. In addition, the modifications did not impact the hydraulic performance of the pump or the coastdown capability of the pump motor. Since the modifications performed by this Engineering Package did not have any adverse effect on plant safety or plant operations, the modifications did not constitute an unreviewed safety question or require changes to the plant technical specifications. Accordingly, prior NRC approval was not required for implementation.
31
PLANT CHANGE / MODIFICATION 98-049 j
{
UNIT : 4 TURN OVER DATE : 04/04/99 l MOV ENHANCEMENTS LIMITORQUE TECHNICAL UPDATE 98-01 l
Summary:
This Engineering Package modified several safety related motor-operated valves (MOVs) to address a change in the application factor published in Limitorque Technical Update 98-01. The application factor is part of the equation used to size a MOV actuator. It is a dimensionless number specified by the actuator vendor which accounts for motor rnanufacturing variations not accounted for by the motor manufacturer. Technical Update 98-01 reduced the application factor from 1.0 to 3 0.9. Changing the application factor can affect the calculated available output torque of an actuator. To ensure that the safety related MOVs will be capable of ;
accomplishing their safety functions with a reduced output torque, this Engineering '
' Package changed the gear ratio on the actuators of valves MOV-4-863A & B, MOV - !
4-864A & B, MOV-4-872, and MOV-4-843A & B. It also replaced the motors on the actuators of valves MOV-4-1417 and MOV-4-1418 with larger capacity motors. The motor overloads and breakers were also upgraded accordingly.
Safety Evaluation:
i The modifications addressed by this Engineering Package increased the available output torque of several safety related MOVs to ensure that they will be capable of accomplishing their safety function under reduced voltage conditions (given the change in application factor). The resulting changes in valve stroke time were evaluated to ensure that existing safety analysis assumptions remained valid. In each case, it was demonstrated that the affected safety system would continue to function within analyzed bounds. Additionally, an engineering review demonstrated that the seismic qualification of the affected valves was not adversely affected by the modifications due to the small weight changes involved. Replacement gears were shown to be consistent with the original gear material, and compatible with the actuator materials of construction. Since the implemented changes did not alter any valve functions, or methods of valve actuation, the modifications implemented by this Engineering Package did not have any adverse effects on plant safety or operation. Consequently, the modifications did not involve an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation.
32
w i PLANT CHANGE / MODIFICATION 99-001 h
e UNIT : 3&4 TURN OVER DATE : 04/07/99
- AFW TRAIN 2 STEAM HEADER MODIFICATION Summary:
t This Engineering Package replaced a section of corroded drain piping connected to the auxiliary; feedwater (AFW) train 2 steam header. A ' through-wall leak had-previously developed in this section of piping during the Unit 3 reactor trip that occurred on 2/16/98. THs Engineering Package was written to permit replacement of the affected piping seguent with Units 3 and 4 on line. In order for the piping replacement activity to be completed successfully and in a safe manner during the AFW allowed outage time, a new 4-inch gate valve was installed upstream of the corroded piping segment to provide an isolation barrier for the steam header. Closing
..this isolation valve during the pipe replacement activity would restore the steam supply from' the 3A and 3B steam generators to the "B" and "C" AFW pump turbines, allowing the pipe replacement activity to be performed outside of the 72- ,
hour ~AFW allowed outage time. It would also allow future pipe repair activities to be !
performed on the train 2 steam piping with out impacting the AFW system limiting i conditions' for operation. I The new isolation valve is designed to the same design code as the existing header piping and is suitable for saturated steam service. The valve was installed in a vertical secticn of the drain piping to eliminate the potential for condensate pocketing upstream of the valve, and will permit the AFW steam line drains to accomplish their
~ design function.
I Safety Evaluation:
The modifications addressed by this Engineering Package did not adversely affect the I integrity of the AFW system pressure boundary, or the performance characteristics of i the . AFW pumps during accident conditions.- The new isolation valve will be administratively locked open during plant normal operation and will serve as a passive part of the ~AFW pressure boundary during those accidents that require AFW flow
- deliv'ery. . One failure mode was identified in the evaluation and it pertained to the :
increased potential for external ' system leakage. This failure mode was adequately addressed by the use of a valve that has a higher design pressure and temperature rating than.the associated AFW piping. The ability of the modified piping system to
- meet the applicable Updated FSAR loading conditions was demonstrated by l calculation. Based on the evaluation criteria contained in the Engineering Package,
.the modifications did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required
- for implementation.
33 J
I!,
?
SECTION 2 SAFETY EVALUATIONS
)
34
U ,
l l
SAFETY EVALUATION JPE-M-87-136 REVISION 1 UNIT : 3&4 APPROVAL DATE : 11/25/98 CCW BASKET STRAINER BACKWASHING Summary:
1 This safety evaluation was prepared to address the 10 CFR 50.59 criteria for l performing a " full flow" backwash on the component cooling water system (CCW) {
basket strainers. .The method proposed included closing the affected strainer inlet i valve and opening the strainer drain valve for 15 to 20 minutes. This would allow l approximately 3300 GPM of intake cooling water from the redundant header to back j ilow through the affected basket strainer under normal operating conditions. The i evaluation addressed a) the impact of the proposed alignment on CCW and turbine plant cooling water (TPCW) heat exchanger flow rates, b) the structural integrity of i the basket strainer internals, c) operability of the strainer drain valve and piping, and !
d) restoration of full intake cooling water system (ICW) flow to the CCW heat i exchangers during postulated accident conditions. :
l Revision 0 of this evaluation demonstrated that the operability of the ICW and CCW systems would not be affected by the proposed backwashing procedure as long as the differential pressure across the strainer was less than 2.0 psi prior to initiating the backwash, and full ICW flow was restored to the CCW heat exchangers within five minutes of a design basis accident. The five minute response time was based on pre- !
uprate design parameters.
Revision 1 re-evaluated the proposed backwashing procedure using plant uprate design parameters and analyses. Due to the relaxation of several ICW and CCW temperature restrictions during the uprate process, additional analytical margin was supported by this revision.
Safety Evaluation:
This safety evaluation addressed the impact on plant safety associated with diverting ICW flow from the CCW heat exchangers for basket strainer backflushing during normal operation. It specifically analyzed the affects on safety associated with manual restoration of the ICW system during accident conditions and demonstrated that the CCW system would be maintained in an analyzed configuration during an eventi capable of removing its design basis heat load. The actions and precautions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the identified actions and precautions.
35
SAFETY EVALUATION JPN-PTN-SEEJ-88-042 REVISIONS 9 and 10 UNIT : 4 APPROVAL DATES :Rev. 9 03/02/99 Rev.10 03/22/99 3 DE-ENERGlZATION OF UNIT 4 4160 VOLT ;
SAFETY RELATED BUSSES l
Summary: j 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 9 updated the electrical design configuration and clairied plant restrictions associated with implementing the bus cross-tie.
Revision 10 revised the bus loading to allow the 4D normal containment cooler to be loaded on to the 4G load center during loss of voltage conditions.
Safety Evaluation:
This safety evaluation addressed the technical and licensing requirements for the de-energization of each Unit 4 4160 volt bus and concluded that the proposed plant configuration and mode of operation was bounded by the technical specifications and did not change the analysis of accidents addressed in the FSAR or the results and conclusions of any previous safety evaluations. The actions or procedural changes identified and evaluated in this safety evaluation did not have any adverse effect on !
plant safety or plant operations. The actions and precautions identified in this safety I evaluation did not constitute an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or precautions identified in this safety evaluation.
1 I
36
l SAFETY EVALUATION JPN-PTN-SEEJ-89-085 REVISION 13 L UNIT : 3 I APPROVAL DATE : 07/30/98 l
l 1
DE-ENERGlZATION OF UNIT 3 4160 VOLT ]
SAFETY RELATED BUSSES I l
Summary:
This evaluation developed the requirements and restrictions which must be placed on i 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 En 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 i necessary to permit periodic maintenance, testing, or design modifications of the 4160 voit switchgear. De-energization of a 4160 volt bus would cause de- I l energization of the 480 volt load centers and motor control centers powered from l 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 13 updated the bus loading to coincide with recent design changes and changes in plant operating requirements.
Safety Evaluation:
This safety evaluation addressed the technical and licensing requirements for the de-energization of each Unit 3 4160 volt bus and concluded that the proposed plant configuiation and mode of operation was bounded by the technical specifications and did not change the analysis of accidents addressed in the FSAR or the results and conclusions of any previous safety evaluations. The actions or precautions identified and evaluated in this safety evaluation did not have any adverse effect on plant safety or plant operations. The actions or changes in plant procedures, identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plent technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or precautions identified in this safety evaluation.
l l
l 1
37
}
SAFETY EVALUATION JPN-PTN-SENP-95-007 REVISION 3 UNIT : 3 APPROVAL DATE : 10/02/98 OPERABILITY OF RHR -
' DURING INTEGRATED SAFEGUARDS TESTING Summary:
This: safety evaluation reviewed the 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 v,as 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 3 evaluated the' impact of performing the ESIT in Mode 6 with reactor vessel level two. feet below the reactor vessel flange. It concluded that the RHR system would remain operable as long as pump flow was maintained below 5000 gpm. ,
l Safety Evaluation:
This safety ' 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) to ensure that both RHR loops would
- remain operable during the test sequence. It concluded that the existing test l restrictions and operator actions, along with the 5000 gpm RHR flow limit imposed )
. for' Mode 6. testing,~ were sufficient to ensure that both trains of RHR would be operable at all times during the test. Since all licensing and design basis requirements )
would continue to be met during the ESIT, the proposed changes did not involve an i
l unreviewed safety.. question or require changes to plant technical specifications.
L Thus, prior NRC approval was not required to initiate the test sequences. j i
I 38 I
1 SAFETY EVALUATIONJPN-PTN-SEMS-95-023 REVISIONS 3 and 4 UNIT : 3&4-APPROVAL DATES :Rev. 3 .09/18/98 Rev.4 01/27/99
- PERFORMANCE OF MAIN STEAM SAFETY VALVE SETPOINT VERIFICATION TEST IN MODE 1
' Summary:
Main Steam Safety Valves (MSSVs) require periodic setpoint verification testing in accordance with Technical; Specification 3/4.71 to demonstrate operability. This safety evaluation was issued to address MSSV setpoint verification testing in plant operating Modes 1, 2, or 3 and to examine the impact on operability of the MSSVs with the safety valve test (SVT) rig attached to the valves. Previously, Turkey Point verified. the lift settings of the MSSVs while in Mode 3 with the tested valves declared out of service. In order to reduce critical path evolutions, a change in the testing approach was developed which would allow MSSV testing while in Modes 1,
.2, or 3. Plant Technical specifications require that all-MSSVs remain operable in Modes 1, 2, and 3. 'However, operation may continue with one or more MSSVs inoperable provided that'within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s: (1) either the valve is restored to an operable i status; (2) the power range neutron flux high trip setpoint is reduced; or (3) power is reduced to below.the specified rated thermal power level as required by technical specifications. .This testing is performed utilizing a temporary SVT rig to determine
~ the valve' lift setting. . This evaluation also examined testing while the valves remained in service... If a MSSV failed the testing acceptance' criteria, or failed to reseat, it would be declared inoperable and the applicable technical specification actions taken.
Revision 3 provided a discussion of the affects on safety from installation of the pressure transducer and to maintain NUREG 0578 commitments.
Revision.4 makes administrative changes to valve number designations to correctly
. identify applicability to both Units 3 and 4.
. Safety Evaluation:
This, evaluation examined the MSSV design bases, the seismic qualification of the Lvalves and test equipment during the test, the effect of testing on MSSV operational requirements, failure modes'and effects, plant' operating restrictions during testing,
. and technical specification requirements. The evaluation concluded that the proposed testing approach and SVT equipment had no adverse impact on plant safety or plant operations, and therefore, did not constitute an unreviewed safety question or require changes _to the plant technical specifications. Therefore, prior NRC approval was not required to implement the new test philosophy.
39
r SAFETY EVALUATION JPN-PTN-SEMS-96-003 i REVISION 1 UNIT : 4 APPROVAL DATE : 12/04/97
. 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 I 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 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 I help prevent and detect the presence of loose parts.
Revision 1 incorporated results of the S/G inspections performed during the Cycle 17 refueling outage.
Safety Evaluation:
I Previous safety evaluations documented for each S/G secondary side foreign object have considered the effects of the object upon tube integrity, chemistry, S/G i instrumentation, the main steam system, and S/G blowdown and sampling systems. This current 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 did not constitute an unreviewed safety question or require changes to the plant Technical Specifications. Therefore, prior NRC approval was not required to implement the actions identified within this evaluation.
40
I SAFETY EVALUATION JPN-PTN-SENS-96-011 l REVISION 1 UNIT : 3&4 APPROVAL DATE : 01/12/99 )
USE OF MANUAL ACTIONS TO ISOLATE THE TPCW HEAT EXCHANGERS Summary:
The purpose of this evaluation was to address the 10 CFR 50.59 criteria for ,
utilizing local manual actions to isolate intake cooling water (ICW) flow to the l
turbine plant cooling water .(TPCW) heat exchangers during accident conditions when POV *-4882 and/or -4883 are unavailable during maintenance activities.
Isolation of the TPCW system is necessary during safety injection (SI) events and loss of offsite power (LOOP) events to maintain ICW flow within design limits for post accident heat removal. Of the two accident conditions, Si actuation imposed the most stringent TPCW isolation requirement. A five minute isolation requirement was established for the system based on a thermal analysis of the ICW and component cooling water systems. The five minute time limit for TPCW isolation was conservatively imposed for both Si and LOOP events. To ensure that the five minute isolation requirement can be met, the safety evaluation requires that a dedicated Operator be stationed near the TPCW heat exchangers whenever the l POVs are removed from service.
Revision 1 re-evaluated the requirement to isolate the TPCW heat exchangers ,
during a LOOP event. It was demonstrated that TPCW isolation for a LOOP event could be performed within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> window. It also provided additional analysis to justify the acceptability of the proposed manual actions.
Safety Evaluation:
This evaluation demonstrated that the use of proceduralized manual actions to compensate for the lack of automatic actuation of POV *-4882 and/or POV *-4883 did not cause any safety limits or design limits to be exceeded. It also demonstrated that there are no postulated events or conditions that would prevent the manual actions from being accomplished. Since the proposed actions satisfied all design and licensing requirements, it was concluded that the alternate method of isolating the TPCW system during ' accident conditions did not constitute an unreviewed safety, question or require a change to the technical specifications.
Thus, NSC approval was not required for implementation of contingency measures when the POVs are removed from service for maintenance.
41
o SAFETY EVALUATION JPN-PTN SEMS-96-038-REVISION 2 UNIT : 3
' APPROVAL DATE : 01/27/99 SAFETY EVALUATION FOR UNIT 3 STEAM GENERATORS'
. SECONDARY SIDE FOREIGN OBJECTS 1
Summary: -)
-.This evaluation addressed the potential safety significance of operating the Unit 3 4 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- 1 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 3 S/G tube wall wear times; and (3) provide a single Unit 3 safety evaluation to assess '{
. and document all 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 2 incorporated results.of the S/G inspections performed during the 1998 Cycle 17 refueling outage. Wear time was'also're-evaluated for all foreign objects.
Safe +y Evaluation:
Previous safety 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 constitute an unreviewed safety question or 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
, L evaluation.
42
1 SAFETY EVALUATION PTN-ENG-SENS-96t7' !
REVISION O UNIT : 3&4 APPROVAL DATE : 03/19/98 SAFETY EVALUATION FOR NEl INITIATIVE FOR LICENSING BASIS CONFORMANCE Summary:
This safety evaluation documents the series of reviews performed at Turkey Point to comply with the Nuclear Energy Institute's (NEI) initiative for licensing basis conformance. It also documents the resolution of the various review findings. The NEl initiative requested all utilities to a) assess the adequacy programs currently in use to ensure the plant is in conformance with its licensing basis; b) assess the accuracy of the FSAR descriptions for two safety-related systems and two non-safety related systems; c) ensure any identified non-conforming or degraded ,
conditions are tracked and resolved in a timely manner; and d) communicate the I results back to NEl within six months of project commencement.
i Conformance with the plant licensing basis was assessed by sampling safety evaluations, changes to conditions of license, corrective actions for old material J conditions, procedure changes, FSAR user comments, license amendments, regulatory comrnitment changes, and minor engineering packages. The two safety l related systems reviewed were the chemical and volume control system (CVCS) )
and component cooling water (CCW) system. The two non-safety related systems were instrument air (IA) and the rtandby steam generator feedwater system. A number of FSAR discrepancies were identified in the above reviews with operability i assessments performed and FSAR user comments generated to correct the noted discrepancies. Additional items reviewed included the work around list,' night order book, technical specification position statements, procedures that had steps marked as "N/A," manually operated equipment, and the abandoned equipment program.
Safety Evaluation:
All identified FSAR discrepancies were dispositioned and none were found to impact nuclear safety or safe plant operation. Updates were prepared to correct and clarify the FSAR. Engineering assessment of the findings determined that no operability issues were involved. The actions and FSAR changes identified in this ;
safety evaluation did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required to implement the actions or changes identified within this evaluation.
1 i
I 43
SAFETY EVALUATION PTN-ENG-SEFJ-97-028 REVISION O UNIT : 3&4 APPROVAL DATE : 03/19/98 SAFETY EVALUATION FOR BEST ESTIMATE LARGE BREAK LOCA (BELOCA)
FSAR AND DBD UPDATES FOR TURKEY POINT UNITS 3&4 1
Summary:
.This safety evaluation was prepared to update the Turkey Point FSAR and Design Basis Documents (DBDs) to reflect NRC approval of the Best Estimate LOCA (BELOCA) analysis. It also confirms that residual heat removal (RHR) valves MOV-
- 744A and MOV *-7448 will be open during a large break LOCA event with loss of offsite power to permit flow to the cold legs when reactor coolant system (RCS) pressure decreases below the RHR pump shutoff pressure.
Attachment 1 documents that the two RHR valves will be fully open for up to 3.85 i seconds prior to RCS pressure decreasing to the RHR shutoff pressure. Attachment 2 contains the recommended FSAR changes. Attachment 3 contains the recommended DBD changes.
1 Safety Evaluation: !
- The proposed FSAR and DBD changes summarized the methodology and the results of the BELOCA safety analysis reviewed and approved by the NRC. The document updates did not alter the sequence of events during the accident or the equipment required to mitigate its consequences. Consequently, the document changes identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.
44
w ,
SAFETY EVALUATION PTN-ENG-SECS-97-061 REVISION 1 UNIT : 3&4 APPROVAL DATE : 03/04/99 SAFETY EVALUATION FOR CONTAINMENT POLAR CRANE MAINTENANCE INSPECTION PROCEDURE S y jry; This safety evaluation assessed the acceptability of spreading the required containment polar crane maintenance inspections over the course of an outage to reduce critical path time. The current practice of performing all of the required crane inspections at the beginning of each outage delays many of the critical path activities.
As a basis for establishing the new inspection schedule, licensing commitments and industry _ standards-were reviewed to determine the minimum set of inspections that were applicable to the polar cranes. ' The evaluation examined those inspections that had to be performed on a periodic basis and those that had to be performed on a frequent basis (i.e., monthly or daily). The periodic inspection requirements were separated into pre-service activities, preventive maintenance activities, and post-
- service activities. Pre-service inspections were considered to be valid for one year.
Preventive maintenance and post-service inspections were considered to be valid for two years. The identified activities (periodic and frequent) were incorporated into a temporary inspection procedure and _ scheduled to be performed during the Unit 4 Cycle 17 refueling outage.-
Revision 1 incorporated inspection relief for the low limit switch of the main hoist and !
clarified the inspectian frequency for the upper limit switch.
Safety Evaluation:
The containment polar cranes do not perform a safety related function so there was no safety significance associated with the proposed activity. Since all licensing commitments were maintained by the proposed inspection plan, the actions or document changes identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant technical specifications.
Therefore, prior NRC approval was not required for implementation of the phased polar crane inspection plan.
45
, SAFETY EVALUATION PTN-ENG-SENS-97-067 REVISION O UNIT : 3&4 APPROVAL DATE : 11/18/97 1
TRANSFER TO COLD LEG RECIRCULATION EOP CHANGES Summary:
The plant Emergency Operating Procedures (EOPs) related to transfer to cold leg recirculation following a loss-of-coolant accident (LOCA) have gone through a number of revisions to address high component cooling water (CCW) flow and thermal power uprating. The Operations and Training staff have identified that the current format is difficult to use and can be confusing. The purpose of this evaluation was to address changes to these procedures that will simplify their use for post-accident situations. The changes included removal of the negative response not obtained phrasing and implementation of a more logical layout. The proposed EOP changes did not impose any new operating configurations, or alter existing plant recovery schemes.
Safety Evaluation:
This safety evaluation has determined that the proposed EOP changes are consistent with FPL commitments related to transfer to cold leg recirculation, including the Westinghouse Emergency Response Guidelines (ERGS), thermal power uprate and recent NRC staff concerns in this area. These changes to the EOPs have been
- evaluated against the criteria of 10 CFR 50.59 and determined not to constitute an I unreviewed safety question, or require a change to the plant technical specifications.
Therefore, the procedure changes were implemented without prior NRC approval pursuant to the requirements of 10 CFR 50.59.
F , 46
[-'_
SAFETY EVALUATION PTN-ENG-SEES-97-094 REVISIONS 0 and 1
. UNIT : 4 APPROVAL DATES :Rev.0 12/19/97 Rev.1 08/18/98 TEMPORARY INSTALLATION OF REMOTE MONITOR
.FOR 'C' RCP OIL LEVEL VERIFICATION Summary:
.This safety evaluation addressed the temporary installation of a video transmitter, power supply, NEMA 4 enclosure, and approximately 240 feet of cable inside the Unit 4 containment to monitor the_4C reactor coolant pump (RCP) motor oil level.
Due to the-small weight involved (20 pounds) the equipment was secured to an existing steel support on the 30'-6" elevation of the containment. The video signal from the transmitter was routed to a communication box near the elevator platform on the 30'-6" elevation of' the containment building, . and connected to spare telephone leads which terminated outside containment in the cable spreading room.
Revision.1 evaluated the use of an existing ILRT cable to transfer the video feed outsi#0 containment in' lieu of the phone connection.
Safety Evaluation:
' An engineering review demonstrated:that the equipment would remain in place during a design- basis seismic event, and not damage adjacent equipment considered .. important ' to safety. 'It also demonstrated that the containment hydrogen, free volume, heat sink, and combustible loading analyses would not be adversely affected by the additional equipment due to the small mass of material
' involved. Since: the installation of video transmitter equipment and associated
cabling did not change the operation, function, or design basis of any structure, system, or component important to safety, the actions identified in this safety evaluation'did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC' approval was noi required for implementation of the actions identified within this evaluation.
, 47
I' SAFETY EVALUATION PTN-ENG SEMS97-096 REVISIONS O and 1 UNIT : 3&4 APPROVAL DATES :Rev.0 03/19/98 Rev.1 02/16/99 FIRE RATED PENETRATION SEALS j Summary:
The primary purpose of this safety evaluation was to augment the documentation of design basis and performance requirements for fire rated penetration seals and firestops (generically referred to as "penseals") and to provide for updating the
' FSAR'accordingly. A secondary purpose was to specifically identify penseals with documented bases for penseal validation. Fire rated penseals are used to seal openings in building structures which are defined as fire rated barriers. They are
, typically installed around penetrating pipe, conduit and cable tray; however, they are also used for seismic gaps, expansion joints and structural members.
! Revision 0 provided a listing of all penseals and identified those which had a specific basis for validation. Those that did not have a supporting basis for validation were subsequently evaluated and their respective design and verification bases were _ reflected in Revision 1. Attachment 1 of the evaluation provides the listing of penseals with reference to their basis for design acceptability.
Attachment 2.provides the proposed FSAR changes Attachment 3 provides the results of tests performed on those penseals that could not be verified by-inspection of generic details J
Safety Evaluation: !
This activity involved review and assessment of fire rated penseal performance capabilities and provided for updating the FSAR accordingly. The proposed FSAR i change elaborated on the design bases criteria and descriptions of penseals !
provided in Appendix 9.6A; however, the change did not modify the design or l configuration of any penseal, did not introduca eny new hazards and 'does not ;
change the methods by which penseals perform their fire protection function, as I described in the FSAR. Fire protection technical specifications are defined in the FSAR, and not in the plant technical specifications. Therefore, the FSAR change i did not constitute an'unreviewed safety question or require changes to the plant technical specifications, and therefore, did not require prior NRC approval.
48- l t
SAFETY EVALUATION PTN-ENG-SEES-98-024 REVISION O UNIT : 3 APPROVAL DATE : 04/02/98 EDG LOADING CONSIDERATIONS DUE TO MANUAL-LOADING OF ELECTRIC DRIVEN FIRE PUMP
- Summary:
This safety evaluation examined the affects of loading the electric driven fire pump
. on to its respective emergency diesel generator (EDG 3A) when the diesel driven fire
~~
pump fails or is taken out-of service, and a need for. a fire pump exists. For a postulated plant fire . event, the safe shutdown ~ analysis takes credit for either the diesel driven fire pump or the electric driven fire pump. Only one fire pump is required
.. to meet the demand for fire water during the event. Since the electric driven fire pump may not be available for those plant fire areas where Channel A power train is
-' lost or cables associated with the electric driven fire pump or 3C 480V Load Center are adversely 'affected, it is desirable to have the ability to load the electric driven fire pump on to the 3A EDG. This evaluation analyzed the effects on safety associated with manually loading 1the fire pump on to the 3A EDG. It also examined the technical and licensing issues associated with load limitations on the 3A EDG and safe shutdown analysis for Turkey Point. I An FSAR change package was provided as an attachment to this safety evaluation.
It describes that the electric driven fire pump will trip on loss of offsite power (LOOP)~and that the Operator can manually load the electric driven fire pump on to !
the 3A EDG if needed.
Safety Evaluation: I This safety evaluation provided a loading analysis of the 3A EDG when the electric driven fire pump is loaded onto the emergency bus during a LOOP event. This analysis demonstrated that the 3A EDG will be able to supply power to the fire pump if all' Appendix R loads are loaded on to the EDG in a controlled manner. It
, concluded that the proposed plant configuration and mode of operation was bounded by the existing plant technical specifications. It was also concluded that the proposed procedural changes did not have any adverse effect va plant safety or operation. Since the addition of the electric driven fire pump did not adversely ;
affect any of the EDG performance characteristics, the actions identified in this l
safety evaluation did not constitute an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the evaluated actions.
I I 49
SAFETY EVALUATION PTN-ENG-SENS-98-031 REVISION O UNIT : 3&4 APPROVAL DATE : 6/9/98 ON-LINE FILTERING OF THE CCW SYSTEM Summary:
This safety evaluation was prepared to address the 10 CFR 50.59 criteria for the temporary connection of a portable filter skid to the component cooling water (CCW) system, and on-line filtering of the CCW fluid. The function of the filter skid was to remove any organic matter that may become entrained in the coolant stream following biocide treatment. The addition of biocides to the CCW system is necessary to minimize the potential for microbiological corrosion of the heat
. exchanger tubes, and maintain long term integrity of the CCW system pressure boundary.
The filter unit was to be connected to the system at previously installed supply and return connections on the "B" header. The use of these connections required that the "B" CCW header be isolated from the surge tank during plant operation. This safety evaluation analyzed: a) the effects of a hypothetical failure of the filter unit during plant operation, b) the ability to accommodate a design basis CCW leak while operating in the proposed CCW system alignment, and c) the use of manual actions to restore the normal CCW system alignment during abnormal or accident conditions.
It was concluded that use of the filter skid on a temporary basis would not prevent accomplishment of the CCW system safety functions. To prevent any adverse interactions during severe weather conditions, the safety evaluation imposed a
. requirement that the portable filter unit be removed to a secure area.
Jafety Evaluation:
The safety evaluation demonstrates that adequate design features and administrative controls were in place to ensure that at least one CCW pump and two CCW heat exchangers would remain operable under design basis conditions.
The safety evaluation concludes that the proposed temporary connection of a portable filter unit. to the CCW system during operation does not involve an unreviewed safety question, or require a change to the plant technical specifications.
Therefore, prior NRC approval for this activity is not required.
50
y w
SAFETY EVALUATION PTN-ENG-SENS-98-033 REVISION O UNIT _
- 3&4.
APPROVAL DATE : 06/09/98 DETERMINATION AND ASSESSMENT OF COMBUSTIBLE OILS IN CONTAINMENT BUILDINGS >
1 Summary:
This evaluation was' developed to determine the potential quantities of combustible
' oils and grease in the Turkey Point containment buildings (Fire Zones 59 and 60) and to assess _the continued adequacy of the fire protection features provided in these fire zones. The largest liquid combustible source in containment is the lube oil associated with the reactor coolant pumps (RCPs). In _ complying with 10 CFR 50, Appendix R, Section ll1.0, Turkey Point Units 3 and 4 are provided with an RCP oil collection system. This evaluation provided an assessment of the RCP oil collection system's
' ability to accommodate the design basis leakage from all potential pressurized and unpressurized leakage sites of the reactor coolant pump = motor lube oil system.
Additionally, this evaluation accounted for remote filling operations due to low level lube oil alarms. Based on the evaluated combustible oil sources in the Unit 3 and 4 containment buildings, the fire protection features provided in containment remain adequate. l An FSAR change package was provided as'an attachment to this safety evaluation. I lt provided a revised description of the RCP- oil collection system and the fire hazards analysis for Unit 3 and 4 containment Fire Zones 60 and 59 consistent with this engineering assessment. !
I Safety Evaluation: 1 i
This safety evaluation documents that the fire protection features provided in l containment Fire Zones 59 (Unit 4) and 60 (Unit 3), specifically the RCP oil !
collection systems, are adequate to accommodate the experienced and estimated i l leakage of RCP lube oil during each unit's operating cycle. Appropriate procedural ;
changes to. limit the addition of lube. oil to RCP motor lower oil reservoirs are
. identified ~and evaluated in this safety evaluation. These procedure changes do not have any _ adverse effect on plant safety or plant operations since there is sufficient reserve provided in the oil collection tanks. This safety evaluation concluded that this safety evaluation did not; constitute an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions identified in this safety evaluation.
1 51
E ;
i M.FETY EVALUATION PTN-ENG-SEYS-98-036 l
REVISION O l h
i UNIT : 3&4 '
APPROVAL DATE : 9/17/98 DIAGNOSTIC TESTING OF ATMOSPHERIC DUMP VALVE CV *-1607 Summary:
' This -' safety evaluation addressed . the' installation of diagnostic > equipment on atmospheric steam' dump' valve (ADV) CV *-1607 and the performance of response time testing while the associated Unit was operating in Modes 1, 2, or 3. The test c.
.was nten i ed d to benchmark the performance of the ADVs and evaluate the affects of proposed actuator changes on valve actuation time. The test required that the ADV be cycled open and closed.for three separate test sequences. It also required that a containtnent isolation valve on the steam generator wet layup branch line of the main steam piping be opened to monitor steam pressure during the test. The evaluation examined: a) the ability of the test configuration to maintain integrity of
, the main steam system pressure boundary; b) the ability of the inservice ADVs to j cool and depressurize the reactor coolant system'(RCS) during normal shutdowns i
- and postulated accidents; c) the impact on plant safety and operation associated with' opening an ADV at power; d) the ability to re-establish containment isolation if conditions. warrant;- and c) the impact of the test configuration on the seismic.
qualification of the main steam piping.
A failure modes and effects analysis was included in the evaluation to demonstrate
.that operation'of the RCS and main steam system remained bounded by the plant m
safety. analyses during the test. Appropriate compensatory actions were imposed
~
i as necessary to ensure compliance with the plant design bases. I I
Safety Evaluation _:
This evaluation examined the main steam system design bases, the seismic aspects 3 of testing, the wffect of testing on main- steam system operational requirements, failure modes'and effects, plant operating restrictions' during testing, and applicable I technical specification requirements. The evaluation concluded that .the proposed ;
testing approach and test equipment had no' adverse impact on plant safety or plant ;
. operations, and therefore, did not constitute an unreviewed safety question or require j changes to the plant Technical specifications. Therefore, prior NRC approval was not l required for installation of. tost equipment or implementation of the testing identified
.within this evaluation. .
52
+ ,
SAFETY EVALUATION PTN-ENG-SEMS-9,8-039 REVISION O UNIT : 3&4 APPROVAL DATE : 09/17/98 i
HHSI PUMP ALTERNATIVE SHUTDOWN FOR CHARGING PUMPS Summary:
i This safety evaluation addresses the ability to use the high head safety injection I (HHSI) pumps as alternative safe shutdown equipment for a postulated fire in Fire Area N or O. Fire Area N houses the Unit 3 charging pumps and Fire Area O houses the Unit 4 charging pumps. Use of the charging pumps for safe shutdown can not be credited for a fire in these areae because insufficient separation exists between the pumps. The evaluation demonstrates that HHSI pumps, in conjunction with the reactor coolant system power operated relief valves (PORVs) can provide the required inventory and reactivity control functions necessary to achieve maintain safe shutdown conditions. It also demonstrates compliance with Sections Ill.G.3 and Ill.L of Appendix R to 10 CFR 50.
An FSAR change package was provided as an attachment to this safety evaluation to revise the description of alternate shutdown to include fire in Fire Areas O and N.
The change was a documentation change only and did not change the physical configuration of the plant or the method or limitations of equipment operation. l Safety Evaluation:
The safety evaluation addressed plant fire protection and safe shutdown capability l using the HHSI pumps and associated equipment as alternate safe shutdown equipment. Use of.the HHSl pumps upon loss of charging flow was already a l design basis for the plant and instruction for performing the reactor makeup and reactivity control functions were already addressed in plant procedures. The safety evaluation concluded that expanding the design bases of the HHSI Pumps and associated components did not alter any of the equipment operational or performance characteristics, did not affect ar, assumptions relative to fire hazard analysis or combustible loading, and did not irr le the accomplishment of post-fire recovery efforts. The safety evaluation concluded the activity did not constitute an unreviewed safety question or require a change to the plant technical specifications. Therefore, prior NRC approval was not required to implement the applicable documentation changes.
53
p SAFETY EVALUATION PTN-ENG-SEMS-98-045 i REVISION O UNIT : 3&4 APPROVAL DATE - : 10/21/98
' CONTROL BUILDING BUILT-UP ROOFING COMBUSTlBILITY Summary:
This safety evaluation was prepared lto assess the combustibility of the control building roof materials. The control building roof consists of a concrete base and a composition built-up tar and gravel covering. The covering design is consistent with ,
ithe original plant construction, and, with the exception of backfit installations for missile barriers and HVAC requirements, is mostly constructed with the original plant roofing material. Based on original design documentation and field walkdowns, this
.. evaluation provides the necessary analysis to: a) document that the control building built-up roofing composite is an equivalent National Fire Protection ~ Association
-(NFPA) 256. Class A construction with a negligible combustible load contribution to I. safe shutdown circuits; b) update the FSAR description of the control building roof configuration and the description of equipment contained in Fire Zones 106R and 118; and.'c) provide Engineering input on a proposed revision to the request for exemption for Fire Zone 106R.
. Attachment 1 to the evaluation provided a supplement to the prcviously submitted
- control building roof exemption. Attachment 2 provided the FSAR change package required to update the description of components'in Fire Zones 106R and 118.
Safety Evaluation: j The safety evaluation analyzed the existing. construction of the control building roof-and the equipment attached'to the roof structure, it also provided the necessary documentation to update the FSAR descrip': ion of components in Fire Zones 106R and 118. The safety evaluation did not implement .any changes to the plant ;
configuration, method of operation, personnel access requirements, or design bases
~
I for facilities in'these fire zones. Since the proposed documentation changes did not I impact any licensing commitments ' relative to fire protection and safe shutdown, !
the safety evaluation did not constitute an unreviewed safety question or require a ;
change:to the'~ plant technical specifications Therefore, implementation of the 1 proposed documentation chan;ps did not require prior NRC approval. j 1
1 I
54 i
SAFETY EVALUATION PTN-ENG-SENS-98-047 REVISION 1 UNIT : 3&4 APPROVAL DATE : 8/6/98 l AFW PUMP LOW-FLOW OPERATING RESTRICTIONS Summary:
This evaluation supported changes to the plant procedures that were necessary to prevent extended operation of the AFW Pumps at low-flow conditions. Extended i operation at low-flow conditions could accelerate pump wear and degradation and potentially impact reliability of the pumps to perform their intended safety functions. To prevent these conditions from occurring, this safety evaluation incorporated procedural guidance into the plant normal and emergency operating procedures to permit one or more operating AFW Pumps to be shutdown if flow delivery capability exceeded system demand. The intent of the procedural action is to shutdown one AFW Pump operating in a two pump train within one hour of the initial AFW actuation signal, followed by the shutdown of a second pump when low flow conditions warrant. The evaluation considered that the AFW Pump area may not be accessible post-accident and allowed for the ability to stop the pumps without entering an area with high dose rates. The proposed actions would permit a shutdown pump (s) to be restarted from either the Control Room or by local operator action in the unlikely event that a change to AFW requirements occurred.
Revisitan 1 of this safety evaluation provided additional documentation of circuit failure analyses and made other minor clarifications throughout the document. The changes did not alter the basis or conclusions of Revision O.
Safety Evaluation:
The procedural actions addressed in this evaluation to maintain long term operability of the AFW pumps did not create any new failure modes for the system, nor prevent accomplishment of any system safety function. The action to shutdown operating AFW pumps did not adversely affect the ability to accommodate single failures since the shutdown methods either a) prepared the pump for restart, b) permitted delayed restart, or c) permitted pump recovery. Since AFW pump performance and availability were not affected, the proposed changes did not involve an unreviewed safety question and do not require a change to the plant technical specifications.
Therefore, prior NRC approval is not required to implement the modified AFW pump operating scheme.
55
1 SAFETY EVALUATION PTN-ENG-SEES-98-049 REVISION O UNIT : 3&4 APPROVAL DATE : 09/18/98 EDG AND SAFETY RELATED BUS LOADING DUE TO INCREASED POWER RATING OF INSTRUMENT AIR DRYERS Summary:
This safety evaluation assessed the impact on plant safety and operation associated with an increased load rating of the instrument air (IA) dryer towers from 25 kW to 32 kW. This condition was discovered during performance of in-service-testing and was documented in Condition Reports 97-1123 and 98-746. The lA dryers are designed to be loaded on to the emergency diesel generators (EDGs) during the first load block. A review of the various EDG loading analyses determined that the additional 7 kW load would not affect the generators ability to: a) successfully start the first block motor loads, and b) recover voltage and frequency before the application of the next sequential load block. The safety evaluation also demonstrated that the existing _lA dryer cables and breakers could accommodate the additional load, and that the condition would not cause any of the equipment design ratings would be exceeded.
- An FSAR change package was provided as an attachment to this safety evaluation to update the applicable load tabulations to reflect the correct kW value.
Safety Evaluation:
The safety evaluation demonstrated that the load addition did not alter the design or operation of safety related equipment, did not affect any assumptions relative to accident initiators, did not impede the accomplishment of pest-accident recovery efforts, or increase the consequences of postulated accidents. Since the change did not adversely affect any of the diesel generator operational or performance characteristics, the increase in EDG loading did not constitute an unreviewed safety
- question or require changes to plant technical specifications. Therefore, prior NRC approval was not required to implement the actions identified in this safety.
evaluation.
I m j l
56
1 SAFETY EVALUATION PTN-ENG-SENS-98-050 REVISION 1
)
UNITS : 3&4 APPROVAL DATE : 02/18/99 OFFSITE DOSE CALCULATION MANUAL REVISIONS RELATED TO THE INTERLABORATORY COMPARISON PROGRAM Summary:
This evaluation supports changes to the Interlaboratory Comparison Program described in the Radiological Environmental Monitoring Program of Turkey Point's Offsite Dose Calculation Manual (ODCM). The intent of the Interlaboratory Comparison Program is to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental samples are performed as part of the quality assurance program for environmental monitoring.
The Interlaboratory Comparison Program requirements had been satisfied by participation in the Environmental Radioactivity Laboratory intercomparison Studies Program conducted primarily by the Environmental Protection Agency (EPA). The EPA discontinued this program. As the EPA program is specifically referenced in the Turkey Point ODCM, an acceptable equivalent program must be implemented and incorporated in to the ODCM. This evaluation provided the basis and analysis of affects on' safety associated with implementing a new Interlaboratory Comparison Program at Turkey Point, and revising the ODCM accordingly.
Safety Evaluation:
]
The safety evaluation documents the acceptability of the replacement Turkey Point Interlaboratory Comparison Program. The acceptabi?ity of the replacement program is based in part on the fact that it includes the monitoring of radionuclides in water, filters (air), milk, vegetation, and soil. The EPA's intercomparison program did not include media such as vegetation or soil. These additional media were included in the Turkey Point program because they represented valid radiation exposure pathways to members of the general public in the vicinity of plant site. The safety evaluation concluded that the proposed Interlaboratory Comparison Program changes were consistent with the ODCM basis and that implementation of the new program did not constitute an unreviewed safety question or require a change to technical specifications. Therefore, prior NRC approval was not required to implement the programmatic radiological monitoring changes.
57
SAFETY EVALUATION PTN-ENG-SEMS98-051 REVISION O UNIT : 4 APPROVAL DATE : 9/15/98 TEMPORARY LOWERING OF UNIT 4 SPENT FUEL POOL (SFP)
LEVELTO SUPPORT MAINTENANCE ACTIVITIES ON THE SFP COOLING.
- DEMINERALIZER RETURN VALVE 4-7988 Summary:
This evaluation was developed to examine the effects of securing the spent. fuel
. cooli ng pumps and reducing the pool level by about 1-foot in order to perform maintenance-on valve 4-7988 in the spent fuel pool (SFP) demineralizer return pioing. 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 was suspended in accordance with Technical Specification _3/4.9.11. The spent fuel pool has been evaluated for elevated pool temperatures and pool heatup from 100 F to 135 F !
was estimated to take about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, which would be a sufficient time to perform l 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 L
was establishe'd as an upper limit during the maintenance activity, at which time work would be secured and SFP cooling restored. !
i l
Safety Evaluation:
This evaluation concluded that reducing the spent fuel pool level for maintenance -
on the demineralizer return 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, i the reduced. pool water level and other actions identified in this safety evaluation
- did not constitute an' unreviewed safety question or require changes to the plant
- technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.
58 L.
p t
SAFETY EVALUATION PTN-ENG-SENS-98-053 REVISION O UNIT : 3 APPROVAL DATE : 09/29/98 I DIFFERENTIAL PRESSURE TESTING OF THE 3B FEEDWATER PUMP MOTOR OPERATED VALVE, MOV-3-1421 Summary:
This safety evaluation' analyzed the impact on plant safety and operation associated with cycling valve MOV-3-1421 open and closed per Temporary Procedure 98-053 with Unit 3 'in Mode 3 or below and Unit 4 at power. The purpose of the )
temporary procedure was to demonstrate that the applicable motor operator developed sufficient thrust and torque to operate under maximum design basis differential pressure conditions. This testing was conducted to satisfy the requirements of NRC Generic Letter 89-10.
'The temporary procedure was to be performed with Unit 3 in Mode 3 or below, utilizing the _ condensate pump (s) to establish the differential pressure across the l valve. To ensure that the minimum flow requirements of the pump were satisfied i
~
-during the test,La pump recirculation flow path was established from the condenser, through the condensate pump (s), the non-running feedwater pumps, and back to the condenser' via the feedwater recirculation valves. Appropriate precautions were established to ensure that additional feedwater could be. added to the steam
_ generators if needed to maintain level within the required range during the test.
i
- Safety Evaluation:
The analyses, evaluations and implementation instructions supporting this activity i
- ensured that no safety related systems, equipment or structures .were adversely {
affected by the test sequence. Operability of the auxiliary feedwater system was not impacted by the test boundary. No new failure modes were created and no adverse interactions with other equipment important to safety were introduced. The actions or temporary plant changes identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant technical specifications.
.Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.
59 l
1 1
SAFETY EVALUATION PTN-ENG-SENS-98-054 REVISION O l UNIT : 3 ;
l~
APPROVAL DATE : 10/13/98 DIFFERENTIAL PRESSURE TESTING OF MOTOR OPERATED VALVE MOV-3-856A Summary:
L This. safety evaluation analyzed ,the impact on plant safety and operation associated L
with cycling valve MOV-3-856A open and closed per Temporary Procedure 98-039 l
with Unit .3 in Mode 5 or 6 and Unit 4'. at power. The purpose of the temporary procedure was to demonstrate .that the applicable motor operator developed sufficient thrust' and torque to operate under maximum design basis differential
- pressure conditions. This testing was conducted to satisfy the requirements of NRC Generic Letter 89-10. l l
The temporary procedure was to be performed with Unit 3 in Mode 5 or 6, utilizing a high head safety injection (HHSI) pump operating in its inservice test alignment to
- establish the differential pressure across the valve. To ensure that the minimum flow j requirements of the pump were satisfied during the test, a pump racirculation flow j path was established from the. Unit 4 refueling water storage tank (RWST), through j the' applicable Unit 3 high head safety injection pump and back to the Unit 4 RWST L through MOV-4-856A and B. Appropriate precautions were established to: a) ensure
- that the Unit 4 RWST volume would not decrease below the minimum technical specification requirement during the test due to the cycling of MOV-3-856A, and b) ensure' that the flow path to the Unit 3 RWST would be isolated during accident L conditions. A pressure transducer.was temporarily installed upstream of the MOV to ;
record the test pressures.
i
' Safety Evaluation:
1 . The analyses, evaluations and implementation instructions suppor o ng this activity L
ensured that no safety related systems, equipment.or structures were adversely affected by the test sequence, Temporary installation of the pressure transducers ;
. for testing purposes did not result in a change in the design function in any of the associated . systerns. No new -failure modes were. introduced and no adverse interactions 1 with other; equipment important to safety were introduced. The i l actions or temporary plant ' changes identified in this safety evaluation did not constitute an unreviewe'd safety question or require changes to the plant technical 1 specifications. Therefore, prior NRC approval was not required for implementation l of the actions or changes identified within this evaluation.
1
-60
SAFETY EVALUATION PTN-ENG-SECS-98-058 REVISION O
. UNIT .. : 3
. APPROVAL DATE : 10/21/98
. 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 acticns and restrictions were identified for implementation within the evaluation.
' Safety Evaluation:
The safety . evaluation concluded that the proposed items identified within the safety evaluation can' safely remain- within containment during all modes of operation, provided that all 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 for Unit 3. Consequently, the '
requirements and restrictions identified in this safety evaluation did not constitute ;
- an. unreviewed safety question or require changes to the plant technical '
specifications. Therefore, prior NRC approval was not required for implementation of the requirements or. restrictions identified within this evaluation.
61 J
r l
j SAFETY EVALUATION PTN-ENG-SENS-98-063 i REVISIONS 0 and 1 4
g UNIT : 3&4 APPROVAL DATES :Rev.0 10/15/98 ,
Rev.1 03/24/99 j FREEZE SEAL FOR REACTOR COOLANT PUMP SEAL INJECTION NOZZLE INLET FLANGES l l
Summary: .;
l j
A This safety evaluation was prepared to assess the performance and use of freeze {
. seals when' conducting repairs on the reactor coolant pump (RCP) seal injection line l
. flanges. .it was generated in response to an identified gasket leak on the 3C RCP '
seal injection inlet flange. To avoid plant operation in a reduced inventory condition, it was proposed that a freeze seal be used to maintain reactor coolant system pressure boundary integrity during the repair process. The evaluation examined: a) the impact of cryogenic temperatures on the piping material, b) the j weight of the . freeze seal jacket on the load bearing capability of the injection line '
piping, and c) the impact of reduced piping temperatures on the repair process. i The controlled plant procedure governing freeze seal application was referenced in j the evaluation, and contingency - plans were established to restore ' pressure !
boundary integrity for the open system upon indication of freeze seal deterioration. l Application of the freeze seal was prohibited during fuel movement or with the
' transfer canal gate valve open.
, Revision 1 expanded.the applicability of the evaluation to all seal injection flanges on l all RCPs. l
. Safety Evaluation:
1
-This evaluation addressed the temporary uncoupling of the RCP seal injection nozzle
~ flange ~ to - replace . a gasket, the impact ~ on ' plant operation, and the various ;
precautions imposed to ensure the safe conduct of maintenance. Strict controls ,
1were imposed ' on the freeze seal process and contingency. measures were j '
developed to. establish pressure boundary integrity for the open system should the
~
' freeze seal start to thaw.
Based on the precautions identified, the evaluation .
concluded that the maintenance could be performed, and that this activity did not involve an unreviewed safety question 'or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the activities identified within this evaluation.
o 62 :
F 1 e
l SAFETY EVALUATION JPN-PTN SEYS-98-069 REVISION O !
l l
UNIT : 3&4 i APPROVAL DATE : 04/03/99 I l
FIRE PROTECTION j SURVElLLANCE REDUCTION TASK !
Summary: i i
This safety evaluation provided the basis for several fire protection surveillance I procedure changes that were initiated to reduce maintenance costs. The reliability and availability of-these fire protection features is maintained and enhanced by !
periodic inspection and testing. A selected set of plant procedures were reviewed to
{
determine if surveillance frequency extensions could be adopted without affecting the reliability or availability of the equipment. This evaluation provided an assessment of the selected procedures and the technical justification necessary to support a surveillance interval extension based on current NFPA codes, NML/NEIL requirements, past performance data, and experienced engineering judgement.
An FSAR change package was provided as an attachment to this safety evaluation i to update Section 7.0 of Appendix 9.6A to reflect the revised surveillance !
frequencies.
4 Safety Evaluation:
This safety evaluation provided an analysis of the potential impact to systems, !
structures, and- components due to the extension of certain fire protection i surveillances. It. confirmed that implementation of the proposed procedure changes 1 would not affect the validity of the plant fire protection program or the overall safe operation of the plant. The changes identified and evaluated in this safety evaluation did not have any adverse impact on plant safety or plant operations. The document changes did not constitute an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required to implementation of the changes identified in this safety evaluation.
63
SAFETY EVALUATION PTN ENG-SENS-98-071 REVISION O l UNIT : 3&4 APPROVAL DATE : 2/18/99 FSAR ACCURACY REVIEW CHANGES FOR CHAPTER 1 Summary:
Industry events and regulatory concems have resulted in an increased emphasis by the NRC on the accuracy of the facility and procedure descriptions in the Final Safety Analysis Reports (FSAR). FPL has performed several self-assessments of the Turkey Point FSAR for accuracy over the last several years. Although these self-assessments did not identify significant concerns, a number of FSAR discrepancies were identified. In the Turkey Point response to the NRC 10 CFR 50.54(f) request for information regarding the adequacy of and availability of design basis information, FPL committed to perform an FSAR assessment using an approach outlined in NEl 96-05. FPL also committed to perform an additional detailed review of portions of the FSAR over a two year period to identify and correct documentation discrepancies.
The scope of the detailed review of the entire FSAR is described in FPL letter L 143 dated July 1,1997.
FSAR Chapter 1 provides a basic overview of the plant design and construction. The review of Chapter 1 identified a number of editorial discrepancies and minor technical discrepancies. No operability issues were identified as a result of this review.
Safety Evaluation:
This review has determined that the identified FSAR discrepancies do not impact safe operation of the plant, do not constitute an unreviewed safety question and do not require a change to the Technical specifications. Consequently, pursuant to the requirements of 10 CFR 50.59, the resulting updates to the Turkey Point FSAR for correctness and clarification can be made and do not require NRC approval prior to implementation.
64
SAFETY EVALUATION PTN-ENG-SENS-98-072 REVISION O UNITS : 3&4 APPROVAL DATE : 02/18/99 INSTALLATION AND OPERATION OF FEEDWATER ULTRASONIC FLOW MONITORING EQUIPMENT Summary:
This evaluation supported the temporary installation of ultrasonic flow test equipment to allow an assessment of feedwater venturi fouling during normal plant operation.
The test equipment used was a Caldon high precision ultrasonic flowmeter, which was mounted on the transducer mounting assembly. The transducer mounting assembly was clamped around the main feedwater piping and connected to dedicated data collection equipment. - The non-intrusive nature of the test equipment installation precluded any adverse effects on plant operation or the response of the plant to accident conditions. The evaluation examined the installation and (pipe) surface preparation requirements for the monitoring equipment to ensure that the integrity of the feedwater piping would not be impacted during the data collection period, it also evaluated the impact of the installed equipment on the seismic qualification of the feedwater piping and associated supports.
Safety Evaluation:
This evaluation considered the seismic effects of the transducer mounting on the seismic qualifications of the feedwater piping and associated supports. With the transducer mounted around the outside of the piping, the pressure retaining capability of the piping was not affected. The data collection equipment had no potential for seismic interaction with any safety related equipment. No plant restrictions were identified, and this configuration did not affect plant technical specifications. Consequently, the special test equipment did not adversely affect any safety related equipment or functions and the proposed testing configuration identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the technical specifications. Therefore, prior NRC approval was not required for installation of test equipment or implementation of the testing identified within this evaluation. ;
I i
l 65
[
t SAFETY EVALUATION PTN-ENG-SEFJ-99-001 REVISION O UNIT : 4 APPROVAL DATE : 2/25/99 TEMPERATURE / POWER COASTDOWN FOR TURKEY POINT UNIT 4 CYCLE 17 Summary:
The purpose of this evaluation is to support a temperature / power coastdown at the end of Cycle 17 for Turkey Point Unit 4. The proposed coastdown is needed to overcome a cycle energy shortfall and allow the plant to continue operation until the scheduled end-of-cycle date of March 15, 1999. The proposed coastdown consists of a 5 degree-F reduction in RCS Tavg at a rate of approximately 1% per day followed by a power reduction of about 5% at a rate of approximately 1% a day. The total cycle exposure is not to exceed 12,396 effective full power hours, i
The purpose of this safety evaluation was to allow Turkey Point Unit 4 to extend I
the length of Cycle 17, after reaching the end of reactivity at nominal hot full power conditions, by using a combined Tavg and power coastdown. The reactivity necessary to extend operation was obtained from the negative moderator temperature and power coefficients. Because these coefficients were negative, a decrease in either the moderator average temperature or the reactor power would result in a positive reactivity addition to the core, offsetting the reactivity loss from burnup. The Tavg/ power coastdown was started at the end of Cycle 17 by reducing primary Tavg by 5 F at a rate of approximately 1 F per day. The temperature coastdown was followed by a power reduction of 5% at a rate of approximately 1 % per day.
Safety Evaluation: '
This safety evaluation demonstrated that the planned temperature / power i coastdown did not affect any assumptions relative to accident initiators, did not j impede the accomplishment of post-accident recovery efforts, or increase the consequences of postulated accidents. Since plant design requirements continued to be met and the integrity of the reactor coolant system pressure boundary was -
l not challenged, it was concluded that the assumptions employed in the calculation of offsite radiological doses remained valid. The actions or plant procedure changes )
identified in this safety evaluation did not constitute an unreviewed safety question !
or require changes to the plant technical specifications. Therefore, prior NRC i approval was not required to initif,te the temperature / power coastdown.
66
- SAFETY EVALUATION PTN-ENG-SEFJ 99-002 l REVISION O j UNIT : 3&4 APPROVAL DATE : 4/1/99 ;
i IMPLEMENTATION OF THE SINGLE-POINT INCORE / EXCORE CALIBRATION l
Summary: - '
Turkey Point Technical specifications require that an incore/excore calibration be performed on a quarterly basis. This test generates the calibration data for: ;
i a) the axial flux difference (AFD) control room meter, b) the K values of the process computer for AFD calculation, and i
c) the G values of EAGLE 21 for the AFD penalty function determination of the OTAT trip setpoint.
The full _ power flat power currents are also determined during the calibration of these instruments.
This calibration is presently performed using a multi-point technique where a xenon oscillation is induced in the reactor core. During the xenon oscillation, power range detector currents, calorimetric power and flux maps are obtained. The analysis of this data then generates the calibration constant to be used.
Westinghouse has developed a single-point calibration technique which eliminates the axial xenon oscillation that is presently used in the multi-point technique. The single point provides calibration data which has been shown to be equivalent to those derived from the current multi-point methodology.
i i e ,
Safety Evaluation: I; > ;
This safety evaluation addres2ed the applicability of the ' single-point methodology for Turkey Point and implementation of the revised calibration technique. It concluded that use of the revised technique would not impact the operation or design bases of the incore and excore detectors. Since no new failure modes were ?
introduced by the new methodology, implementation at Turkey Point did not represent an'unreviewed safety question or require a change to the plant technical specifications. Accordingly, prior NRC approval is not required to implement the single-point calibration technique.
67
l
' SAFETY EVALUATION PTN-ENG-SEMS-99-002 REVISION O UNIT : 3&4 APPROVAL DATE : 3/12/99 FSAR ACCURACY REVIEW CHANGES FOR CHAPTER 6 Summary:
Industry events and regulatory. concerns have resulted in an increased emphasis by the NRC on the accuracy of the facility and procedure descriptions in the Final Safety Analysis Reports (FSAR). FPL has performed several self-assessments of the Turkey . !
Point FSAR for accuracy over the -last 'several years. Although. these self-assessments did'not identify significant concerns, a number of FSAR discrepancies were identified. In the Turkey Point response to the NRC.10 CFR 50.54(f) request ,
for.information regarding the adequacy of and availability of design basis information, I FPL committed to perform an FSAR assessment using an approach outlined in NEl 96-05. FPL also committed to perform an additional detailed review of portions of the FSAR over a two year period to identify and conect documentation discrepancies.
"The scope of the detailed review of the entire FSAR is described in FPL letter L 4 143 dated July 1,1997.
FSARL Chapter 6 provides a basic overview of .the Engineered Safety Features including the Safety injection System, the Emergency Containment Cooling System and Filtering System, Containment Spray System, and Leakage Protection System.
The review of Chapter 6 identified a' number of editorial discrepancies and minor technical discrepancies. One technical discrepancy identified is being assessed under a separate 50.59 safety. evaluation. ' No operability issues were identified as a result of this review.
Safety Evaluation:
- This' review has determined that the identified FSAR discrepancies and the proposed ; changes. associated with these in this evaluation do not impact safe operation of the plant, do not constitute an unreviewed safety. question and do not
- require a ; change to the Technical specifications. Consequently, pursuant to the requirements of 10 CFR 50.59, the resulting updates to the Turkey Point FSAR in this evaluation. for correctness and clarification can be made and do not require NRC approval prior to implementation. ~ One technical discrepancy identified is being assessed under a separate 50.59 safety evaluation.
68
i SAFETY EVALUATION PTN-ENG-SENS-99-OO8 l REVISION O !
I UNIT : 4 APPROVAL DATE : 3/29/99 CONDUCTING RCS FILL AND VENT ACTIVITIES DURING ENGINEERED SAFEGUARDS INTEGRATED TESTING l
- Summary:
The_ Engineered -Safeguards integrated Test (ESIT) is performed at the end of each refueling outage.to demonstrate that'the accident mitigating equipment is functioning properly prior to a plant startup. Over the past several years, it has been general practice to perform.the test early in the post-refueling startup sequence while the RCS is depressurized in a " loops not filled" condition. Startup activities such as fill and vent' would typically be performed after successful completion of the ESIT and
. generally take 2 - 3 shifts to complete. In an effort to improve the post-refueling
~
startup schedule, this safety evaluation looked at performing RCS fill and vent p activities during,' or prior to, the safeguards test. It examined the impact of l
performing the ESIT in Mode 5 with~ the RCS pressurized. It also examined the command and control aspects associated with integrating the two major startup evolutions together.
Each system and' component utilized in the fill and vent process was reviewed to ensure that the required equipment would be available during the various safeguards tests. Appropriate operating restrictions were established to prevent the RCS from i
. exceeding the overpressure mitigating system actuation setpoint when the various pumps connected to_ the RCS start and stop under the simulated accident signals
- with water solid conditions in the pressurizer.
The evaluation demonstrated that integrating RCS fill and vent activities with the ;
post-refueling safeguards test did not introduce any new failure modes for the RCS or i
'its support systems. The RCS would continue to be operated within analyzed limits. j i
Safety Evaluation:
The safety evaluation demonstrated that RCS fill and vent activities can be performed during those windows of opportunity when actual safeguards train testlig is not in progress and that the integration of the two activities did not involve an unreviewed safety question, or require changes to the plant technical specifications. Therefore, l prior NRC approval was not required to implement the ESIT procedure changes.
69
f SAFETY EVALUATION PTN ENG-SEMS-99-010 REVISION O UNIT : 3&4 APPROVAL DATE : 03/02/99 RCS CHEMICAL DEGASSING Summary:
This safety evaluation analyzed the impact on plant safety and operation associated with using hydrogen peroxide to chemically remove dissolved hydrogen from-the reactor coolant during plant cooldowns. Hydrogen peroxide is routinely used in the reactor coolant system (RCS) at cold shutdown conditions to provide a controlled solubilization of radio-cobalt for subsequent removal via the chemical and volume control system (CVCS) demineralizers. Removal of radio-cobalt limits radiation exposure from reactor coolant borne radio-cobalt sources when the plant is shut down. The proposed chemical degassing process extends the hydrogen peroxide treatment to provide for reactor coolant dissolved hydrogen removal. Industry experience has demonstrated that hydrogen petoxide will react rapidly with dissolved hydrogen in cold borated coolant, in near stoichiometric proportions, with pure (unborated) water as the product. The process enables RCS degassing to be completed in parallel with plant cooldown. The evaluation addressed: a) the potential to form flammable mixtures in the RCS and CVCS gas spaces, (b) the impact on core reactivity caused by the reaction product (pure water), and c) the potential to increase process instrument corrosion. I Safety Evaluation:
This safety evaluation defined the necessary plant configuration, precautions and method of adding hydrogen peroxide to the RCS that will ensure safe and efficient chemical degassing. It demonstrated that the addition of hydrogen peroxide to accomplish the degassing function did not affect any assumptions relative to accident initiators, did not impede the accomplishment of post-accident recovery efforts, or increase the consequences of postulated accidents. Since plant design requirements continued to be met and the integrity of the reactor coolant system '
pressure boundary was not challenged, it was concluded that the actions or plant j procedure changes identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant technical specifications. ;
Therefore, prior NRC approval was not required to implement the alternate RCS degassing procedure.
I 70
SAFETY EVALUATION PTN-ENG-SECS-99-012 REVISION O
)
1 UNIT : 3&4 APPROVAL DATE : 03/12/99 USE OF ACOUSTIC EMMISSON TECHNOLOGY AS AN ALTERNATE METHOD FOR NDE OF SPECIAL LIFTING DEVICES Summary:
This safety evaluation addressed the use of acoustic emission (AE) technology as an I alternate method of performing in-service inspection of special lifting devices. It provided the technical and licensing bases necessary to use AE technology as an alternative to conventional nondestructive examination (NDE) methods on the reactor vessel head and internal lift rigs and the reactor coolant pump (RCP) motor lift rig. i Both the NRC and EPRI have recognized AE technology as an acceptable NDE '
method. The AE technology is capable of identifying and locating active flaws present in a structure under load. However, it is unable to discern the type and size of a flaw. After a potential flaw is detected and located using AE technology, a measurement and characterization of the flaw would have to be accomplished using an alternate NDE method. This evaluation concludes that the AE technique provides assurance that any existing flaw in the critical welds and parts of the special lifting devices will be detected, and that the process will not affect the integrity of the load bearing enmponents.
An FSAR change package was provided as an attachment to this safety evaluation I to allow the use of the AE technology as an acceptable alternate method of inspection.
Safety Evaluation:
The safety evaluation demonstrated that the use of AE technology to perform volumetric examinations of special lifting devices would not adversely impact plant safety or operation. Since the load restraining characteristics of the special lifting devices were not affected, it was concluded that the use of AE methodology did not constitute an unreviewed safety question or require a change to the plant technical specifications. =Therefore, prior NRC approval was not required to implement the alternate examination technique.
4 71
SAFETY EVALUATION PTN-ENG-SEES99-017 REVISION O UNIT : 4 APPROVAL DATE : 3/18/99 TEMPERATURE MONITORING EQUIPMENT IN CONTAINMENT DURING ALL MODES OF OPERATION Summary:
This evaluation addressed the acceptability of leaving temperature monitoring equipment inside the Unit 4. containment during all modes of operation. The monitors were installed to collect temperature data at specific locations inside conta.1 ment to evaluate whether the qualified life of installed equipment can be extended to a 60-year plant life. The specific locations were selected to monitor equipment that would have a significant cost increase due to the increased maintenance requirements in going from a 40-year to 60-year plant life. The equipment was secured to existing plant structures using fasteners and fastening techniques approved for containment service. 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 equipment addressed in the evaluation would remain in place during plant operation, both generic and specific actions and restrictions were identified for implementation within the evaluation.
I Safety Eveluation:
The safety evaluation concluded that the proposed items identified within the safety evaluation can safely remain within containment during all modes of operation, provided that all the restrictions and requirements identified within the l 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 4. Consequently, the requirements and restrictions i identified in this safety evaluation did not constitute an unreviewed safety question i 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.
I i
72
SAFETY EVALUATION PTN ENG-SENS-99-018 REVISION O UNIT : 4 APPROVAL DATE : 03/12/99 USE OF FREEZE SEAL IN SUPPORT OF MAINTENANCE ON RV-4-791E Summary:
4 This safety evaluation addressed the use of a freeze seal to support scheduled maintenance on component cooling water (CCW) system relief valve RV-4-791E.
Valve RV-4-791E is installed downstream of the reactor coolant pump (RCP) seal l water heat exchanger. The inlet of the relief valve is located upstream of the heat exchanger outlet throttle valve. The relief valve discharge is routed to the downstream (lower pressure) side of the throttle valve. The freeze seal was applied to the relief valve discharge line so that continued CCW cooling to the non-regenerative heat exchanger, spent fuel pool heat exchanger, and charging pump coolers could be maintained during the repair process. The controlled plant procedure governing freeze seal application was referenced in the evaluation, and contingency plans were established to restore pressure boundary integrity for the open system upon indication of freeze seal deterioration.
Safety Evaluation:
The freeze seals were relied on to perform a CCW system boundary function during i the short repair duration. The strict controls imposed on the freeze seal process, the contingency measures, relatively low pressure of the contained fluid, and small i size of the piping opening ensured that all CCW safety functions would remain unimpaired throughout the installation. Based on the precautions identified in this safety evaluation, it was concluded that the freeze seal could be performed, and that the activity did not involve an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this avaluation.
73
~
SAFETY EVALUATION PTN-ENG-SENS-99-024 REVISION O UNIT - : 3&4 APPROVAL DATE : 3/12/99 LEAK INSPECTION OF RHR (PIGGY-BACK)
RECIRCULATION FLOW PATHS
'l .
Summary:
This safety evaluation addressed the impact on plant safety and operation associated with performing a leak test of the post-LOCA
- piggy-back" recirculation flow path at near residual heat removal (RHR) pump shutoff head conditions. The testing was required to satisfy FSAR. commitments related to relative leak tightness' of the
~
external recirculation loop piping. ~The test boundary addressed in this safety evaluation extended from valves MOV *-863A/B to the suction of the high head
~ safety injection (HHSI) pumps and containment spray pumps. The evaluation established the appropriate test pressure, RHR pump operating requirements, system
- alignments, and plant restrictions that must be met to perform the test. -It also addressed the ability to conduct the test with an external pressure source in lieu of using the' discharge head of an operating RHR pumps. Since any recirculation loop leakage in the test mode would represent' a reduction in reactor coolant system inventory, this safety evaluation required that the unit in test be defueled prior to performing the test with an RHR pump.
'An FSAR_ change package was provided as an Attachment to this evaluation to allow j the use of an external pressure source, in lieu of. the installed ' RHR pumps, to i pressurize portions of the recirculation piping.
' Safety Evaluatior.: !
This evaluation defined the requirements needed to satisfy the FSAR leak inspection commitment. An assessment of the' fluid conditions concluded that the test would not ' adversely impact the integrity of any component included within the test boundary. . The actions or temporary plant conditions identified in this safety l evaluation did not constitute an unreviewed safety question or require changes to the ;
' plant Technical Specifications. Therefore, prior NRC approval was not required for !
implementation.of the actions or tempoiary plant conditions identified within this evaluation.
? -
U 9.
74 i
3 SAFETY EVALUATION PTN-ENG-SECS-99-025
' W,
- r REVISIONS 0 and 1 L/. .,
UNIT' : 4 APPROVAL DATES :Rev.0 03/31/99 Rev.1 04/06/99 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 .4 ' containment. structure during all modes of plant
- operation. The items to be stored, and the storage locations within the Unit 4 containment, were specifically -identified within the evaluation. The purpose of leaving these_ tools and' equipment within containment following refueling outages L
was to reduce the_ usage demand on the Unit 4 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 l . addressed in the evaluation were safely stored during plant operation, both generic and specific actions and restrictions were identified for implementation within the ,
evaluation.
l Revision 1 evaluated the-impact of leaving two additional tools in containment.
These additional tools were inadvertently left inride containment when the equipment
. hatch was closed..
Safety Evaluation:.
l l
l , IThe safety evaluation cor.cluded_ that the proposed items identified within the l
- safety f evaluation can safely remain within . containment during all modes of operation, provided that all 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 1 compromise the ~ safety; and~ licensing bases for Unit 4. Consequently, the requirements ~and' restrictions identified in this safety evaluation did not constitute an;unreviewed safety question or require changes to - the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the requirements or. restrictions identified within this evaluation. l I
-75 ;
'tn SAFETY EVALUATION SECL-99-034 REVISION O a , UNIT .
- 4 q APPROVAL DATE : 4/03/99 FOREIGN MATERIAL IN REACTOR VESSEL Summary:
This safety evaluation was prepared- by Westinghouse Electric Corporation to )
assess the impact on plant safety and operation associated with a rubber glove or small piece of ' polyethylene located under the reactor vessel lower core plate. The foreign material was sighted during the Unit 4-Cycie 18 refueling outage but could
- not be recovered without disassembling the reactor vessel. The safety evaluation provided an assessment of the potential impact of the unrecovered object on the operability and integrity of the reactor coolant system (RCS) and interfacing safety-related auxiliary systems during' future operating cycles. The assessment considered the potential-impact on materials compatibility, fuel integrity, core thermal-hydraulics, core physics characteristics, RCS components, auxiliary components, and instrumentation and control systems.
" Safety Evaluation:
The safety evaluation examined the physical changes that would occur to rubber and polyethylene during RCS heatup to normal operating temperatures, Rubber was expected to initially soften.but remain mostly intact with degraded physical properties. Exposure .to elevated temperatures and radiation would eventually cause the rubber to disintegrate into smat pieaes. Removal of the material would
- eventually be removed via the_ chemical and volume control system. Polyethylene was expected to soften during heatup and completely melt prior.to reaching the normal RCS operating . temperature. It was further expected that the melted
. polyethylene material' would become dispersed _into the RCS and deposited on
. cooler surfaces under low flow conditions. The safety evaluation demonstrated that these phenomena would not affect the operability or integrity of the reactor fuel, RCS, or the interfacing auxiliary systems. In addition, the decomposition of rubberLor polyethylene material due to extended thermal and radiation exposure
-- would not alter the ' results of any previously _ performed radiological dose calculations. Based on this assessment, the evaluation determined that the presence 'of'a' rubber glove or.small piece of polyethylene ~ material in the RCS did not constitute an unreviewed safety question or require changes to the plant technical- specifications. Therefore, prior NRC approval was not required for
' continued operation of the plant. '
76 o
SAFETY EVALUATION PTN ENG-SEMS-99-035 REVISION O UNIT : 4 APPROVAL DATE : 3/20/99 ALTERNATIVE RHR FLOW PATH AND OTHER PLANT CONDITIONS TO SUPPORT MOV-4-865B REPAIR ACTIVITIES Summary:
i This safety evaluation examined the impact on plant safety and operation associated with an alternative residual heat removal (RHR) system flow path during refueling i operations when 23 feet of water exists above the reactor vessel flange. The alternative RHR system alignment was required to permit maintenance activities on the 4B safety injection accumulator discharge isolation valve, MOV-4 865B. The normal refueling alignment is to provide RHR flow to both the A and B RCS cold legs. i The proposed alternative alignment involved isolating flow to the B reactor coolant system (RCS) cold leg and directing all RHR return flow through the A cold leg.
Surveillance tests performed each refueling outage verified that the alternative configuration would provide the required cooling flow to satisfy plant technical specification requirements. Additionally, an engineering review of the system hydraulics demonstrated that the use of a single return point for the RHR cooling flow would maintain adequate core flow distribution. Stringent leakage limitations were imposed on the RCS and RHR valves used to provide the repair boundary, to ensure that the refueling cavity inventory _would not be compromised.
Safety Evaluation:
This evaluation demonstrated that the operational and performance requirements of !
the RHR system would not be affected by the alternative alignment. Adequate core flow distribution and refueling cavity integrity would be maintained during the duration of the MOV-4-8658 process. It concluded that the temporary changes did !
not create any new failure modes for the RCS or RHR system, and that the systems would be restored to their original design condition upon completion of the maintenance activity. The actions and precautions identified in the safety evaluation l did not impact safe operation of the plant, did not constitute an unreviewed safety question, and did not require a change to the technical specifications. Therefore, prior NRC approval for this activity was not required.
77 i
E 1 l
l l
l SECTION 3 RELOAD SAFETY EVALUATIONS i
I j
i i
l l
i i
l 78 L i
m PLANT CHANGE / MODIFICATION 98-014 UNIT : 3 TURN OVER DATE : 01/25/99 TURKEY POINT UNIT 3 CYCLE 17 RELOAD DESIGN Summary:
This Engineering Package provided the reload core design for the Turkey Point Unit 3 Cycle 17 reload. The primary design change to the core for Cycle 17 was the replacement of 48 irradiated assemblies with 48 fresh Optimized Fuel Assembly (OFA) Region 19 fuel assemblies. Similar to past reloads, these fresh assemblies were all Debris Resistant Fuel Assemblies (DRFA) and all contained a nominal 6-inch axial blanket of natural UO, pellets at both the top and bottom of the fuel stack. The maximum fuel enrichment was 4.4 w/o which was the first use of fue! with an enrichment above 4.2 w/o at Turkey Point.
Minor changes were made to the Region 18 fuel assembly design for use in Region
- 19. These changes included the use of ZlRLO alloy for the fuel rod cladding, guide tubes, instrument thimbles, and mid-span spacer grids. A variable pitch fuel rod plenum spring was used and the fuel rod length was increased by 0.372 inches. The helium backfill pressure was also reduced to 100 psig. Additionally, the mid-span spacer grid sleeve length was decreased by 0.1 inches and the inside diameter was
. increased by 0.008 inches. None of these manufacturing-related design changes had any impact on fuel performance.
Cross core fuel bundle shuffles were utilized in the Cycle 17 loading pattern to minimize potential power asymmetries. The fuel was arranged in a low leakage pattern with no significant differences between the Cycle 16 and Cycle 17 patterns.
Safety Evaluation:
The Unit 3 Cycle 17 reload core design was evaluated by FPL and by the fuel s
supplier, Westinghouse Electric Corporation. The Cycle 17 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 SAR. These changes had no impact on fuel rod performance, dimensional stability or core operating limits. The Cycle 17 core reload did not have any adverse effect on plant safety or plant operations. Consequently, the Cycle 17 core reload package did not involve an unreviewed safety question or require changes to plant technical specifications.
Therefore, prior NRC approval was not required for implementation. J 79
PLANT CHANGE / MODIFICATION 98-015 UNIT .: 4 TURN OVER DATE- .:. 07/19/99 TURKEY POINT UNIT 4 CYCLE 18 RELOAD DESIGN Su'mmary- I This Engineering Package provided the reload core design for the Turkey Point Unit 4 Cycle ~ 18 reload. The primary design change to the core for Cycle 18 was the replacement of 57 irradiated assemblies with 57 fresh Optimized Fuel Assembly (OFA) Region 20 fuel assemblies. Similar. to past reloads, these fresh assemblies were all Debris Resistant Fuel Assemblies (DRFA) and all contained a nominal 6-inch axial blanket of natural UO2 pellets at both the top and bottom of the fuel stack. The maximum fuel enrichment .was 4.4 w/o which was the first use of fuel with an enrichment above 4.2 w/o for Unit 4.-
s ,
Minor changes were made to the. Region 19 fuel assembly design for use in Region
- 20. These changes included the use of ZlRLO alloy for the fuel rod cladding, guide tubes, instrument thimbles, and mid-span spacer grids. A variable pitch fuel rod plenum spring was used and the fuel rod length was increas' e d by 0.372 inches. The
- helium backfill pressure was reduced to 100 p'sig. The mid-span spacer grid sleeve length was decreased by 0.1 inches and the inside diameter was increased by 0.008 inches. Additionally, annular cellets were used to provide the axial blanket of natural uraniurr at the top and bottom of each feel rod. None of these manufacturing-related design changes had any. impact on fuel performance.
Cross ' ' core fuel - bundle. shuffles .were utilized to minimize potential- power
. asymmetries and a low. leakage loading pattern was utilized similar to past core designs.
I Safety Evaluation:
The Unit 4 Cycle 18. reload core design was evaluated by FPL and by the fuel .
supplier, Westinghouse Electric Corporation. The Cycle 18 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 SAR. These changes had no impact on fuel' rod performance, dimensional stability or core operating limits. The Cycleu18 core reload did not have any adverse effect on plant safety or plant
. operations. Consequently, the Cycle 18 core reload package did not involve an unreviewed safety. question or require changes to plant technical specifications.
Therefore, prior NRC approval was not required for implementation.
80
I SECTION 4 REPORT OF POWER OPERATED RELIEF VALVE (PORV) ACTUATIONS 81
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- II.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 A PORV actuation occurred on February 16,1998 during a reactor trip from 100%
power. Two PORVs lifted during the transient.
Unit 4 No PORV actuations have occurred on Unit 4 between October 14,1997 and April 8, 1999.
i e l 82
)
l i
I SECTION 5 i i
STEAM GENERATOR TUBE INSPECTIONS FOR TURKEY POINT 1
I l
1 l
l 83 l
FORM NIS-88 OWNERS' DATA REPORT FOR EDDY CURRENT EXAMINATION RESULTS As required by the provisions of the ASME Code Rules EDDY CURRENT EXANINATION RESULTS PLANT: Turkey Point Unit 3 EXAMINATION DATE: October 5, Igg 8 through October 8,1998 Tubes Steam Total Total Tubes Plugged as Tubes Plugged Total Plugged Generator Tubes Preventative This Outage Tubes in S/G Inspeded 20%-39% >40% Maintenance 3E210A 3194 6 go 0 0 0 20 3E2108 3187 11 to 0 1 ra 1 28 3E210C 3179 27 go 0 0 0 35 LOCATON OF INDICATIONS pi (20 % 100%)
Support Support Top of Top of Total Total ;
Steam Freespan AVB Locations Locations Tubesheet Tubesheet ind^ cations indications Generaw Bars ,
1 thru 6 1 thru 6 06H.08C to #t TSP to #1 TSP 20%-39% 240 %
Cold Leg Hot Leg U-Bend Cold Leg Hot Leg 3E210A 11 to 0 0 0 0 0 11 to 0 3E2108 22 tom 0 0 0 0 0 22 mm 0 3E210C 43 :n 0 0 0 0 0 43 cy 0 Remarks:
(1) Mechanical weer damage at antWibration bars (AVB) was depth sized using qualified bo'ubin coR sizing technique.
(2) One tube was preventatively plugged in SG 38 for AVB wear progression.
(3) Some tubes may have more than one AVB wear indication reported.
Date: /8/8//8 Prepared by: N
- CSI S/G Eddy Current doordinator Date: 10/2.ta /9A Reviewed by- e 1I inspections Superdsor i Date: 10 /11r /f f Reviewed by: 2 I An4/ I 8 w CSI S/G(echnical Sf fecialist 84
[ CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT / 3 09/98 COMPONENT : S/G A Page : 1 Date :
10/13/98 Examination Dates : 10/05/98 thru 10/08/98
-Total Number of Tubes Inspected .....: 3194 Total Indications Between 2 0% and 3 9 % . . . . . . . . . . . . : 11 Greater than or equal to 40% ...: 0 Total Corrosion type Indications "VOL" 0 Total Tubes Plugged as Preventive Maint : 0 Total Tubes Plugged ....................: 0 Location Of Indications 20% to 100% & "VOL"
-Hot Leg Cold Leg ;
TSH .5 to 01H -2.1 : 0 TSC .S to Olc -2.1 : 0 O1H -2.0 to 06H +2.0 : 0 Olc -2.0 to 06C +2.0 : 0 06H +2.1 tc AV1 -3.1 : 0 06C +2.1 to AV4 -3.1 : 0 AV1 -3.0 to AV4 -3.0 : 11 85
IN0lCATIONS/ TRENDING REPORT PTN 3 OUTAGE : 09/98 COMPONENT S/0 A p.g, , j
' 9E,$CRIPfl0N : ALL*20% TO 39% INDICATIONS g,g,; ggf,37,,
flee : 15:55:32 l l -fatent l l 09/98 l l N/A g l8oul Col l Log l***lTst/ Nets l Reel l' Probe l Location l Volts lDeglCh l X l0lf fl Location l Volts l Des lChl3l
.......l...l...l........l........l............l.............l.....l...l...l...l....l..............j.....l...l...l.,,,
j 33l 41l C l 'lTENTECPSlACO25 lA-720 M/ULC lAV1 .0 l .9l lP 2l 26l l l l l l g l l lCl lTENTECPSlAC025- lA720M/ULClAV2 .0 l .4l -lP 2l 23l l l l l l g l -l lCl . liENTECPSlACO25 lA.720M/ULClAV3 .0 l 1.4l lP2l29l l l l l l l l.l lCl lTENTECLtlACO25 lA-720-M/ULClAV4 .0 l .4l lP2l23l l l l l l g l 31l 44l N l lTECTEN lAN025 lA-720M/ULClAV3 .0 l .4l lP 2l 29] l l l l l l l 34l 45l C l lfENTECPSlAC024 l4-720-M/ULClAV2 .0 l 1.8l lP2l35l l l l l l g l l lCl .lTENTECPSlAC024 lA720M/ULClAV3 .0 l .8l lP2l26l l l l l l g l 37.l 47l C l lTENTEC lAC027 lA.720-M/ULClAV3 .0 l 1.0l lP2l28l l l l l l g l 30l 52l C l jfENTECPClAC030 lA720-M/ULClAV3 .0 l .4l lP2j29l l l l l l l l 28l 59l C l jfENTEC lAC034 lA.720M/ULClAV2 .0 l .7l lP2l25l l l l l l l l l' lCj lTENTEC lACO34 lA.720-M/ULClAV3 .0 l .5l lP2l23l l l l l l l
.. . . l . . . l . . . l . . . l . . . . . . . . l . . . . . . . . . . . . . . . . . . . . l . . . . . . . . . . . . . l . . . . . l . . . l . . . l . . . l . . . . . . . . . . . . . . . . . l . . . . . l . . . l . . .
Nuuher of .4ECOR05 Selected from Current Outage a 11 N.auber of TUSES Selected from Current Outage : 6 86
F
)
I i
.,m CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT /.3 09/98 l
COMPONENT :-S/G B Page : 1 ;
Date :
10/13/98 !
Examination Dates : 10/05/98 thru 10/08/98 Total Number of. Tubes Inspected .....: 3187 Total Indications i Between 20% and 39% ............:
~
22 {
Creater than or equal to 40% ...: 0 1 Total Corrosion type Indications "17L" 0 j Total Tubes Plugged as Preventive Maint : 1 Total Tubes Plugged ....................: O l 1
l Location of Indications 20% to 100% & b'VOL" Hot Leg Cold Leg TSH .5 to ' 01H -2.1 : 0 TSC .5 to Olc -2.1 : 0 01H -2.0 to 06H +2.0 0 Olc -2.0 to 06C +2.0 : 0 06H ' +2.1 to AV1 -3.1 : 0 06C +2.1 to AV4 -3.1 : 0 AV1 -3.0 to AV4 -3.0 : 22 1
i 87 i
i l
i l
INDICAil0NS/TEENDINC REPORT PfN-3 CUTAGE : 09/94 CtercutNT 3 5/C 3 l' Page : 1 DESCRIPil0N : ALL 60% TO 1905, VOL & PTP INDICATIONS Date : 10/13/93 Tlee
- 15:s8:
e................................................................................................................21,,, i l l Estent l l. 09/98 l l N/A g ,
l Row l Col lLosl
- lfst/Netel aeel l Probe l Location l Volts l pes lCh l X l0iffl Location lveltsl Des lCh 3 1
.... l . . . l .. . l . .. j . . . . . . .. l .. .. . .. . l . . . . . ... ... . l . . .. . ... . . . . . l .. . .. l . .. l . . . g . .. l . . . . ;l l 4 l ssl n lminciaN lsmou .newute kvi .0 l 1.sl le:lrftl l l l l 1
- e. . . l . . . l . . . l . . . l . . . . . . . . l . . . . . . . . l . . . .. . . . .. . . l . . . . . . . . . . . . . l . . . . . l . . . l . . . li ph of RfC040$ selected from Current Outage : 1 '
kuuber of TUBES Selected fras Current Gutege : 1 I
)
88 o
. j l
INotcAf ton /TRENofWe REPoti i PTN.3 ouTAGr : 09/98 conPontNT : s/c 5 Pete s i 0(scRIPfloN : ALL roe To 395 INDicAitous oste : 10/13/98 Time 8 is:so:si I l ratont l l o9/98 l l N/A j p lceipe,l~.ps:/ mot:1 ant l cree l Leestien Ivoitsloesien l x loisel tocation Iveti.loesten l lg
....l...l...l...l........l........l............l.............l.....l...l...l...l....l.............l.....l...l...l.,,, J l zal ul a l Inenuta puoti p no nfute pv4 .o l .si p zl 2:1 l l l l l l lz9luinl lncuntspnen p.no-nfute pvi .o l .4l p zl ul l l l l l l26lzalnI psenn paars p no-n/ute pv4 .o l .sl y zl zil l l l l l g l54lsilnl guerieslents p.no n/ute pva .o l .sl lezlzil l l l l l l l l lul prennspunis p.no n/ute pvs .o l .sl y zl 251 l l l l l l l nj 34l a l prennspnozi p.no n/utt kvi .o l .sl y 21 sol l l al l l l l lml lreeuwslenozi .no.n/ute pn .o l .si p al ari l l 3 l l l l l lul pecnespnori p.no.n/ute pvs .o l .9l lrzlszl l l l l l l l l lul pscunspuozi .no-nfute pv4 .o l .si p 21 zrl l l l l l l j l 34l 46l u l -lrrenesienot, p.72o-neuteIAvz .o l i.nl lp l sol l l l l l l ;
l l lnl lncTrwslenoi, p.72o n/ute pvs .o l 1.4l le21371 l l l l l l !
lssl6slnl pecunspacia p.no-n/ute kvr .o l .4l pzlrsi l l l l l l l l l lwl lnennspuoi6 p.no.n/ute pvs .o l .sl p zl 271 l l l l l l liilsolel lusucschcon p.no neute pvi .o l .il y l zsi l l l l l l l l341silul lncuesjonai6 p.no.n/ute kvi .o l .zl p 2 zil l l l l l l l l 1 lul lnennspuois p.no n/ute pn .o l i.il p 21 34l l l l l l l 1 l lul lnenesionia p.no n/ute pvs .o l .sl y l ul l l l l l l l34lsainl lnens pants p.no-n/ute kv .o l .si p2126l l l l l l l l l lNl Incun pai6 p.no-n/ute un .o l .6l p l zal l l l l l l ,
I l lul prenn punis p-no-neute pvs .o l .sl y zl ul l l l-l l l l42lssinI prens Isais p.no-n/ute pn .o l .zl p 21 ni l l l l l l l l lul lucun puois lA.no-n/ute kys .o l i.sl lr al nl l l l l l l
....l...j...l...l........l........l............l.............l.....g...l...l...l....l.............l.....l...l...l....
Nudwr of RECoRot selected from current Outage 22 Nudwr of TuSEs betected from current Outage si e
89
u.
CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT # 3 09/98 COMPONENT : S/G C Page : 1 Date : 10/1379g
. Examination Dates :- 10/05/98. thru 10/08/98 Total Number of Tubes Inspected . . . . . : 3179 Total Indications Between 2 0% and 3 9 % . . . . . . . . . . . . : 43 Greater'than or equal to 40% ...: 0 Total Corrosion type Indications avoL" 0 Total Tubes Plugged as Preventive Maint : 0
. Total Tubes Plugged ....................: O Location of Indications 20% to 100% & "VOL" Hot Leg cold Leg TSH .5 to 01H -2.1 : 0 TSC .5 to Olc -2.1 : 0 01H -2.0 to 06H +2.0 : 0 Olc -2.0 to 06C +2.0 : 0 06H +2.1 to AV1 -3.1 : 0 06C +2.1 to AV4 -3.1 : 0 AV1'-3.0 to AV4 -3.0 : 43 l
i i
90
thoicAffoNs/ TRENDING stroRT PfN 3 DJTAGE 09/9s conr0 MENT s/G e page i DEscairTIoM : ALL zot To 39: INoicAf ton: cate to/13/9s alme : 16:o4:32 e.............................................................................................................. ,,,,,
l l Estent l l o9/9s l l n/A g lRowjcell Leg l***lfst/ Note l seel l Probe l Location lvaltsjoeglch l X lalffl tocation lvoltsjoeslch l 1 e . . . l . . . g . . . l . . . g .. . . . . . . l . . . . . . . . l . . . . . . . . . . . . l . . . . . . . . . . . . . l . . . . . l . . . g . . . l . . . l .
l sil ul c l' penscrsiccess p.non/utepvt .o l .zl irzlrol l l l l l g l sol tri c le prnecesiccess .no.nme pvi .o 'l .il yzlzij l l l l l g l sol isl c l penecesiccess p.no ame kvi .o l .zl irzlrs l l l 1 l l l ul sil e l penecesiccan .no-ame pvs .o 1 4l lealzsl l l l l g l 341 sil c l- lTenecesiccosr .no-n/ute pn .o l .sl p 21 ni l l l l l g l 1 l c V penececiccosr .no.n/ute pvs .o l .71 p 21 al l l l 1 l l l 421 sil e p prerewslcuan p.no n/ute kys .o l .4l p zl zal l l l l l l
. I ni ni c le pmeceslecess p.no-n/ute n .o l .sl y zl ni l l l 1 l l l 1 l c p p m scesiccess no-n/ute pvs .o 1 .31 p zl ni l l l l g l l 431 ni e 1/ pAnecac ccess p.no-n/ute pvs ..: l .sl p 21 ni j l l l l l l ssi ni c l/ peneceslccess i k.no.n/ute pvs .o l .sl lezlzil l l l l l l l 4sl ssi c I/ pmsceslccess .no-n/ute pn .o l .zl lr zl zil l l l l l l l ssi ul c 1/ p msceniccosa p.no-n/ute pva .o l .4l p al ni l l l l l l l l lcI/ pensceniccou p.no-ame pvs .o I .sl lezl241 l l l l 1 l l 34l 4 l c l< pmscesiccosr .no-ame kvi .o l .sl lrzl sol l l l l l l 1 I I e y pmecesiccasi k.no.nmc pn .o l .sl y zl al l I i l l l l l l c p pmecesiccost p.no-n/ute pn .o l .61 lezlzal l l l l l l l 1 l c P p m eer:Iccu r p.no-nac kys .o l .71 y zl z91 l l l l l l l ssl 431 e l pmscesiccut p.no-n/ute pn .o l .9) lezlrsi l l l l l l l l l c p inneceslecui p.no-n a c pn .o l i.zl irzl291 l l l l l l l l l c P p m sceslecui p.no-n/ute pv4 .o l .si. p 21 24l l l l l l l l 34l 44l c y p m scesiccuo p.no.nmc pvs .o l .41 p zl nl l l l l l l l351441 e P p m ecesiccui p no-nfute n .o l i.sl y zl ni l l l l l l l l l c F p m sceslecai k.no.nme pn .o l 1.61 p al 34l l l 1 l l l l l lc y penscesiccui p.no-nfute pv4 .o l .6l p zl zil l I l l l l l zal 4al e p tr m eersicco46 p-no-n/ute pn .o l .sl y zl nl l l l l l l l 35l 49l c l# liEnEceslcco43 p-72o-H/utt pv4 .o l- .sj lezl22l l l l l l l l 39l 541 e l/ pencersiccus p-no-n/ute pn .o l .sl lezlzal l I l l l l l 26l sal e I/ pmecesicceso p.no-n/ute kvi .o l .4l p zi 26l l l l l l l l l lcI/ pmscesleceso p no-nfute pn .o l .71 y al sol l l l l l l l l l c I/ prmerslcceso p.no-n/ute pvs .o l .sl p 21 zsl l l l l l l 1 z'l 591 e y p m scsclccoso p.no n/ute pvi .o l 4l p 21 zil l l l l l 1 l rol sol c y pr m erslccosa p.no.nfute pvz .o l .zl Irzlrsl l l l l 1 l l sol eil e le pmece:Iccoso p.no-nac pn .o l .61 lr zl 29l l l l l l l l sal 6tt c F prmersiccus p no-nac pn .o l .31 p 21 25l l 1 l l l l 1251621 e 1/ pentecesicco49 14.rzo n/ute pvz .o l .31 lpalzil l l l l l l l l l c 1/ pcmcesicco:9 p.rzo-n/ute pvs .o l .sl lpalzal l l l l l l l 24163l c p pemersicceso p.no n/ute pn .o l .4l Ir21rsi l l l l 1 l l l l c p pe m ersiccoso p.no n/ute pvs .o l .6l p al zal l l l l l l l ssl 4si n l/ pectrwslemon p.no.nme pn .o I .si p 21 zzi l l l [ l l l 1 -l n l/ pecuwslcnon -no ame pv3 .o l 4l p al zog l l l l l 1 l l l u 1/ preneslcuoz6 p.no-n/ute kv4 .o l .6l p al ul l l l l l l
-l sal ril u l/ pecirnesicNoz6 14.rzo-n/utelavs .o l .6l lpzlrsi l l l l l l
....l...l...l...l........l........l............g.............l.....l......g...l....l.............l.....l...l..l....
91
I:
lNDICAfl0K8/Tafslus REPost PTN-3 CUTACE : 09/98 COMPONENT S/S 'C peg,8 2 DESCalPfl0N : ALA 205 70 395 lNotCAflous Date a soft 3j,,
Time
- 16;04th e.......................................................................................................... ,,,,,,,,
.l l Estent.l l 99/98 l l N/A l Row l Col lLegj'**lTet/ Nets l Reet - l Probe l Lacetion lValtslDe8lCh l 1 jolffl Location lValtsl0eslChjgl
. . . . l . . . l . . . j . . . l . . . . . . . . l . ... . . . . l . . . .. . . . . . . 1. . . . . . . . . . . . . l . . . . . l . . . l . . . l . . . l . . . 1. . . . . . . . . . . . . l . . . . .
- e. .. ; .. . l .. . l . . . l . .. . . . . . g . . . . . . . . l .... . . .. . . .. l . . . . . . .. . .. .. l . . . . 1....g...,,,, . . l . .. l . . . l . . . .
thaber of RECORDS Belected from Current Gutepe : 43 Ihadser of FUSES Selected from Current Gutete : 27 l
92