ML17354A885
| ML17354A885 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/13/1997 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17354A884 | List: |
| References | |
| NUDOCS 9804160442 | |
| Download: ML17354A885 (337) | |
Text
TUEQCEY POINT PLANT UNITS 3 AND 4 DOCKET NUMBERS 50-250 AND 50-251 CHANGES, TESTS AND EXPERIMENTS MADE AS ALLOTTED BY 10 CFR 50.59 FOR THE PERIOD COVERING APRIL 8, 1996 THROUGH OCTOBER 13 1997
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ZNTRODUCTXON, This report is submitted in accordance .with 10 CFR 50.59(b),
which requires that:
i) changes in the facility as described in the SAR ii) changes in procedures as described in the SAR, and iii) tests and experiments not described in the SAR, which are conducted without prior Commission approval, be reported to the Commission for the same period as required by 50.71(e) for the Turkey Point UFSAR update. This report is intended to meet this requirement for the period covering April 8, 1996, through October 13, 1997.
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 section summarizes 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 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 Xtem 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. Both Units 3 and 4 had a steam generator tube inspection during this reporting period.
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TABLE OP CONTENTS SECTION 1 PLANT CHANGE MODZP1CATXONS PAGE 94-141 BORIC ACID EVAPORATORS AND GAS STRIPPER 12 ABANDONMENT 06/06/96 95 -'063 SJAE SPING MONITOR DRYER ENHANCEMENTS 13 12/02/96 95-072 IN-PLACE ABANDONMENT OF LIQUID WASTE 14 DISPOSAL SYSTEM WASTE EVAPORATOR PACKAGE AND SOLID WASTE D1SPOSAL SYSTEM HOLDUP AND MIXING 04/17/97 95-080 REMOVAL OF LEAD/LAG MODULES FROM LOW PRESSURIZER PRESSURE TRIP INSTRUMENTATION 09/18/96 95-097 REPLACEMENT OF MAIN STEAM SAFETY RELIEF VALVE DISCHARGE PIPING 04/10/96 l 95-147 EMERGENCY CONTAINMENT COOLER START LOGIC 17 DESIGN CHANGE 10/03/96 95-148 EMERGENCY CONTAINMENT COOLER START LOGIC 18 DESIGN CHANGE 10/21/96 95-157 REMOVE TIME DELAY FOR BLOWDOWN ISOLATION 19 VALVES CV-6275A,B,C 07/24/96 95-168 REMOVE TIME DELAY FOR BLOWDOWN ISOLATION 20 VALVES CV 6275AI B I C 06/11/96 95-170 THERMAL POWER UPRATE SETPOZNT/SCALING 21 01/14/97 95-171 THERMAL POWER UPRATE SETPOINT/SCALING 22 01/24/97 96-008 POTENTIAL EDG LOCKOUT FOLLOWING NORMAL STOP 23 03/25/97 96-012 UNIT 3 BORON INJECTION TANK BYPASS 24 MODIFICATION 03/24/97
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SECTION 1 PLANT CHANGE MODIFICATIONS (Continued) PAGE 96-022 THERMAL POWER UPRATE IMPLEMENTATION 25 01/29/97 96-036 SPRING/SETPOZNT CHANGE FOR THE PILOT OPERATED 26 LOCKUP VALVES FOR THE ECC CCW SUPPLY/RETURN ISOLATION VALVES 05/22/96 96-039 SPRING/SETPOINT CHANGE FOR THE PILOT OPERATED 27 LOCKUP VALVES FOR THE 'ECC CCW SUPPLY/RETURN ISOLATION VALVES 05/22/96 96-040 APPENDIX R DOCUMENTATION CHANGE IN SUPPORT 28 OF CRs96-023, 96-474,96-754, 96-951, 96-1060; Ec 96-1152 02/13/97 96-076 ENHANCEMENT OF RCP BUS UNDERVOLTAGE TIME 29 DELAY RELAY CIRCUIT 03/17/97 96-084 RADIANT ENERGY SHIELDS INSIDE CONTAINMENT 30 03/28/97 96-085 RADIANT ENERGY SHIELDS 'INSIDE CONTAINMENT 31 10/02/97 96-093 UNIT 4 ADDITION OF CCW HEAD TANK 32 10/08/97 97-012 THERMAL OVERPRESSURIZATION OF ISOLATED PIPING 33 10/10/97 97-026 MOV-4-744A AND MOV-4-744B REPACKZNG AND 34 EQUALIZING LINE INSTALLATION 10/06/97 97-035 MONITOR TANK "B" DIAPHRAGM ELIMINATZON 35 08/27/97 97-041 CORE EXIT THERMOCOUPLE COLUMN REMOVAL - UNIT 4 36 10/11/97 SECTION 2 SAFETY EVALUATIONS SEEJ-88-042 DE-ENERGIZATION OF UNIT 4 4160 VOLT SAFETY 38 RELATED BUSSES 09/05/97
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SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SEEJ-89-085 DE-ENERGIZATION OF UNIT 3 4160 VOLT SAFETY 39 RELATED BUSSES 02/20/97. 03/09/97; 03/13/97 SEMS-90-041 10. CFR 50.59 SAFETY EVALUATION - ACCEPTAB1LITY 40 OF AS-FOUND CONDITION FOR RHR CHECK VALVE 3-753A 03/26/97 SEMS-93-010 SAFETY EVALUATION FOR ICW VALVE REPLACEMENT AND B HEADER CRAWL THROUGH INSPECTION 09/04/97 SECS-93-013 SAFETY EVALUATION FOR SCAFFOLDING AND COVER FOR "42 SPENT FUEL POOL ROOMS 09/04/97 SENP-95-007 SAFETY EVALUATION FOR OPERABILITY OF RHR DURING 43 INTEGRATED SAFEGUARDS TESTING 09/10/96 SENP-95-023 SAFETY EVALUATION FOR OPERABILITY OF RHR DURING 44 INTEGRATED SAFEGUARDS TESTING 10/01/97; 10/01/97 SECP-95-046 SAFETY EVALUATION FOR UNIT 3 AND UNIT 4 TWENTY- 45 FIFTH YEAR CONTAINMENT TENDON SURVEILLANCES 09/11/96 SENS-95-052 SAFETY EVALUATION FOR REVISION OF UNIT 3 DIESEL 46 FUEL OIL TRANSFER SYSTEM TECHNICAL SPECIFICATION BASES 03/03/97; 03/13/97 SEMP-96-004 EVALUATION OF AP ARNAFLEX INSULATION FOR 47 FOR CONDENSATION DURING TEMPORARY CONTAINMENT COOLING 02/27/97 SEMS-96-014 SAFETY EVALUATION FOR A TEST OF THE USE OF SUB- 48 MICRON ULTRAFINE FILTERS IN THE CVCS AND SFP SYSTEMS 02/06/97 SENP-96-017 SAFETY EVALUATION FOR UFSAR CONSISTENCY REVIEW 49 AND UPDATE 05/09/96 SENS-96-024 SAFETY EVALUATION TO SUPPORT A CROSS-CONNECTED 50 RHR SYSTEM REAL'IGNMENT DURING A REFUELING OUTAGE 03/21/97
0 SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SENP-'96-025 SAFETY EVALUATION FOR REMOVAL OF THE RCC CHANGE FIXTURE FROM THE FSAR 06/04/96 SENS' 028 ,SAFETY EVALUATION UFSAR OPERATIONAL REVIEW 52 AND UPDATE 09/19/96 SECS-96-033 SAFETY EVALUATION FOR INSTALLATION OF REMOTE 53 CAMERA FOR 'B'CP OIL LEVEL VERIFICATION 08/29/96; 04/25/97 SEMS-96-037 SAFETY EVALUATION FOR THE TEMPORARY INSTALLATION'4 OF A FREEZE SEAL AND BLIND FLANGE PER TSA-04-96-046-3.1 TO SUPPORT MOV-4-350 VALVE REPAIRS 05/31/96 SEMS-96-038 SAFETY EVALUATION FOR UNIT 3 STEAM 55 SIDE 'FOREIGN OBJECTS GENERATORS'ECONDARY 08/22/96; 03/31/97 SEMS-96-040 SAFETY EVALUATION FOR THE TEMPORARY INSTALLATION 56 OF DRAIN HOSES AND PERFORMANCE OF HOT SPOT FLUSHES ON THE RHR SYSTEM PER TP-96-046 AND TP-96-047 07/25/96 SEIS-96-042 SAFETY EVALUATION FOR THE OPERATION OF RAD-*-6458 57 THE REACTOR VESSEL HEAD, LEAK DETECTION SYSTEM ON A NON-CONTINUOUS BASIS 08/23/96 SEIS-96-044 SAFETY EVALUATION FOR ABANDONING SPENT FUEL 58 PIT BRIDGE'CRANE LASER RANGE FINDER 08/01/96 SEES 046 SAFETY EVALUATION FOR ESSENTIAL EQUIPMENT 59 AFFECTED BY A HEAVY LOAD DROP RESULTING FROM EPS MODIFICATIONS 08/26/96 SEMS-96-048 SAFETY EVALUATION FOR MEASURING ICW HEADER 60 FLOW PER TP-96-070 08/27/96 SECS-96-059 EVALUATION OF TEMPORARY SCAFFOLDING WITHIN CONTAINMENT FOR USE IN REPLACING RELIEF VALVE RV-3-203 09/25/96 SENS-96-062 SAFETY EVALUATION TO PERMIT WIDER PRT LEVEL 62 OPERATING BAND 09/27/96
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SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SECS-96-065 10 CFR 50.59 SAFETY EVALUAT10N FOR UNIT 3 63 CYCLE 16 REFUELING OUTAGE TURBINE OVERHAUL SUPPORT 12/17/96; 01/30/97 SENS-96-068 SAFETY EVALUATION RELATED TO REPQSIT1ONING OF VALVE HV-7 ON THE POST ACCIDENT CONTAINMENT VENTILATION SYSTEM 12/05/96 SEMS-96-078 SAFETY EVALUATION RELATED TQ TSA No. 03-96-20-09'5 FOR REPAIR OF MOV-3-832 12/09/96 SEES-96-081 SAFETY EVALUATION FOR TEMPORARY SYSTEM ALTER- 66 ATION (TSA) NO. 4-96-003-16 FOR CONTAINMENT ZNSTRUMENT AIR BLEED VALVE CV-4-2826 CIRCUIT MODIFICATION 12/13/96 SEMS-97-006 SAFETY EVALUATION TO ABANDON AUXILIARY STEAM IN 67 THE AUXILIARYAND RADWASTE BUILDINGS 02/11/97 SEIS-97-013 SAFETY EVALUATION FOR THE ABANDONMENT, REPAIR 68 QR RESTORATION OF RVLMS SENSORS 02/27/97 SEMS-97-014 SAFETY EVALUATION TO ADDRESS THE OPERATION OF 69 THE CONDENSATE POLISHING DEMINERALIZER SYSTEM 05/29/97 SEMS-97-016 SAFETY EVALUATION FOR INSTALLATION OF AN OIL 70 DRAIN LINE ON THE 3B RCP MOTOR 03/18/97 SEIS-97-017 SAFETY EVALUATION RELATED TO THE ABANDONMENT OF MONITORING FUNCTION OF PARTICULATE DETECTOR CHANNELS OF THE SPING EFFLUENT MONITORS 05/20/97 SENS-97-023 SAFETY 'EVALUATION RELATED TO SEVERE ACCIDENT 72 MANAGEMENT GUIDELINE IMPLEMENTATION 06/13/97 SENS-97-027 SAFETY EVALUATION RELATED TO TEMPORARY ALTER- 73 ATION OF THE "C" GAS DECAY TANK SAMPLE PATH 02/28/97 SEFJ-97-030 10 CFR 50.59 SAFETY EVALUATION - EVALUATION FOR 74 PERFORMING THE ROD DROP TEST FROM ALL RODS OUT CONDITION 09/02/97
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SECTION 2 SAFETY EVALUATIONS (Continued) PAGE
'SEMS-97-033 SAFETY EVALUATION FOR UNIT 3 REACTOR CORE 75 FOREIGN OBJECT 03/28/97 SENS-97-038 SAFETY EVALUATION FOR FREEZE SEALS ON SEAL 76 TABLE THIMBLE GUIDE TUBE 04/07/97 SENS-97-040 SAFETY EVALUATION RELATED TO STEAM GENERATOR "77 BLOWDOWN CONTAINMENT ISOLATION FEATURES 04/29/97 SEIS-97-041 SAFETY EVALUATION FOR'ORE EXIT THERMOCOUPLE 78 ABANDONMENT OR REPLACEMENT 06/27/97 SECS-97-051 10 CFR 50.59 EVALUATION FOR THE PLACEMENT 79 OF LEAD SHIELDING ON VALVE LCV,-4-460, AND SHIELD WALL BETWEEN VALVE LCV-4-460 AND REGEN HEAT EXCHANGER 04/25/97 SENS-97-054 SAFETY EVALUATION RELATED TO TEMPORARY REMOVAL 80 OF THE 3A CCW HEAT EXCHANGER CHANNEL HEADS AND TEMPORARY SUPPORT INSTALL'ATION 05/09/97 SENS-97-058 SAFETY EVALUATION FOR REPEALING THE FIRE 81 BARRIER REQUIREMENTS ASSIGNED TO PENETRATION No. 041E-E001 IN THE EAST WALL OF THE BORIC ACID STORAGE TANK ROOM (FIRE AREA M/41) 07/24/97 SEIS-97-059 SAFETY EVALUATION FOR TEMPORARY INSTALLATION 82 OF A RECORDER ON RCS FLOW LOOP F-4-426 (TSA 04-97-049-3) 07/07/97 SECS-97-061 SAFETY EVALUATION FOR CONTAINMENT POLAR CRANE 83 MAINTENANCE INSPECTION PROCEDURE 09/04/97
. SEES-97-062 SAFETY EVALUATION FOR SURVEILLANCE REQUIREMENTS FOR APPENDIX R CIRCUITS AT 4.16 KV SWITCHGEAR AND 480 V LOADCENTER BUSSES 08/06/97 SEMS-97-065 SAFETY EVALUATION TO DOCUMENT THE CHANGE FOR 85 AUXILIARY FEEDWATER SYSTEM STEAM TRAPS ST-33, ST-34, AND ST-35 DRAINING INTO THE ADJACENT TROUGH IN LIEU OF THE EXISTING CONDENSER DRAIN-PATH 07/29/97
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SECTION 2 SAFETY EVALUATIONS (Continued) PAGE SENS-'97-066 SAFETY EVALUATION FOR AUGMENTED RCP OZL 86 COLLECTION SYSTEM REQUIREMENTS 08/21/97 SENS-97-076 SAFETY EVALUATION FOR TEMPORARY ALTERATION 87 OF CONTAINMENT PURGE VALVE POV-4-2600 CONTROL AND INDICATION CIRCUIT WIRING 09/08/97 SECS-97-077 EVALUATION FOR STORAGE OF TOOLS AND EQUIPMENT 88 IN CONTAINMENT DURING ALL MODES OF OPERATION 10/02/97; 10/10/97 SEMS-97-078 SAFETY EVALUATION FOR CONTINUOUS FERE WATCH 89 10/14/97 SENS-97-084 SAFETY EVALUATION FOR CONTAINMENT SUMP SCREEN 90 DESIGN REQUIREMENTS 10/02/97 SECL-97-200 REACTOR COOLANT SYSTEM LOOSE PARTS EVALUATION 91 09/25/97
'SECTION 3 RELOAD SAFETY EVALUATIONS96-024 TURKEY POINT UNIT 3 CYCLE 15 RELOAD DESIGN 93 (THERMAL POWER UPRATE) 11/13/96 96-025 TURKEY POINT UNIT 4 CYCLE 16 RELOAD DESIGN 94 (THERMAL POWER UPRATE) 11/13/96 96-071 TURKEY POINT UNIT 3 CYCLE 16 RELOAD DESIGN 95 05/07/97 97-014 TURKEY POINT UNIT 4 CYCLE 17 RELOAD DESIGN 96 10/22/97 SECTION 4 REPORT OF POWER OPERATED RELIEF VALVE PORV ACTUATIONS UNIT 3 '98 UNIT 4 98
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SECTION 5 STEAM GENERATOR TUBE INSPECTIONS FOR TURKEY PO1NT PAGE'NIT 4 100-105 UNIT 3 106-115 10
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SECT10N 1 PLANT CHANGE / MODXPXCATIONS 11
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PLANT CHANGE MODIPICATION 94-141 UNIT 3 & 4 TURN OVER DATE : .06/06/96 BORIC AC1D EVAPORATORS AND GAS STRIPPER ABANDONMEKT.
~SllEER This Engineering Package provided the design change documentation required for abandonment of the boric acid evaporators and gas stripper equipment. The original purpose of this equipment was to process reactor coolant effluent in order to conserve boric acid and reduce the processing 'requirements placed on the liquid radioactive waste. disposal system. The boric acid evaporators and gas strippers were 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 ion exchange, gas stripping, and evaporation process was separation and reclamation of boric acid and primary water. By the mid-1980s, this method of liquid waste disposal had proven too costly to operate and maintain.
Presently, all radioactive liquids requiring cleanup are 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 and mechanical equipment associated with the evaporators and gas strippers were abandoned in-place. Power feeds were isolated by lifting leads and de-energizing breakers. The various system isolation valves, including the component cooling water valves to the boric acid evaporators, were previously isolated. A circular plate was installed in a flange upstream of valves .CV-*-6598 and CV-*-6599 to prevent potential back leakage from the waste disposal vent header from entering the gas stripper package.
Safet Evaluation:
The boric acid evaporators and gas strippers did not serve any safety related functions and were not required to support safe shutdown of the plant ~ The in-place abandonment of this equipment had no adverse impact on plant safety or plant operations. As demonstrated 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 these modifications.
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PLANT CHANGE MODIPICATION 95-063 UNIT 4 TURN OVER DATE 12/02/96 BJ'AE SPING MONITOR DRYER ENHANCEMENTS
~Summa
. This Engineering Package replaced the chiller unit used to condense and remove moisture from the steam jet air ejector SPING sample stream because it was corroded beyond repair. The. steam jet 'air ejector SpING monitors the condenser exhaust gases for particulate, iodine,'and noble gas radioactivity as required by Regulatory Guide
. 1.97. The replacement system utilized a moisture separator tank to, remove entrained liquid from the sample stream, and heat tracing to raise the temperature of the sample prior to entering the SPING unit to prevent condensation. The moisture separator tank was installed on the mezzanine. deck of the Turbine Building, next to the SPING unit.
The heat tracing and thermal insulation was designed to preclude condensation and maintain process temperatures below the maximum value .specified by the SPING vendor. The .heat tracing was not designed to meet seismic criteria because any credible failure of the heat tracing would not damage any safety related equipment.
Electrical power for'he heat tracing circuit was taken from the non-vital side of MCC 4A which has no impact on station battery loads, emergency diesel generator loads, or emergency diesel generator load sequencing.
Safet Evaluation:
The steam jet air ejector SPING monitor does not perform any nuclear safety related functions. However, to preclude the possibility of any adverse interactions with safety related structures, systems and components, the mounting and relocation of drain and sample line tubing, conduit, and associated structural supports, were evaluated in accordance with .seismic criteria contained in the UFSAR. Tubing and supports located above the operating deck of the Turbine Building were also evaluated for applicable hurricane wind loads in accordance with UFSAR criteria.
Based on the evaluation criteria addressed in this PC/M, 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 these modif ications.
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PLANT CHANGE MODIFICATION 95-072 UNIT 3 Ec 4 TURN OVER DATE 04/17/97 IN- PLACE ABANDONMENT'P LI UID WASTE DISPOSAL SYSTEM WASTE EVAPORATOR PACKAGE AND SOLID WASTE DISPOSAL SYSTEM HOLDUP AND MIXING
~Summa This Engineering Package provided the design change documentation required for abandonment of the liquid waste evaporator package and the solid waste disposal system holdup and mixing equipment.
The abandoned liquid waste evaporator package included an evaporator, absorption tower, evaporator condenser, vent condenser, distillate cooler, distillate demineralizer, distillate f ters, distillate pump, feed preheater assembly, il stripping column assembly, and concentrate pump. The abandoned solid waste disposal system holdup and mixing equipment included the radwaste holdup/mixing tanks and their associated pumps. The original purpose of this equipment was to concentrate dissolved and suspended . solids from liquid waste using an 'evaporation, condensation, and filtering process. Over the years, this method of liquid waste disposal has proven too costly to operate and maintain. Presently, all radioactive liquids requiring cleanup are processed by the waste disposal portable demineralizer skid located in the Radwaste Building, and released to the circulating water system upon verification that the discharge meets site release requirements. The only solid wastes currently requiring management and disposal are the spent resins. These are currently transferred to shipping containers, dewatered, and sent off-site for final processing and disposal.
All electrical and mechanical equipment associated with the above liquid and solid waste processing equipment are abandoned in-place.
Power feeds were isolated by lifting leads and de-energizing breakers. The various system isolation valves were isolated.
Safet Evaluation:
The liquid waste evaporator package and the solid waste disposal system holdup and mixing tanks and pumps did not serve any safety related functions and were not required to support safe shutdown of the plant. The in-place abandonment of this equipment had no adverse impact on plant. safety or plant operations. As demonstrated 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 these modifications.
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PLANT CHANGE MODIPICATION 95-080 UNIT 3 & 4 TURN OVER DATE : 09/18/96 REMOVAL OP LEAD LAG MODULES PROM LOW PRESSURIZER PRESSURE TRIP INSTRUMX2TTATION
~Summa This Engineering Package removed the lead/lag modules from the low pressure portion of the Unit 3 and 4 pressurizer pressure protection channels I, II, and III. The lead/lag units were originally installed in the pressurizer pressure instrumentation loops to provide an amplification of the pressurizer pressure signal as a function of its rate of decline. This was intended to generate an anticipatory low pressure reactor trip signal during a
, cooldown/depressurization event which otherwise did not warrant a safety injection actuation.
The function of the lead/lag units was not, credited in any of the UFSAR safety analyses. The plant safety analyses assume that the reactor trips at the set pressure value which corresponds to the low pressurizer pressure technical specification limit minus the instrument loop uncertainties. Removal of the lead/lag unit does not affect the ability of the pressurizer pressure protection channels to initiate a reactor trip at the assumed setpoint.
Removal of the lead/lag units was based in part on a Westinghouse analysis that concluded lead/lag modules do not perform their intended function. Westinghouse determined through analysis of plant operational transients that events which are severe enough to actuate the low pressurizer pressure trip will generally also actuate the low pressurizer pressure safety injection setpoint, even with lead/lag compensation.
Safet Evaluation:
Removal of the lead/lag units did not affect the ability of the protection channels to trip the reactor due to low'ressurizer pressure at the value assumed in the plant-safety analyses. In addition, the implemented changes did not affect any setpoints or operating characteristics of any safety system required to prevent or mitigate the consequences of design basis accidents. Since no new failure modes were created. by the lead/lag module removal, 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 of the protection channel changes.
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PLANT CHANGE MODIFICATION 95-097 UNIT 4 TURN OVER DATE 04/10/96 REPLACEMENT OP MAIN STEAM SAFETY RELIEF VALVE DISCHARGE PIPING
~Summa This Engineering Package replaced the ten-inch diameter discharge piping for each main steam safety valve (MSSV) on Unit 4 with twelve-inch diameter piping to support plant thermal uprate project implementation. The implementation of this design package involved.
the replacement of approximately twenty five feet of piping and modifications to two supports for each of the twelve MSSVs. Zn addition, -permanent drains were provided for the discharge piping.
The MSSVs serve to protect the secondary plant system from overpressurization. Four MSSVs are located'n each of the three main steam lines, upstream of the main steam isolation valves (MSXVs). Based on analyses'erformed for the plant uprated condition, it was determined that back pressure on the MSSVs resulting from the mass flow rate at uprate conditions coupled with the unusually long discharge lines would be excessive, and could lead to high blowdown rates. High blowdown rates could result in reactor coolant system temperatures dropping below the "no-load" temperature, potentially affecting the fatigue analysis of several reactor coolant system components. Zn order to preserve all of the reactor coolant system fatigue analysis margins, increasing the diameter of the discharge piping would result in lower backpressures and lower blowdown rates in the range of 3% to 8.,
thereby, ensuring that fat'igue analysis .margins would remain adequate to accommodate reactor coolant system transient conditions.
.Saf et Evaluation:
The modifications addressed by this Engineering Package ensure that fatigue analysis margins would remain adequate to accommodate reactor coolant, system transient conditions. 'These modifications, therefore, did not alter the design bases, functions, or, operation of the main steam saf ety relief system and did not create any adverse interactions with any other safety related structures or plant systems. Consequently, 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 these modifications.
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PLANT CHANGE MODXPXCATXON 95"3.47 UNIT 3 TURN OVER DATE : 10/03/96
, EMERGENCY CONTAXNMENT COOLER'TART LOGXC DESXGN CHANGE
~Sl1IQQlK This Engineering Package modified the start logic of the emergency containment coolers (ECCs) to support plant thermal uprate project implementation. The original design provided an automatic start.
capability for all three ECCs. However, analysis .has shown that, with the fouling factor assumptions used for thermal power uprate, the component cooling water (CCW) system thermal . analysis temperatures could be exceeded start during a limiting case event.
if all three ECCs were allowed to The ECC start logic was changed to coincide with the containment integrity reanalysis performed at .uprated conditions. The reanalysis demonstrated that containment temperature and pressure values would remain below design limits provided that, as a minimum, one ECC automatically started at the onset of the event and a second ECC started within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. of the 'event for containment temperature reduction. The modif ications implemented by this Engineering Package revised the ECC start logic such that two ECCs, the dedicated train A and B ECCs, would automatically start on a safety injection signal. The third (i.e., swing) ECC would be capable of manual operation in the event a single active failure caused one of the dedicated train A or B ECCs to fail.
Accordingly, this Engineering Package revised the swing ECC control circuit to remove the auto start capability and permit only manual .
operation.
Safet Evaluation:
The modifications addressed by this Engineering Package and the associated technical specification changes were approved by the NRC in license amendments. Accordingly, the design change to eliminate the automatic start feature of the third (swing) ECC on a safety inj ection signal had received NRC approval. The implemented changes ensure that CCW system will be capable of performing its safety related heat removal function during a limiting case event.
In addition, the capability of the ECCs to perform the required
. containment pressure mitigation, temperature reduction, and hydrogen mixing functions were not compromised by the design change. Consequently, the ECC control circuit changes did not constitute an unreviewed safety question. Since the emergency containment cooler technical specifications were approved in advance of the Engineering Package implementation, specific NRC approval "
was not required for implementation of these modifications.
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PLANT CHANGE MODIFICATION 95-148 UNIT TURN OVER DATE : 10/21/96
,EMERGENCY CONTAINMENT COOLER START LOGIC DESIGN CHANGE
~Sl1EER This Engineering Package modified the start logic of the emergency containment coolers (ECCs) to support plant thermal uprate project implementation. The original design provided an automatic start "capability for all three ECCs. However, analysis has shown that, with the fouling factor assumptions used for thermal power uprate, the component cooling water (CCW) system thermal analysis temperatures could be exceeded if, all three ECCs were allowed to start during a limiting case event.
The ECC start logic was changed to coincide with the containment integrity reanalysis performed at uprated conditions. The reanalysis demonstrated that containment temperature and pressure values, would remain below design limits provided that, as a minimum, one ECC automatically started at the onset of the event and a second ECC started within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the event for containment temperature reduction. The modif ications implemented by. this Engineering Package revised the ECC start logic such that two ECCs, the dedicated train A and B ECCs, would on a safety injection signal. The third (i.e., swing) ECC automatically'tart would be capable of manual operation in the event a single active failure caused one of the dedicated train A or B ECCs to fail.
Accordingly, this Engineering Package revised the swing ECC control circuit to remove the auto start capability and permit only manual operation.
Safet Evaluation:
The modifications addressed by this Engineering Package and the associated technical specification changes were approved by the NRC in license amendments. Accordingly, the design change to eliminate the automatic start feature of the third (swing) ECC on a .safety injection signal had received NRC approval. The implemented changes ensure that CCW system will be capable of performing its safety related heat removal function during a limiting case event.
In addition, the capability of the ECCs to perform the required containment pressure mitigation, temperature reduction, and hydrogen mixing functions were not compromised by the design change.. Consequently, the ECC control circuit changes did not constitute an unreviewed safety question. Since the emergency containment cooler technical specifications were approved in advance of the Engineering Package implementation, specif ic NRC approval was not required for implementation of these modifications.
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PLANT CHANGE MODIPICATION 95- 157 UNIT 4 TURN OVER DATE : 07/24/96 REMOVE TIME DELAY POR BLOWDOWN ISOLATION VALVES CV-6275A B C
~Summa This Engineering Package removed the .five minute time delay in the opening circuits of steam generator blowdown stop valves CV 6275A, B, and C.. The time delay was included -in the stop valve circuits to allow time for the system bypass valves to open and pressurize the downstream piping between the isolation valves and the blowdown flow control valves (FCV-4-6278A, B, and C), to .
prevent 'potential water hammers from occurring. However, due to excessive leakage past the flow control valves, the downstream section of piping could no't be pressurized by the bypass flow.
"Consequently, Operations has modified their operating procedure for the blowdown system such .that the bypass valves are not used to pressurize the system, or prevent piping water hammers. The design changes implemented by this Engineering Package, were initiated to eliminate an Operator "work-around" item.
=The control circuit of valves CV-4-6275A, B, and C was modified to eliminate the opening time delay. In addition, the bypass valve operators were removed to. prevent the valves from opening electrically. This maintains the valves in a'losed position which is the fail-safe position for operation of the auxiliary feedwater system, containment isolation, and the i;nterlock with sample isolation valves MOV-4-1425, 1426, and 1427. . Since position indication for the bypass valves was disconnected with removal of the valve actuator, the upstream manual isolation valve in the bypass loop (SGB-4-082A, B, and C) was locked'losed and designated as the containment isolation valve for the bypass portion 'of the containment penetration. The bypass valves remain as passive elements included in the system seismic boundary.
Safet Evaluation:
The modif ications perf ormed on the steam generator blowdown isolation valves by this Engineering Package maintained the existing containment penetration design bases and did not introduce
.any new 'failure modes not previously analyzed. The level of protection provided by penetration Nos. 28A, B, and C was enhanced by, the new bypass line arrangement because valves .SGB-4-082A, B, and C do not have to change .state to achieve the isolation function. Consequently, the margin of safety .associated with penetrations 28A, B, and C. has not decreased as a result of the modif ications. 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 of these, modifications.
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PLANT CHANGE MODIFICATION 5"168 UNIT 3 TURN OVER DATE : 06/11/96 REMOVE TIME DELAY FOR BLOWDOWN ISOLATION VALVES CV-6275A B C
~Summa This 'Engineering Package removed the five minute time delay in the opening circuits of steam generator blowdown stop valves CV 6275A, B, and C. The time delay was included in the stop valve circuits to allow time for the system bypass valves to open and pressurize the downstream piping between the isolation valves and the'lowdown flow, control valves (FCV-3-6278A, B, and C), to prevent potential water hammers from occurring. However, due to excessive leakage past the flow control valves, the downstream section of piping could not be pressurized by the bypass flow.
Consequently, Operations has modified their operating procedure for the blowdown system such that the bypass valves are not used to pressurize the system, or prevent piping water hammers. The design changes implemented by this Engineering Package were initiated to eliminate an Operator "work-around" item.
The control circuit of valves CV-3-6275A, B, and C was modified to eliminate the opening time delay. In addition, the bypass valve operators were removed to prevent the valves from opening electrically. This maintains the valves in a closed position which is the fail-safe position for operation of the Auxiliary Feedwater System, containment isolation, and the interlock with sample isolation valves MOV-3-1425., 1426, and 1427. Since position indication for the bypass valves was disconnected with removal of the valve actuator, the upstream manual isolation valve in the bypass loop (SGB-3-082A, B, and C) was locked closed and designated as the containment isolation valve for the bypass portion of the containment penetration. The bypass valves remain as passive elements included in the system seismic boundary.
Safet Evaluation:
The modifications performed on the steam generator blowdown isolation valves by this Engineering Package maintained the existing containment penetration design bases and did not introduce any new failure modes not previously analyzed. The level of protection provided by penetration Nos. 28A, B, and C was enhanced by the new bypass line arrangement because valves SGB-3-082A, B, and C do not have to change state to achieve the isolation function. Consequently, the margin of safety associated with penetrations 28A, B, and C has not decreased as a result of the modifications. 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 of these modifications.
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ll PLANT CHANGE MODIFICATION 95 "170 UNIT 3 TURN OVER DATE 01/14/97 THERMAL POWER UPRATE SETPOINT SCALING
~SlRIER The Engineering Package implemented the various instrumentation changes necessar'y to support the thermal power uprate project. The thermal power uprate affected many of the plant process variables including core Delta-T, decay heat load, steam flow, steam pressure, and feedwater flow. The following parameters and instruments were affected by the uprate: a) Overtemperature Delta-T and Overpower Delta-T reactor trip setpoints, b) rod insertion limit summators, c) Delta-T channel deviation alarm setpoint, d) steam flow / feed flow mismatch comparators and high steam line flow setpoint summators, e) turbine first stage pressure, f) main generator hydrogen temperature alarm setpoint, g) main generator megawatt output, h) spent fuel pool high temperature alarm, i) RCP thermal barrier CCW return high temperature alarm, and j) feedwater pump low suction alarm pressure. FPL received approval from the NRC for the setpoint/scaling changes in advance of implementation of this Engineering Package. The NRC's safety evaluation concluded that operation of the above systems at the proposed power levels was acceptable.
Safet Evaluation:
This RPS and ESFAS setpoint changes addressed by this Engineering Package (including Overtemperature Delta-TOverpower Delta-T, High Steam Line Flow and Steam Flow / Feed Flow Mismatch) protect the integrity of the fuel and fuel cladding,, and mitigate the .
consequences of design basis. accidents. The implemented changes have been shown to be acceptable by reanalyzing all affected accident scenarios. The rescaling of the turbine first stage pressure transmitters further ensures that the high steam line flow setpoint program is enveloped by that used in the revised accident analysis. Turbine first stage pressure also provides input signals to the P-7 logic, which is used to bypass various reactor trip functions below 10: power. The P-7 interlock is now satisfied at approximately 230 MWt rather than 220 MWt. The revised accident analysis demonstrates that this change is also acceptable. The other setpoint/scaling changes within the scope of this Engineering Package relate to control systems and annunciator circuits which are not used to mitigate the consequences of an accident.
Consequently, the modifications addressed by this Engineering Package did not constitute an unreviewed safety question. Since the RPS and ESFAS setpoint technical specification changes were approved in advance of the Engineering Package implementation, specific NRC approval was not required for implementation.
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0 PLANT CHANGE MODIFICATION 95-171 UNIT 4 TURN OVER DATE : 01/24/97 THERMAL POWER UPRATE SETPOINT SCALING
~Summa The Engineering Package implemented the various instrumentation changes necessary to support the thermal power uprate project. The thermal power uprate affected many of the plant ..process variables including core Delta- T, decay heat load, steam flow, steam pressure, and feedwater flow. The following parameters and instruments were affected by the uprate: a) Overtemperature Delta-.T and Overpower Delta-T reactor trip setpoints, b) rod insertion limit summators, c) Delta- T channel deviation alarm setpoint, d) steam flow / feed flow mismatch comparators and high steam line flow setpoint summators, e) turbine first stage pressure, f) main generator hydrogen temperature alarm setpoint, g) main generator megawatt output, h) spent fuel pool high temperature alarm, i) RCP thermal barrier CCW return high temperature alarm, and j) feedwater pump low suction alarm pressure. FPL received approval from the NRC for the setpoint/scaling changes in'dvance of implementation of this Engineering Package. The NRC's safety evaluation concluded that operation of the above systems at the proposed power levels was acceptable.
Safet Evaluation:
This'PS and ESFAS setpoint changes addressed by this Engineering Package (including Overtemperature Delta-T, Overpower Delta-T, High Steam Line Flow and Steam Flow / Feed Flow Mismatch) protect the integrity of the fuel and fuel cladding, and mitigate the consequences of design basis accidents ..
have been shown to be acceptable by reanalyzing The implemented changes all affected accident scenarios'he rescaling of the turbine first stage pressure transmitters further ensures that the high steam line flow setpoint program is enveloped by that used in the revised accident analysis. Turbine first stage pressure also provides input signals to the P-7 logic, which is used to bypass various reactor trip functions .below 10. power. The P-7 interlock .is now satisfied at approximately 230 MWt rather than 220 MWt. The revised accident analysis demonstrates that this change is also acceptable. The other setpoint/scaling changes within the scope of this Engineering Package relate to control systems and annunciator circuits which are not used to mitigate the consequences of an accident.
Consequently, the modifications addressed by this Engineering Package did not constitute an unreviewed safety question. Since the RPS and ESFAS setpoint 'technical specification changes were approved in advance of the Engineering Package implementation, specific NRC approval was not required for implementation.
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PLANT CHANGE MODIFICATION 96-008 UNIT 3 TURN OVER DATE 03/2S/97 POTENTIAL EDG LOCKOUT ROLLOWING NORMAL STOP
~SURES This Engineering Package was developed to eliminate an identified "relay race" between the shutdown relay SDRX (energize to run) and the emergency shutdown relay ESDRX (energize to lockout) in the Unit 3 Emergency Diesel Generators (EDG) control circuitry.
relay race could cause an emergency diesel generator to.
if The'dentified lockout an emergency start signal was received during the coastdown period from 450 to 40 RPM following a normal stop. To eliminate the potential lockout condition, an interlock was added t'o the emergency shutdown circuit to ensure pickup of the SDRX relay before pickup of the ESDRX relay during the subject event.
The interlock was obtained by adding a new time delay dropout relay in the ESDRX circuit. The addition of this relay does not impact
'any of the EDG protective trip functions since the only protective trips energized during an emergency start (overspeed and bus differential) initiate an EDG lockout directly, and do not rely on the. emergency shutdown circuit. The new seismically qualified, Class E relays were seismically mounted in the 3A and 3B EDG idle start panels. Implementation of the circuit changes was allowed in any plant operating mode, but concurrent outages of the 3A and 3B EDQs were prohibited.
This Engineering Package . also corrected several drawing discrepancies that were identified during the EDG circuit review.
Safet Evaluation:
The wiring changes implemented by this Engineering Package enhanced the response of the EDGs to an emergency start signal when received during a normal EDG shutdown. The changes did not impact any other EDQ operating scenario. An engineering review demonstrated that the seismic qualification of the panels in which wiring changes were made was not adversely impacted, due to the small weight changes involved. After modifying the emergency shutdown circuit, no new failure modes were identified that could affect EDG and emergency power safety functions. This Engineering Package established requirements'or functional testing of the revised circuitry to verify performance. Based on the design package evaluation, these modifications did not have any adverse effects on plant safety or operation.. Consequently, these modifications did not involve an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required, for. implementation of the subject modifications.
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PLANT CHANGE MODIFICATION 6- 03.2 UNIT 3 TURN OVER DATE 03/24/97 aUNIT 3 BORON INaTECTION TANK BYPASS MODIRICATION
~Summa This Engineering Package re-routed the Unit 3 safety injection system piping around the boron injection tank (BIT), due to the potential for stress corrosion cracking in the abandoned tank piping. The BIT was an original plant design feature that was rendered obsolete by a re-analysis of the main steam line break (MSLB) event performed for the replacement steam generators in the early 1980s. Over the past fifteen years, the BIT and associated piping has remained a passive part of the safety injection system pressure boundary. Based on an engineering review of observed stress corrosion cracks in the chemical and volume control system boric acid supply piping, the BIT safety injection piping was identified as potentially subject to the same stress corrosion cracking failure mechanisms., Removing the BIT from the pressurized flow path of the safety injection system and re-routing the system piping was determined to be the best technical and most cost effective option. Accordingly, this design package provided a new simplified safety injection flow path, which accommodated the in-situ abandonment of the Unit 3 BIT and associated piping and, equipment. Implementation of the piping changes was allowed in plant operating Modes 5, 6, or defueled. Specific system
-configuration restrictions were imposed by the Engineering Package to ensure that the four high head safety injection pumps remained operable to support Unit 4 operation, in accordance, with plant technical specification requirements.
Safet Evaluation:
The modifications addressed by this Engineering Package were have 'a negligible affect on the hydraulic performance of the shown'o safety injection system during the injection and recirculation phases of an accident, including refueling water storage tank (RWST) drain down time. The piping configuration changes did not alter the response of the Safety Injection System to postulated accident conditions. In addition, no new failure modes were introduced by the flow path change. Based on the design bases and installation criteria evaluated 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 of the piping system changes.
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PLANT CHANGE MODIFICATION 9 6 " 022 UNIT 3 Ec 4 TURN OVER DATE 01/29/97 THERMAL POWER UPRATE IMPLEMENTATION
~Summa This Engineering Package evaluated the power escalation process for thermal uprate, defined specific requirements for escalation above 2200 MW and provided changes to the various engineering documents affected by the change in plant operating parameters, including the UFSAR, Design Basis Documents (DBDs), and Vendor Technical Manuals (VTMs). No hardware changes or accident analysis revisions were specifically implemented by this document. It essentially provided an overview of the impact of the thermal uprate project including associated hardware changes, accident analysis revisions, licensing basis document changes, recommended Emergency Operating Procedure changes, and recommended Off-Normal Operating Procedure changes.
FPL received approval from the NRC for the safety related and power generation system hardware changes and accident reanalyses in advance of implementation of this Engineering Package. The NRC's safety evaluation -concluded that operation of'urkey Point Units 3 and 4 at the proposed power levels was acceptable.
Safet Evaluation:
Implementation of the thermal uprate project did not involve any unusual operating practices or maintenance practices. No changes were made to the designs or methods of operation of components postulated to initiate design basis events such that the probability of their failure was increased. The restrictions and precautions imposed on the power escalation process assured that the plant operating configuration remained within accident analysis limits. The plant procedures associated with power escalation were revised to accommodate uprate for normal, off-normal, and emergency conditions, and were consistent with pre-uprate operating practices and methodology. Since no new equipment or operating practices were invoked for thermal uprate, i the mod'if cat ions did not constitute an unreviewed safety question. The associated technical speci.'fication changes for thermal uprate were approved in advance of the Engineering Package implementation, thus, specif ic NRC approval was not required for implementation the proposed changes.
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PLANT CHANGE MODIFICATION 96-036 UNIT 4 TURN OVER DATE 05/22/96 PRING SETPOINT CHANGE FOR THE PILOT OPERATED. LOCKUP VALVES FOR THE'CC CCW SUPPLY RETURN ISOLATION VALVES
~SUEIIR This Engineering Package installed stiffer springs in the pneumatic actuators of, the component cooling water (CCW)'supply and return isolation valves for the emergency containment coolers (ECCs), to enhance the fail-open operation of the valves, on loss of instrument air. The .actuators utilize a spring-assisted pilot operated lock-
.up valve (POLV) to provide the fail safe operation when instrument air pressure decreases below a nominal 45 psig value. Failure of the lock-up valves to slide to the fail safe position on loss of
.instrument air has been observed during periodic surveillance tests. In each case, an increased friction force on the POLV 0-rings prevented the valves from failing open. To overcome the increased drag force on the POLV spool, a replacement spring with a significantly higher spring rate was installed. The air pressure setpoint at which the POLV actuates was also increased from 45 psig to 60 psig.
This Engineering Package also made the POLV test valves installed by Temporary System Alteration (TSA) 4-96-30-07 a permanently installed part of the CCW return valve actuators. These test valves. allow the POLV to be tested without physically stroking the isolation valve; thereby improving valve reliability. Permanent installation of the test valves does not impact the function of the POLV, the actuator, or the control valve.
Safet Evaluation:
The valve actuator changes implemented by this Engineering Package enhanced the response of the ECCs to accident conditions, when instrument. air is not available. The changes did not impact the ECC control circuit or actuation logic. An engineering review demonstrated that the seismic. qualif ication of the supply and return valves was not adversely affected by the modification due to the small weight changes involved. The replacement springs were shown to be consistent with the original spring material, and compatible with the POLV materials of construction. In addition, functional 'esting of the modified actuators verified proper performance of the new POLV spring and setpoint change. Based on the design package evaluation,'these modifications did not have any adverse effects on plant safety or operation. Consequently, these modifications did not involve an unreviewed safety question'r require changes to plant technical specifications. Therefore;
,prior NRC approval was .not required for implementation of the subject modifications.
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PLANT CHANGE MODIFICATION 96-039 UNIT 3 TURN OVER DATE 05/22/96 PRXNG SETPOINT CHANGE FOR THE P XLOT OPERATED LOCKUP VALVES FOR THE ECC CCW SUPPLY RETURN XSOLATION VALVES
~summa This Engineering Package installed stiffer springs in the pneumatic actuators of the component cooling water (CCW) supply and return isolation valves for the emergency containment coolers (ECCs), to enhance the fail-open operation of the valves on loss of instrument air. The actuators utilize a spring-assisted pilot operated lock-up valve (POLV) to provide the fail safe operation when instrument air pressure decreases below a nominal 45 psig value. Failure of the lock-up valves to slide to the fail safe position on loss of instrument air has been observed during periodic surveillance tests. In each case, an increased friction force on the POLV 0-rings prevented the valves from failing open. To overcome the increased drag force on the POLV spool, a replacement spring with a significantly higher spring rate was installed. The air pressure setpoint at which the POLV actuates was also increased from 45 psig to 60 psig.
This Engineering Package also made the POLV test valves installed by Temporary Sys tern Alterat ion (TSA) 3 - 9 6 06 . a permanently installed'art of the CCW return valve actuators. These test valves allow the POLV to'e tested without physically stroking the isolation valve; thereby improving valve reliability. Permanent installation of the test valves does not impact the function of the POLV, the actuator, or the control valve.
Safet Evaluation:
The valve actuator changes implemented by this Engineering Package enhanced the response of the ECCs to accident conditions, when instrument air is not available. The changes did not impact the ECC control circuit or actuation logic. An engineering review demonstrate'd that the seismic qualif ication of the supply and return valves was not adversely affected by the modifications due to the small weight changes involved. The replacement springs were shown to be consistent with the original spring material, and compatible with the POLV materials of construction. In addition, functional testing of the modified actuators verified proper performance of the new POLV spring and setpoint change. Based on the design package evaluation, these modifications did not have any adverse effects on plant safety or operation. Consequently, these modifications did not involve an unreviewed safety cpxestion or require changes to plant technical specifications. Therefore, prior 'NRC approval was. not required for implementation of the subject modifications.
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PLANT CHANGE MODIFICATION 96-040 UNIT 3 Ec 4 TURN OVER DATE : 02/13/97 APPENDIX R DOCUMENTATION CHANGE IN SUPPORT OP CRs96-023 96-474 96-754 96-951 96-1060 6 96-1152
~Sl1$ UIIR This Engineering Package revised the Turkey Point Units 3 and 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 incorporate changes documented in Condition Report Nos.96-023, 96-474,96-754, 96-951, 96-.1060, and 96-1152.
The revisions implemented by this Engineering Package involved, changes in operator actions, requirements for raceway fire protection, and availability of components and equipment during postulated fire scenarios. Changes to the above documents were also made for miscellaneous cables and equipment associated with systems referenced in the above condition reports.
Safet 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.
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PLANT CHANGE MODIFICATION 9'6 - 07 6 UNIT 3 TURN OVER DATE : 03/17/97 CEMZKT OF THE RCP BUS VNDERVOLTAGE TIME DELAY RELAY CIRCUIT
~summa This Engineering Package replaced the reactor, coolant pump (RCP) bus undervoltage time delay relay in each reactor protection system (RRS) train (UVTD-A 8 UVTD-B) with two parallel undervoltage time delay relays. The reactor coolant system low flow protection subsystem is part of the reactor protection system (RPS) which provides reactor trips to protect the core from exceeding departure from nucleate boiling (DNB) limits during a loss of react'or coolant flow. The RCS low flow protection subsystem includes trips for low flow measured - in the primary piping, RCP bus undervoltage, RCP breaker position, and RCP bus underfrequency. A reactor trip for RCP bus undervoltage is initiated by a one out of two coincident logic of the undervoltage relays from both 4160 volt buses. After the one of two coincidence logic is satisfied, the UVTD-A and UVTD-B rela'ys provide a 1.0 second time delay before the reactor trip (RT) relays drop out'. However, the existing time delay relays are electrically downstream of this coincidence logic. As a result, a single failure of any one time delay relay has the potential to initiate a reactor trip. The new relays will be installed in parallel and will provide coincidence logic, so that, a two out of two failure logic is required for an erroneous reactor will preclude the possibility of a single relay failure.trip. This initiating a reactor trip, thus improving plant reliability for power generation. A single failure that prevents the -relay contacts from changing state on coil deenergization (e.g., contacts stick or become welded) would not adversely affect operation of the RPS due to the availability of the redundant train circuitry.
Safet Evaluation:
These modifications were implemented during a refueling shutdown in accordance with requirements imposed by plant technical specifications. This design modification did not alter the logic of any safety function of the RPS and will enhance the reliability of the RPS, protecting it from erroneous reactor =trips originating from the single failure of 'an RCP bus undervoltage timed delay relay. This Engineering Package did not alter any design bases, safety analysis, operational or test'ing requirements of the reactor protection system. Consequently, 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 these modifications.
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PLANT CHANGE MODIFICATION 96"084 UNIT 3 TURN OVER DATE: 03/28/97 RADIANT ENERGY SHIELDS INSIDE CONTAINMENT
~Sunana This Engineering Package added stainless steel sheet metal lagging over existing Thermo-Lag wrapped raceways inside containment to satisfy 10CFRSO, Appendix R, Section IZZ.G. 2. f requirements for radiant energy shields. Stainless steel sheet metal is known for its ability to provide radiant energy shielding and has been used, in the past inside containment to protect various components and cables'he addition of stainless steel lagging on these components augmented the existing Thermo-Lag material. The lagging provided: a) an augmented radiant energy shield for the Thermo-Lag and protected raceway for a minimum of 30 minutes, b) a vapor barrier that prevents open flaming from developing on the surface of the Thermo-Lag (if subjected to fire), and c) a moisture shield that prevents the Thermo-Lag from falling away from the installed location if subjected to fire and/or hose streams.
The lagging was installed snugly around the Thermo-Lag covered components to prevent moisture intrusion. Stainless steel banding and/or stainless steel pop rivets were used to anchor the lagging.
in place.
Safet Evaluation:
The modifications performed by this Engineering Package restored the 30 minute fire rating to protected raceways inside containment, and resolved the combustibility and degradation issues surrounding Thermo-Lag radiant'nergy shields. An engineering review demonstrated that the seismic qualif ication of the modif ied conduits and terminal/pull boxes would not be adversely impacted by the addition'of lagging materials, due to the small weight changes involved. The review also demonstrated that the lagging would remain in place during a design basis seismic event, and not become a source of debris for the co'ntainment sump screens. The review concluded that the lagged raceways would be enveloped by the existing ampacity correction factors, and that none of the affected circuits would have to be derated. Since the installation of lagging 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 of these modifications.
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PLANT CHANGE MODIFICATION 96-085 UNIT 4 TURN OVER DATE 10/02/97 RADIANT ENERGY SHIELDS INSIDE CONTAINMENT
~Sunnna This Engineering Package added stainless steel sheet metal lagging over existing Thermo-Lag wrapped raceways inside containment to satisfy 10CFR50, Appendix R, Section IIX..G.2.f requirements for radiant energy shields. Stainless steel sheet metal is known ability to provide radiant energy shielding and has been used for'ts in the past inside containment to protect various components and cables. The addition of stainless steel lagging on these components augmented the existing Thermo-Lag material. The lagging provided: a) an augmented radiant energy shield for the Thermo-Lag and protected raceway for a minimum of 30 minutes, b) a vapor barrier that prevents open flaming from developing on the surface of the Thermo-.Lag (if subjected to fire), and c) a moisture shield that prevents the Thermo-Lag from falling away from the installed location if subjected to fire and/or hose streams.
The lagging was installed snugly around the Thermo-Lag covered components to prevent moisture intrusion. Stainless steel banding and/or stainless steel pop rivets were used to anchor the lagging in place.
Safet Evaluation:
The modifications performed by this Engineering Package restored the 30 minute fire rating to protected raceways inside containment, and resolved the combustibility and degradation issues surrounding Thermo- Lag radiant energy shields. An engineering review demonstrated that the seismic qualif ication of the modif ied conduits and terminal/pull boxes would"not be adversely impacted'by the addition of lagging materials, due to the small weight changes involved. The review also demonstrated that the laggi'ng would remain in place during a design basis seismic event, and not become a source of debris for the containment sump screens. The review concluded that the lagged raceways would be enveloped by the existing ampacity correction factors, and that none of the affected circuits would have to be derated. Since the installation of lagging 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 of these modifications.
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~ PLANT CHANGE MODIRICATION 96- 093 UNIT 4 TURN OVER DATE 10/08/97 UNXT 4 ADDITION OR CCW HEAD TANK
~Summa 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 was retained by 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-4-716A&B spring packs to accommodate the required increase in closing torque.
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 assist Operations personnel with level control during periods of reduced CCW inventory.
Safet 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 UFSAR seismic and hurricane wind loads.
The logic change for RCV-4-609 did not adversely affect any system safety function. Modifications performed on the excess letdown heat 'xchanger and system relief valves ensured that system.
inventory and function would not be adversely a'ffected by anticipated overpressure 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 the CCW system changes.
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II PLANT CHANGE MODIFICATION 97-012 UNIT 4 TURN OVER DATE 10/10/97 THERMAL OVERPRESSURIZATION OP ISOLATED PIPING
~Summa 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 potentially 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) drilling 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 containment portion of affected piping associated with containment penetrations. The discharge of each relief valve was piped to a collection tank inside containment.
Safet 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
'ffected 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 UFSAR 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 exist'ing containment penetration design bases and did not introduce any new failure
- The UFSAR commitment modes not previously analyzed. that containment isolation be established assuming an independent single active failure remains intact with the modif ied design.
Accordingly, the implemented changes did not involve an unreviewed safety guestion or require changes to plant technical specifications. Therefore, prior NRC approval was'not required for implementation.
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PLANT CHANGE MODIFICATION 97-026 UNIT 4 TURN OVER DATE 10/06/97 MOV-4-744A AND MOV-4-744B REPACKXNG AND E UALXZING LINE INSTALLATION
~Summa This Engineering Package installed bonnet equalizing lines on the residual heat removal (RHR) system discharge isolation valves, MOV-4-744A and MOV-4-744B. 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.
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 from 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 3/8" stainless steel tubing, associated tube clamps, and supports.
A check valve was installed in each equalizing branch line and a single manual isolation valve was installed in the common tubing run. The manual isolation valve is. required to be locked in the open position during plant operation.
Safet Evaluation:
The modifications addressed by this Engineering Package eliminated a potential failure 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 supports were evaluated in accordance with seismic criteria contained in the UFSAR to 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 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.
34
PLANT CHANGE MODIFICATION 7 - 03 5 UNIT 3 Ec 4 TURN OVER DATE 08/27/97 MONITOR TANK "B" DIAPHRAGM ELIMINATION
'~StlhEIR This Engineering Package removed the diaphragm in the "B" boric aci.'d recycle monitor tank, commonly known as the monitor tank, because it was tom and no. longer needed to satisfy a system functional requirement. The primary function of the tank diaphragm was to maintain the discharge from the plant evaporators in degassed state so that it could be returned'o the primary water storage tank (PWST) without further processing. The tear causes the diaphragm to settle below the water level blocking the outlet nozzle, and impeding normal tank drain down efforts.
Presently, the monitor tank is used for processing laundry waste water. Since laundry waste water is not returned to the PWST, preventing oxygenation of the tank contents is no longer a plant design requirement.
The diaphragm was removed from the tank by trimming the membrane material as close'o the interior tank attachment bracket as possible.
Safet Evaluation:
The Monitor Tank does not serve any safety related .functions and is not required to support safe shutdown of -the plant. Removal of the tank diaphragm was evaluated and found to have no adverse affect, on plant safety or plant operations. .ln addition, the removal process did not impact the existing tank structure or:original design code.
As demonstrated in the Engineering Package, this modification did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of this design change.
35
i
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PLANT CHANGE MODIRICATION 97-041 UNIT 4 TURN OVER DATE 1O/11/97 CORE EXIT THERMOCOUPLE COLUMN REMOVAL - UNIT 4
~Suaana This Engineering Package permanently removed the upper portion of core. exit thermocouple column 53 from the reactor vessel because was damaged beyond repair during removal of the reactor vessel it closure head, in preparati;on for the Unit 4 Cycle 17 refueling outage. The damaged instrument column was removed by cutting the column support tube and enclosed conduits at an elevated location above the upper support plate and sealing the lower portion of the instrument column (and exposed conduit ends) with weld metal. The reactor vessel head penetration was sealed usi.ng the existing seal configuration with a specially fabricated core exit thermocouple nozzle plug that duplicated the existing thermocouple column seal block. The abandoned core exit thermocouples were part of the inadequate core cooling system (ICCS) instrumentation, which was designed to yield information on emergency" core cooling, system operation, and long term surveillance of post-accident decay heat removal. During normal plant operation, i:t was possible to use the core exit temperature information to confirm reactor core design parameters and calculate the quadrant power tilt ratio. The core.
exit thermocouples are classified as Regulatory Guide 1.97 Type A Category 1 variables. Abandonment of the core exit thermocouples did not impact the control room qualified safety parameter display system (QSPDS).
Safet Evaluation:
The modifications described in this Engineering Package permanently disabled 13 core exit thermocouples and left a portion .of the intact support column attached to the reactor vessel internals.
Due to the passive nature of the thermocouple instruments and the thermocouple column sealing device, .no new. failure modes were created by the design change. The configuration of the reactor vessel upper internals package and the head .nozzle plug assembly were analyzed and determined to meet all applicable load combinations. Since the number of remaining operable core exit thermocouples continued to satisfy technical specification requirements, and FPL's commitment to'provide 'core exit temperature monitoring during mid- loop operations, there was no saf ety significance associated with the abandonment of the damaged core exit thermocouple column. Based on the evaluation criteria
. contained in the Engineering Package, the design change did not constitute an unreviewed safety question or. require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation.
36
SECTION 2 SAFETY EVALUATIONS 37
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SAPETY EVALUATION JPN-PTN-SEEZ-88-042 REVISION 8 UNIT APPROVAL DATE : 09/05/97 E-ENERQIZATION OP UNIT 4 4160 VOLT SAPETY RELATED BUSSES
~SURER This evaluation was developed to establish the requirements and restrictions which must be placed on the operation of Units 3 and 4 and their equipment when a Unit 4 4160 volt bus was de-energized and train "A" and "B",load centers were cross-connected. Also examined were technical and licensing concerns associated with de-energizing safety related equipment .and ef fectively 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.
8 updated the electrical design configuration and imposed I'evision new restrictions to accommodate modifications being implemented by PC/M 96-096, "C-Bus Re'liability Improvements Modifications."
Safet Evaluation:
This safety evaluation addressed the technical and. licensing requirements for the de-energization of each Unit 4 4160 volt bus and concluded that the proposed plant configuration and mode of operation was bounded by the technical specifications and did not change the analysis, of accidents addressed in the UFSAR 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 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.
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Ih SAPETY EVALUATION JPN'-PTN-SEEJ-89"085 REVISIONS 10, 11, .and 12 UNIT 3 APPROVAL DATES : Rev.10 02/20/97 Rev.ll 03/09/97 Rev.12 03/13/97 DE-ENERGIZATION OP UNIT 3 4160 VOLT SAPETY RELATED BUSSES
~SllEER This evaluation developed the requirements and restrictions which must be placed on the 'operation of Units 3 and 4 and their equipment when a Unit 3 4160 volt bus was de-energized and Train
>>A>> and >>B>> load centers were cross-connected. Also examined were technical and licensing concerns associated with de-energizing saf e'ty related equipment and effectively removing an emergency diesel generator (EDG) from service as the result of a Unit 3 4160 volt bus outage. The de-energization of a Unit 3 4160 volt safety related bus, with Unit 3 in cold or refueling shutdown (Modes 5 and
- 6) or de-fueled and Unit 4 at. power operation (Mode 1) or below, is sometimes necessary to permit periodic maintenance, testing, or design modifications of the 4160 volt switchgear. De-energization of a 4160 volt bus would cause de-energization of the 480 volt load, centers and motor control centers powered from that bus, if and a loss of power to equipment which may be required to maintain any, 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.
Revisions 10, 11, and 12 updated the bus loading to coincide with recent design changes and changes in plant operating requirements.
Safet 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 configuration and mode of operation was bounded by the technical specifications and did not change the'nalysis of accidents addressed 'in the UFSAR 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 plant changes in procedures, identified in this safety evaluation did not constitute an unreviewed safety question or require changes to plant technical specifications.
Therefore, prior NRC approval was not required for implementation of the actions or precautions identified in this safety evaluation.
39
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SAPETY EVALUATION JPN-PTN" SEMS- 0-041 REVISION 6 UNIT ~
3 APPROVAL DATE : 03/26/97 10 CPR 80. 59 SAPETY EVALUATION - ACCEPTABILITY OP AS-POUND CONDITION POR RHR CHECK VALVE 3 "753A
~Summa This safety evaluation examined the as-found metallurgical defects in the residual heat removal (RHR) system 3A pump discharge check valve 3-753A. In response -to Significant Operating Experience (SOER) 86-3, Turkey Point implemented a disassembly and inspection program on a sampling basis to ensure check valve internals were intact and were not experiencing abnormal wear. During visual
,inspection of the 3A RHR pump discharge check valve, three linear indications were identified on the valve seat. One of the indications cut across the stellite seat and extended into the austenitic stainless steel valve body.. A liquid penetrant examination determined that the other two'efects met the acceptance criteria of ASME Section III. A flaw evaluation was conducted consistent with the analytical flaw evaluation methods contained in ASME Section XI (IWB-3600). Based on this review and the material behavior for the cast austenitic stainless steel valve body, the only relevant degradation expected was fatigue. Due to the low calculated crack growth for the estimated valve duty cycles, it was concluded -that the valve would provide acceptable operation until the end-of-service life of the plant.
Revision 6 documented the valve re-inspection performed during the Unit 3 .Cycle 16 refueling outage. The inspection revealed that there was no increase in the length of the flaw on the valve seat.
Safet Evaluation:
Re-inspection of the linear indications showed that no crack growth occurred over the approximate 60 month operating period. This confirmed the stationary nature of the linear indications and the conservatism of previous analyses. Since the as-found condition of the valve will not impact the capability of the RHR system to perform its safety functions (effectively until the end-of-service life of the plant), the actions or plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant technical specifications.
Therefore, prior NRC approval was not required for implementation of the actions or conditions identified within this evaluation.
40
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SAFETY EVALUATION JPN-PTN-SEMS 010 REVISION 2 UNIT 4 APPROVAL DATE 09/04/97 SAFETY EVALUATION FOR ICW VALVE REPLACEMENT AND B HEADER CRAWL THROUGH INSPECTION
~S11IDER This safety evaluation addressed nine intake cooling water (ICW) isolation valves which were scheduled to be replaced during the 1993 Unit 4 refueling outage. A crawl through inspection and repair of the Unit 4B ICW header and the C ICW pump discharge.
piping was performed during the same outage. The purpose of this safety evaluation .was to assess all potential safety'oncerns caused by activities associated with valve replacement and the piping crawl through inspection/repair. Some of the valve replacements required each of the ICW headers to be individually removed from service. To ensure residual heat removal (RHR) and spent fuel pit (SFP) cooling requirements were met, ICW operations were controlled in accordance with the applicable technical specifications and system operating procedures.
Revision 2 addressed replacement of valve 4-50-310 during the 1997 refueling outage, and the crawl through inspection and repair of the B ICW header.
Safet Evaluate.on:
Because plant operating procedures assure that decay heat removal through the RHR system and the SFP cooling system will be maintained, the ICW valve replacement and the insp'ection/repair crawl through did not adversely impact plant safety and operation.
Removal of valve bodies and piping components during the replace-ment and crawl through activities were evaluated and did not affect the seismic qualif ications of the remaining piping or its operability. The piping met the allowable stresses as defined in, the UFSAR. The actions or plant changes (procedures and/or hardware) identified'n this safety evaluation did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not requir'ed for implementation of the actions or changes identif ied within this evaluation.
41
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SAFETY EVALUATION JPN-PTN-SECS-93"013 REVISION 2 I
UNIT 3 Ec 4 APPROVAL DATE . 09/04/97 SAFETY EVALUATION 'FOR SCAFFOLDING AND COVER FOR SPENT FUEL POOL ROOMS
~Summa This evaluation addressed the installation of scaffolding .and pool covers in the Units 3 .and 4 spent fuel pool rooms to allow the application of coatings to walls and ceiling areas. The purpose .Of this safety evaluation was to demonstrate that the proposed installation of temporary scaffolding, pool covers, and the use of the spent fuel pool cranes during coating application would not adversely affect plant safety 'or operation during any mode, except while the affected unit is in Mode 6, or while handling fuel or inserts in the affected spent fuel pool. These restrictions limited pool heat load to a recently discharged 1/3 core, plus past discharged. cores, consistent with the design basis of the spent fuel cooling system. Scaffolding was installed to provide access to all or most walls, and ceiling areas not located directly above the ,pools. The spent fuel pool gantry crane was used as a traveling staging platform to access adjacent walls, ceiling areas located directly above the pools, and potentially for installation of the pool cover system.
Revision" 2 included calculated pool heatup data supporting the initial 1/3 core restriction.. It also included a Deviation Request Form to allow Engineering to review minor deviations from the cover installation details.
Safet Evaluation:
The temporary structures covered by the scope of this safety evaluation were designed to withstand all applicable loads, including seismic loads. These temporary structures did not modify or actively interact with any plant equipment important to safety.
No new permanent plant ecpxipment was added by the activity, and restrictions. were placed in this evaluation to preclude implementation of the coating work in Mode 6, or while fuel or inserts were handled in the affected pool. The actions or 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.
42
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SAFETY EVALUATION JPN" PTN" SENP" 9 5 - 0 07 REVISION 2 UNIT 3 APPROVAL DATE : 09/10/96 SAFETY EVALUATION FOR OPERABILITY OF RHR DURING INTEGRATED SAFEGUARDS TESTING
~Summa This safety evaluation addressed engineered safeguards integrated testing (ESIT), .performed during each refueling outage, with respect to a generic Westinghouse concern related to the decay heat removal'apabilities of the steam generators (SGs) during plant shutdowns in Mode 5. Westinghouse identified that during Mode shutdowns, there was the potential for gas formation within the steam generator U-tubes, which makes the use of steam generators and natural circulation for decay heat cooling ineffective. Thus, in accordance with plant technical specifications, both trains of the residual heat removal system (RHR) must remain operable during the period of safeguards testing performed in Mode 5. Since safeguards testing was normally performed during Mode 5 and could have involved the potential for temporary alterations in engineered safety features systems (including the RHR system), this evaluation was developed to document that the RHR system remained operable during the period when safeguards testing was conducted with the plant in Mode 5. 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 address the generic Westinghouse concerns, since the ESIT procedures ensure that both RHR trains remain operable at all times during the ESZT without having to perform any compensatory actions.
Revision 2 addresses the temporary test configuration of each 4160 volt bus during the ESIT. The evaluation concluded that both busses will remain available to support RHR pump operability.
Safet Evaluation:
The engineered safeguards integrated test (ESIT) is performed each refueling outage to demonstrate the readiness of emergency power systems and sa feguards equipment.
. This evaluation examined the operability of both RHR trains during ESIT testing because of the potential for steam generator U-tube voiding in Mode 5. This evaluation concluded that ESIT procedures ensure that both RHR trains remain operable at all times during testing without requiring compensatory actions. Consequently, ESIT procedures did not involve an unreviewed safety question nor did they require changes to plant technical specifications. Therefore, prior NRC approval was not required to perform the next outage related ESIT saf eguards tests or to address the generic implications of SG voiding as this could have applied to existing ESIT procedures.
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ll SAPETY EVALUATION JPN- PTN- SENP - 9 5 - 023 REVISIONS 2 and 3 UNIT APPROVAL DATES : Rev.2 10/01/97 Rev.3 10/01/97 SAPETY EVALUATION FOR OPERABILITY OP RHR DURING INTEGRATED SAPEGUARDS TESTING
~Summa This safety evaluation addressed engineered safeguards integrated testing (ESZT), performed during each refueling outage, with respect to a generic Westinghouse concern related to the decay heat removal capabilities of the steam generators (SGs) during plant shutdowns in Mode 5. Westinghouse identified that during Mode 5 shutdowns, there was the potential for gas formation within the steam generator U-tubes, which makes the use of steam generators and natural circulation for decay heat cooling ineffective. Thus, in accordance with plant technical specifications, both trains of the residual heat removal system (RHR) must remain operable during the period of safeguards testing performed in Mode 5. Since safeguards testing was normally performed during Mode 5 and could have involved the potential for temporary alterations in engineered saf ety features systems (including RHR), this evaluation was developed.to document that the RHR system remained operable during the period when safeguards testing was conducted with the plant in Mode 5. 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 address the generic Westinghouse concerns.
Revision 2 addresses performance of the ESIT with Unit 4 in Mode 6 with reactor coolant system level two feet below the reactor vessel flange. Revision 3 addresses RHR operability during test transients with respect to technical specification compliance. It is concluded that both 4160 volt busses will remain available to support RHR pump operability.
Safet Evaluation:
This safety evaluation examined the operability of both RHR trains during ESIT testing because of the potential for steam generator U-tube voiding in Mode 5. The examination concluded that the ESIT procedures ensure that both RHR trains remain operable at all times during testing without requiring compensatory actions.
Consequently, ESIT procedures did not involve an unreviewed safety question nor did they require changes to plant technical specifications. Therefore, prior NRC approval was not required to perform the next outage., related ESIT safeguards tests or to address the generic implications of SG voiding as this could have applied to existing ESIT procedures.
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0 SAPETY EVALUATION JPN-PTN-SECP-95-046 REVISION 0 UNIT 3 & 4 APPROVAL DATE : 09/11/96 SAPETY EVALUATION FOR UNIT 3 AND UNIT 4 TWENTY-FIFTH YEAR CONTAINMENT TENDON SURVEILLANCES
~Snnnna Containment tendon surveillance testing is performed every fifth year from the date of the initial containment structural integrity.
test, in accordance with plant technical specifications. This safe'ty evaluation examined the implementation activities (e.g.,
mobilization, equipment erection on the containment dome, heavy load lifts, -extraordinary weather precautions, demobilization activities) associated with the twenty-fifth year tendon surveillances on Units 3 and 4 to ensure that adverse interactions with safety related equipment are precluded during the heavy load handling and "construction-like" activities associated with tendon testing. Appendix B to this safety evaluation contained a list of the restrictions imposed on. the overall test sequence. The safety evaluation concluded that sufficient precautions and limitations exist to permit performance of the surveillance tests in any plant operating mode.
Safet Evaluation:
This safety evaluation examined the construction and testing equipment to be used during the surveillance test and demonstrated that any potential adverse interactions could be accommodated by a) the design attributes of the surveillance equipment and structures, b) the facilities with which the surveillance equipment or structures could interact, or c) the restrictions imposed by Appendix B of the safety evaluation. Zn addition, the safety evaluation examined the detensioning procedure and concluded that the =-surveillance activity will not affect the ability of the containment to withstand maximum design basis loads. The safety evaluation concluded that the activities associated with twenty-fifth year tendon surveillance test 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 identified within this evaluation.
0 ik
SAPETY EVALUATION ZPN-PTN-SENS-95-'052.
REVISIONS 1 and 2 UNIT 3 APPROVAL DATES : Rev.1 03'/03/97 Rev.2 03/13/97 SAPETY'VALUATION POR REVXSION OP UNIT 3 DIESEL PUEL OXL TRANSPER SYSTEM TECHNICAL SPECXPXCATXON BASES
~SURER An earlier evaluation, JPN-PTN-SENS-95-050, documented the operability of the Unit 3 emergency diesel generators (EDGs) and
.compliance with technical specification requirements. for EDG diesel fuel oil transfer from the storage tank to the day tank using manual actions. This safety evaluation addressed changes in the UFSAR to clarify the use of manual actions to fulfill the Unit 3 EDG fuel transfer function on loss'f instrument air, and proposed changes in the bases of technical specifications to clarify the design differences between Unit 3 and Unit 4 diesel fuel transfer systems. The diesel fuel transfer valves. on Unit 3 are designed to fail closed on loss of instrument air which prevents the automatic transfer of fuel oil from the Unit 3 diesel oil storage tank to the EDG day tanks. The use of portable compressed air bottles to manually restore function to the EDG automatic fill isolation valves were proceduralized as part of the controlling procedure for loss of instrument air. This evaluation concluded that the .use of manual actions to fulfillthe Unit 3 EDG fuel transfer function is within the current design basis of the plant.
Revision 1 clarified the position that the use of manual actions in place of automatic'peration is an acceptable compensatory measure under loss of instrument air conditions, and is consistent with the design intent. Revision 2 further clarified that the EDGs are operable under loss of instrument air conditions, provided the fuel transfer valves can be operated with a portable air source.
Safet Evaluation:
This evaluation supports revisions to technical specification bases for the Unit 3 EDG diesel'uel transfer system surveillance requirements .which'eflect that design basis EDG diesel fuel transfer functions are satisfied by using either existing procedura'lized manual actions on- loss of instrument automatic action of the fill air'r transfer isolation valves. Since by manual actions to fulfill the Unit 3 EDG fuel oil transfer functions are within the current plant'esign basis, this evaluation concluded that the proposed clarifications to technical specification bases and the UFSAR 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 identified within this evaluation.
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SAFETY EVALUATION JPN-PTN-SEMP 004 REVIS 1ON 1 UNIT 3 R 4 APPROVAL DATE 02/27/97 EVALUATION OF AP AEQGWLEX INSULATION FOR CONDENSATION DURING TEMPORARY CONTAINMI2FP COOLING
~SIIMIIS This safety evaluation addressed the Appendix R requirements for the use of foam insulation which was installed on portions of the component cooling water (CCW) system. Under a separate ication (PC/M) 95- 054, permanent plant modif ications.
Plant'hange/Modif were implemented to allow the use of chilled water in the CCW system to provide chilled water to the normal containment coolers (NCCs) to cool the containment buildings during refueling outages.
The PC/M identified. the permanent installation of foam insulation for portions of the affected CCW piping to minimize condensation within the auxiliary building. The areas identified were the Unit 3 and 4 boric acid evaporator rooms, the Unit 3 pipe and valve room and other areas as required. This safety evaluation addressed the Appendix R requirements for the use of the foam insulation and associated material qualif ications and suitability requirements.
Based on a review of the material properties, FPL determined that
.the material was acceptable for use in the auxiliary building. The evaluation within PC/M 95-054 addressed the piping and pipe supports associated with temporary containment cooling.
Revision 1 revised the scope of the safety evaluation to include the additional CCW piping that was insulated under PC/M 95-177.
Safet Evaluation:
This safety evaluation concluded that the use of insulating foam on the CCW piping did not adversely affect the performance of CCW pumps or heat exchangers, since the basis for system heat removal did not consider the piping as a heat transfer surface. The evaluation also examined the seismic performance, the additional mass attached to CCW system piping, and the chemical composition, flame retardant characteristics, and temperature rating of the foam insulation. The total amount of foam material added to each fire zone was evaluated. The use of "-the foam insulation did not adversely affect CCW system performance or plant operations.
Consequently, the use of foam insulation on selective portions of the CCW piping did not constitute an unreviewed safety question or require changes to the plant technical specifications. Based oq the above, prior NRC approval was not required for the use of foam insulation on those portions of CCW piping identified within this safety evaluation.
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SAPETY EVALUATION JPN-PTN-SEMS-9 6- 014 REVIS ION 1 UNIT 3 & 4 APPROVAL DATE . 02/06/97 SAFETY EVALUATION POR A TEST OP THE USE OF SUB-MICRON ULTRAPINE FILTERS IN THE CVCS AND SFP SYSTEMS
~Summa This evaluation served to allow the temporary use of the ultrafine cartridges with absolute filtration ratings in the reactor coolant (RCS), seal water injection, and seal water return filters in the.
chemical volume and control system to reduce plant radiation levels and to extend the life of reactor coolant pump seals. The ultrafine filter program will proceed in three phases. Because the filters proposed for use must be specifically designed for the individual filter filter housings, fit Phase I will involve a demonstration for proper and performance of near equivalent rated absolute filters cartridges. Only one test cartridge will be installed in the parallel filter paths at a time; the other path(s) will contain rated filters of the type currently used. Phase II of the testing program is a gradual reduction in the absolute rating of the filters used. This will gradually filter out finer and finer particles as the overall RCS particulate inventory is reduced. This will continue until the desired RCS cleanliness is level is reached. Phase III involves the permanent use of under formal plant design change documentation. Phase I of these'ilters the program was evaluated in a previous safety evaluation. This evaluation only addressed Phase II.of the ultrafine filter program.
Safet Evaluation:
This evaluation addressed the use of ultrafine filter cartridges for the RCS, seal water return, and seal water injection filters.
This evaluation concluded that these ultrafine filters will meet all current design criteria for the systems identified above.
Failure modes were evaluated and precautions have been established to monitor these filters more closely during the test'eriod. The use of these filters does not change system design bases, functions, and operation of any safety related equipment, and will not adversely affect any safety related structures, systems or components. Therefore, 'the testing implementation and plant 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.
i SAFETY EVALUATION JPN- PTN- SENP - 9 6 " 017 REVISION 0 UNIT 3 R 4 APPROVAL DATE 05/09/96 SAPETY EVALUATION FOR UPSAR CONSISTENCY REVIEW AND UPDATE
~Summa In response to NRC Information Notice '95-54, a self-assessment activity was undertaken to review the Turkey Point UFSAR and determine the nature and extent of discrepancies between UFSAR descriptions and the design and procedural configuration of the plant. Chapters 3, 4, 6, 7, 8, 9, 11, 12, and 14 of the UFSAR were included in the review. Ninety-six findings were identified by the self-assessment team and forwarded to Engineering for review, final disposition, and incorporation into the UFSAR as appropriate.
Further review by engineering found that 67 of the findings were already addressed by in-process change documents,, or determined not to be a discrepancy. Twenty-four of the findings identified administrative clarifications or addressed historical information in the-UFSAR. Five of the 96 findings documented statements in the UFSAR that were inconsistent,."with the plant design or plant operating procedures. No operability issues were, identified and no physical design changes were-required to resolve the discrepancies. to this safety evaluation lists each review finding, its classification, and status. Attachments 2 through 6 contain the FSAR 'User Comment Forms for the five identified discrepancies. documents the resolution of each finding.
Safet Evaluation:
The safety evaluation documented that all .UFSAR findings were dispositioned and that none of the discrepancies impacted plant safety or operation. 'he evaluated discrepancies did not change the operation,,function, or design bases of any structure, system, or component important to safety .as described in the UFSAR.
Consequently, the UFSAR 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 changes identified within this evaluation.
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SAPETY EVALUATZON JPN-PTN-SENS 024 REVISION 1 UNIT 3 & 4 APPROVAL DATE 03/21/97 SAPETY EVALUATZON TO SUPPORT A CROSS-CONNECTED RHR SYSTEM REALZGNMENT DURZNG A REPUELZNG OUTAGE
~SllESR This, safety 'evaluation was developed to address the impact of a proposed residual heat removal (RHR) system realignment on plant technical specifications during refueling activities. This temporary realignment was required to ensure that RHR cooling would remain available to remove decay heat during the period when repairs were performed on the nozzle of the 4B RHR heat exchanger.
At the same time these repairs were performed, the 4B RHR pump was unavailable due to an outage on the 4B 4160 volt bus.
Additionally, maintenance was performed on the RHR pump discharge check valves, and IST surveillance testing was required to ensure operability of these valves. This evaluation assessed the sequence of RHR pump discharge check valve IST surveillance testing to ensure the credited RHR "loop" remained operable. As a result of these refueling outage maintenance activities, the operable RHR "loop rec(uired by Technical Specifications was composed of the'A RHR pump combined with the 4B RHR heat exchanger. Technical Specifications require one operable and operating RHR loop while in operating Mode 6, provided that the refueling. pool water level remains 23 feet or more above the reactor vessel flange., In addition to RHR technical specification requirements, procedures imposed stringent administrative controls on support systems associated with each operable RHR decay heat cooling loop. These procedures address Component Cooling Water (CCW) and Intake Cooling Water (ICW) systems and their power supplies.
Revision 1 extended the scope of the evaluation to include the generic alignment of any RHR pump with its opposite train heat exchanger.
Safet Evaluation:
Thi;s evaluation examined critical design and licensing criteria which involved 'single failure criteria, support system alignments, power supply alignments, and seismic criteria. This evaluation found that the temporary realignment of RHR and associated maintenance activities did not adversely affect safety functions and met all technical specification and procedural requirements.
Consequently, the RHR realignment and .associated .maintenance activities did not involve an unreviewed safety cpxestion or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the RHR realignment and I
other actions .identifi'ed within this evaluation.
50
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SAFETY EVALUATION JPN-PTN" SENP-96-025 REVISION 0 UNIT 3 & 4 APPROVAL DATE 06/04/96 SAPETY EVALUATION POR REMOVAL OP THE RCC CHANGE PIXTURE PROM THE PSAR
~Summa This safety evaluation was prepared to delete the description of the rod control cluster (RCC) change fixture from the UFSAR. The RCC change fixture is a necessary refueling device during fuel reloads performed with an in'-core fuel shuffle. This was the practice at Turkey Point- until the replacement of the steam generators in the early 1980's. Subsequently, refuelings have been performed by offloading the entire core to the spent fuel pool and performing the RCC change activities in the spent fuel pool. The RCC change tool used in the spent fuel pool provides an equivalent function to the RCC change fixture inside containment. The proposed UFSAR changes were provided as Attachment 1 to the safety evaluation.
Safet Evaluation:
Abandoning the RCC change fixture in place and removing reference to it design in both the UFSAR and plant procedures did not alter the basis, functions, or operation of any safety related equipment. The practice of performing full core offloads has been evaluated and found to improve safety with respect to a number of concerns, including reduced consequences for refueling cavity seal failures and minimizing operating time at reduced inventory conditions. The actions and 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.
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SAFETY EVALUATION JPN" PTN-SENS-9 6- 028 REVISION 0 UNIT 3 Ec 4 APPROVAL DATE : 09/19/96 SAFETY EVALUATION UFSAR OPERATIONAL REVIEW AND UPDATE
~SllSER In response to NRC Inspection Report Nos. 50/250-94-01 and 50/251-94-01, a self-assessment activity was undertaken to review the Turkey Point UFSAR and determine the nature and extent of discrepancies between UFSAR descriptions and .the plant procedures.
The review consisted of selected portions of the FSAR identified as containing operational or procedural information. Forty-five findings were identified by the self-assessment team and forwarded to Engineering for review, final disposition, and incorporation into the UFSAR as appropriate. Further review by engineering found that 21 of the findings were already addressed by in-process change documents, or determined not to be a discrepancy. Sixteen of the findings identified administrative clarifications or addressed historical information in the UFSAR. Eight of the 45 findings documented statements in the UFSAR .that were inconsistent wi;th the way in which the plant was operated. No operability issues were identified and no physical design changes were required to resolve the discrepancies. Attachment 1 to this safety evaluation lists each review finding, its classification, and status. Attachments 2 through 12 contain the FSAR User Comment Forms for the identified discrepancies.
Safet Evaluation:
The safety evaluation documented that all UFSAR findings were dispositioned and that none of the discrepancies impacted plant safety or operation. The evaluated discrepancies did not change the operation, function, or design bases of any structure, system, or component important to safety as described in the UFSAR.
Consequently, the UFSAR 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 .
changes identified within this evaluation.
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a SAFETY EVALUATION JPN-PTN-SECS- 6- 033 REVISIONS 1 and 2 UNIT 4 APPROVAL DATES : Rev.1 08/29/96 Rev.2 04/25/97 SAPETY EVALUATION POR INSTALLATION OP REMOTE CAMERA POR 'B'CP OIL LEVEL VERIFICATION
~Summa This safety evaluation addressed the installation of a remote camera assembly inside containment to monitor the 4B reactor coolant pump (RCP) motor oil level. The assembly consisted of bent steel conduit wi;th a miniature video camera and light attached to the bent end. The total weight of the conduit and camera assembly was approximately 10 pounds, resulting in insignificant loads being applied to the RCP motor oil piping and shield wall.
The video feed from the camera was routed to a communication box near the elevator platform on the 30'-6" elevation of the containment building. The video cable was connected to spare telephone leads which terminated outside containment in the cable spreading room.
Revision 1 addressed minor administrative changes to the evaluation. Revision.2 addressed relocation of the camera assembly to monitor leakage at the pump seal housing to main flange joint.
Safet Evaluation:
a An engineering review demonstrated that the seismic qualification of the RCP motor oil piping and supports would not be adversely impacted the addition of the camera and conduit assembly, due to the small weight changes involved. The review also demonstrated that the camera assembly would remain in place during a design basis seismic event, and not damage adjacent equipment considered important to safety. Since the installation of camera assembly 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 not required for implementation of the actions identified within this evaluation.
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0 SAPETY EVALUATION JPN-PTN-SEMS-96" 037 REVISION 0 UNIT 4 APPROVAL DATE . 05/31/96 SAFETY 'VALUATION POR THE TEMPORARY INSTALLATION OR A PREEZE SEAL AND 'LIND FLANGE PER TSA- 04 "9 6- 046- 11 TO SUPPORT MOV-4-350 VALVE REPAIRS
~summa This safety evaluation addressed the use of a freeze seal. to temporarily isolate a section of the chemical and volume control system (CVCS) to perform maintenance activities on the emergency boration valve, MOV-4-350. The freeze seal was to be applied downstream of check valves 4-351 and 4-352 as. a backup feature in the event that check valve 4-351 leaked past its seat. The freeze seal was considered to provide a housekeeping function since the upstream check valves provided the pressure boundary function for the charging pump suction header during the maintenance evolution.
This evaluation addressed plant operation with the emergency boration line isolated during the repair process, and the potential impact on charging system operation due to freeze seal failures.
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. A blind flange was installed at the upstream flange connection for FT-4-110 to allow the boric acid flow path from the Boric Acid Storage Tanks (BASTs) 'to be maintained as an operable flowpath during the maintenance evolution.
Safet Evaluation:
This evaluation addressed the temporary uncoupling of the emergency boration flowpath caused by removal of FT-4-110, the impact on plant operation, and the various precautions imposed to ensure the safe conduct of maintenance. Strict controls were imposed on the freeze seal process and, contingency measures were 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.
SAPETY EVALUATION JPN-PTN-SEMS 038 REVISIONS 0 and 1 UNIT 3 APPROVAL DATES : Rev.0 08/22/96 Rev.1 03/31/97 SAPETY EVALUATZON POR UNXT 3 STEAM GENERATORS ~
SECONDARY SIDE POREXGN OBJECTS
~Summa This evaluation addressed the potential safety significance of operating the Unit 3 steam generators '(SG) 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 have addressed the acceptability of continued Unit 3 operation while these foreign objects remained in the steam generators and associated systems. The purpose of this evaluation was: (1) to re-examine the analyses, results, requirements, and restrictions of previous evaluations while applying recent industry standards; (2) to document the methodology for determining the interval between steam generator eddy current tests as affected by estimated steam generator tube wall wear times; and (3) to provide a single Unit 3 safety evaluation to assess and document all the Unit 3 steam generator foreign obj ect estimated wear times as adjusted by updated steam generator eddy current data and steam generator Foreign Object Search and Retrievals (FOSAR) results.
Revision 1 addressed the impact of one new unretrievable foreign object identified in the 3B steam generator during the Cycle 16 refueling outage.
Safet Evaluation:
Previous safety evaluations prepared for each SG secondary side foreign object have considered the effects of the object upon tube integrity, chemistry; SG instrumentation, the main steam system, and SG 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 steam generators 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 steam generators, or endorsement of the programmatic actions identified within this evaluation.
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SAFETY EVALUATION JPN" PTN-SEMS-96-040 REVISION 0 UNIT 3 Ec 4 APPROVAL DATE 07/25/9 6 SAFETY'EVALUATION FOR THE TEMPORARY INSTALLATION OF DRAIN HOSES AND PERFORMANCE OF HOT SPOT FLUSHES ON THE RHR SYSTEM PER-TP-96-046 AND TP"96-047
~Summa This safety evaluation examined the procedure for flushing the residual heat removal (RHR) system with water from the refueling water storage tank, to eliminate radioactive hot spots at various sys'em drain locations. The affected drains were located in the suction piping from the refueling water storage tank and south containment recirculation sump, and at the RHR heat exchangers.
The flushes were performed by installing a flush adapter and tygon hose to the discharge of each drain valve, routing the hose to the nearest suitable floor drain, and opening the drain valve for approximately 20 seconds. The affected drain valves were flushed one. at a time while the RHR system remained operable in the normal standby valve lineup (Modes I - 3) . A flushing flow rate of'5 gpm was expected through the drain piping based on the static head of the refueling water storage tank. As a precautionary measure, an additional ball valve was used with the flush adapter on the refueling water storage tank suction piping, drains, to provide a backup isolation capability (in lieu of RWST isolation) in the event that the piping drain valve could not be re-closed when the flushing activity was complete.
Safet Evaluation:
This evaluation addressed the temporary configuration of the system with the installed flushing adaptor, the impact on plant operation, and the various precautions imposed to ensure safe conduct of the maintenance activity. Strict controls were imposed on the flushing process and contingency measures were developed to establish pressure boundary integrity for the open system should a drain valve fail to re-close, or actuation of the engineered safety features occur. "
Based on the precautions identified, the evaluation concluded that the maintenance activity could be performed in Modes 1 - 3, 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 identif ied within this evaluation.
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SAFETY EVALUATION JPN- PTN- SEIS - 9 6 - 042 REVISION 0 UNIT '3 & 4 APPROVAL DATE 08/23/96 SAFETY EVALUATION FOR THE OPERATION OF RAD" *-6458 THE REACTOR VESSEL HEAD LEAK DETECTION SYSTEM ON A NON-CONTINUOUS BASIS
~S'umma This evaluation analyzed the impact associated with operation of
.the reactor vessel head leak detection system on a non-continuous basis, rather than a continuous basis as described in the UFSAR.
The reactor. vessel head leak detection system was installed at the request of plant management in the late 1980's to enhance reactor vessel head leak detection capability. The system is designed to detect an increasing trend in radioactivity level in the control rod drive mechanism (CRDM) cooler ductwork over containment background levels which would indicate the presence of a reactor vessel head leak. Performance of the system is based on the capability of the skid to detect a difference in activity levels between the two sample points. Due to the recent upgrade in containment atmosphere gaseous and particulate radiation monitoring capability, efforts to more thoroughly seal the reactor coolant system'RCS) pressure boundary (e.g., CRDM canopy seal clamps, crush resistant O-rings), and a lessened tolerance for RCS leakage, the usefulness of the reactor vessel head leak detection system has diminished since its original installation. Rather than removing the system completely, this safety evaluation provided the necessary justification to reduce its hours of operation, and reduce the dose rate received by maintenance personnel attempting to maintain the system operating in a continuous sampling mode.
Safet Evaluation:
The reactor vessel head leak detection system was installed at the request of .plant management and did not perform a safety related function. Changing its operating philosophy did not adversely affect any structure, system, or component considered important to safety. Since no physical plant changes are required to implement the new monitoring scheme, 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 not required for implementation of the actions or changes identified within this evaluation.
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SAFETY EVALUATION JPN-PTN-SEIS-96-044 REVISION 0 UNIT 3 & 4 APPROVAL DATE 08/01/96 SAFETY EVALUATION FOR ABANDONING SPENT FUEL PIT BRIDGE CRANE LASER RANGE FINDER
~Slllnmll This safety evaluation analyzed the impact on plant safety and operation associated with abandonment of the laser range finding systems on the Unit 3 and 4 spent fuel pit bridge cranes. The existing equipment was considered to be broken beyond repair and new units, different in form and fit, would be required to restore the systems to operable status. Use of the fixed scale and pointer system is preferred by plant personnel for fuel handling since is simple, accurate, reliable, and maintenance-free. Additionally, it the scale and pointer is system is referenced directly to individual fuel 'storage cells, whereas .the laser range finding system indicates the distance in feet from the spent fuel pit walls, and does not provide direct indication of storage rack positions. The laser range finding system is physically and electrically independent of the crane's load handling system. The required UFSAR changes were provided as Attachment 1 to the safety evaluation.
Safet Evaluation:
The laser range finding system did 'not serve any safety related functions and had no effect on the operational capability of the spent fuel pit bridge crane, or its load handling system. Since the in-place abandonment of this equipment had'o adverse impact on plant safety or fuel handling operations,.it was concluded that the actions and 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.
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SAFETY EVALUATION PTN-ENG-SEES-96-046 REVISION 0 UNIT 3 & 4 APPROVAL DATE '08/26/96 SAFETY EVALUATION FOR ESSENTIAL E UZPMENT AFFECTED BY A HEAVY LOAD DROP RESULTING'ROM EPS MODZFZCATZONS
~SEAS This safety evaluation re-verified the acceptability of the safe load paths for the turbine gantry crane and spent fuel pool cask.
crane in light of discrepancies noted'between the'plant procedure for heavy load handling, and commitments made to the NRC regarding NUREG 0612. In addition,.field walkdowns identified the presence of several conduits installed during recent plant modifications such as the Emergency Power Upgrade Project which had the potential to be impacted by a heavy load drop. The areas of'he plant examined in the safety evaluation included a) the area north of the Unit 3 switchgear rooms, b) the auxiliary building roof and volume control tank roof areas, and c) the areas above the Unit 3 and 4 component cooling water room roof grating. Consistent with NUREG 0612, this evaluation used the basic criterion that sufficient equipment be available to bring the plant to cold shutdown conditions during a postulated heavy load drop event. The essential equipment required for safe shutdown,was based on the
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established Appendix R Essential Equipment List. The Appendix R
,Safe Shutdown Analysis was used as guidance in developing the applicable safe shutdown assumptions.
Safet Evaluation:
The safety evaluation demonstrated that additional restiictions were required to preserve safe shutdown capability during heavy load handling activities north of the 'Unit 3 switchgear rooms.
Consequently, additional restrictions were imposed on turbine gantry crane operation to preclude heavy load handling in the affected area. These actions did not constitute an unreviewed safety question- or require changes to plan't technical specifications. Therefore, prior NRC approval was'not required for implementation of the actions or changes identif ied within this evaluation.
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SAFETY EVALUATION PTN-ENG" SEMS-96-048 REVISION 0 UNIT 4 APPROVAL DATE : ~ 08/27/96 SAPETY EVALUATION FOR MEASURING ICW HEADER PLOW PER TP-96-070
~Summa This safety evaluation was prepared to evaluate a temporary re-
.alignment. of the intake cooling water (ICW) system for flow measurement purposes in accordance with Temporary Procedure TP 96-
- 70. Obtaining an accurate measurement of ICW flow was a commitment made in LER 50/250-95-003, "Intake Cooling Water System Flow Rate Found less Than Required By Design Basis," to facilitate an assessment of the condition of the ICW basket strainers. The proposed ICW system alignment was similar to that used for ICW pump in-service testing in that two ICW pumps were aligned to the operable header supplying two component cooling water (CCW) heat exchangers, with the third. ICW pump aligned to the two turbine plant cooling water (TPCW) heat exchangers and the third CCW heat exchanger. The CCW heat exchanger on the inoperable header was required to be taken out of service during the test with the CCW flow isolated. The test was performed by closing the basket strainer outlet isolation valves to establish zero flow and then incrementally throttling the valves open to for'alibration obtain the required flow data. The test alignment required entry into the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> technical.,specification Action Statement for one inoperable ICW header. In addition, the test alignment caused increased flows through the TPCW heat exchangers. The TPCW heat exchangers are located downstream of the safety related (Quality Group C) boundary and serve non-safety related functions only.
'I Safet Evaluation:
The ICW and CCW systems are capable of performing their design basis heat removal functions with the minimum amount of equipment required to be operable by plant technical specifications.
Although a relaxation of the single failure criterion is permitted when operating under a technical specification Action Statement, contingency measures were developed to restore the normal ZCW and CCW system alignments should an emergency or abnormal ICW temperatures occur. The actions authorized 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. test procedure.
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SAFETY EVALUATION PTN-ENG-SECS-96-059 REVISION 0 UNIT 3 APPROVAL DATE 09/25/9 6 EVALUATION OF TEMPORARY SCAFFOLDING WITHIN CONTAINMENT FOR USE IN REPLACING RELIEF VALVE RV-3-203
~Summa This safety evaluation addressed the installation of scaffolding inside containment to allow replacement of the Unit 3 letdown line relief valve, RV-3-203. The purpose of the evaluation was to demonstrate that the temporary scaffolding structure would not adversely affect plant safety or operation if erected in Mode 3 The evaluation examined the potential for adverse seismic or'elow.
interactions between the scaffolding and adjacent safety related equipment,.the potential impact on containment free volume and heat sink analyses due to .the metal scaffold support members, the potential impact on the containment hydrogen generation analysis due to the zinc based scaffold support member coating, the potential interaction with the containment sump, and containment combustible loading.
Safet Evaluation:
The temporary structure covered by the scope of this safety evaluation was designed to withstand all applicable loads, including seismic loads The temporary structure did not modify or
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actively interact with any plant equipment important to safety. An engineering review demonstrated that the applicable containment analyses were not affected by the composition of the scaffolding structure (i.e., zinc, metal, wood) due to the small amount of material involved. The actions or 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.
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SAFETY EVALUATION P TN- ENG - SENS - 9 6 - 0 62 REVISION 1 UNIT 3 APPROVAL DATE 09/27/9 6 SAFETY EVALUATION TO PERMIT WIDER PRT LEVEL OPERATING BAND
~Summa This safety evaluation was prepared to address plant operation with the pressurizer relief tank (PRT) level indicator LT-3-470 operating beyond its stated accuracy. Vendor literature indicated.
that the nominal instrument error should be 1% (indicated error) while actual in-situ operation displayed an error of approximately 3'; ~ The purpose of this safety evaluation was to address operation of the plant with the existing operating level band assuming a higher instrument error. The setpoints associated with PRT operation were established by Westinghouse. The low level setpoint of 68% (z 1% accuracy) is intended to insure that the design basis pressurizer steam space release would not result in exceeding the PRT design limit of 200 F. The 83% high level limit is specified such that on a pressurizer steam space release, the pressure in the PRT would not exceed 50 psig. Indication of PRT level is a Regulatory Guide 1.97, Category D variable. 'The stated regulatory guide function is to monitor plant operation.
Revision 1 addressed the impact of the higher instrument error on the reactor coolant system leak rate calculation.
'Safet Evaluation:
The PRT does not perform a safety related function. Consequently, inadvertent .operation of the PRT rupture disk due to level uncertainty is not a malfunction of equipment that could adversely affect the plant's ability to respond to an accident or transient.
The conditions identified in this safety evaluation did not constitute an unreviewed'safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the actions or changes identified within this evaluation.
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SAFETY EVALUATION PTN-ENG" SECS-96-065 REVISIONS 0 and 1 UNIT 3 APPROVAL DATES : Rev.0 12/17/96 Rev.1 01/30/97 10 'CPR 50.59 SAPETY EVALUATION FOR UNIT 3 CYCLE 16 REFUELING OUTAGE TURBINE OVERHAUL SUPPORT
~StlEIIR This safety evaluation was prepared for the Unit 3 Cycle 16 refueling outage to address a) use of the Turkey Point Units 1 & 2 turbine gantry crane on the Unit 3 turbine operating deck; b) lifting and transport of several heavy turbine components over safety related equipment and outside the approved safe load paths; and c) transport of the high pressure turbine rotor from the Unit 4 laydown area to the turbine staging area located south of the site cafeteria building.
Several restrictions were identified in the evaluation to permit use of the Units 1 a 2 turbine gantry crane near safety related equipment. In addition, compensatory measures applicable to both cranes were identified to address lifting and transport of the turbine loads outside of the safe load path zones. Rigging options were also identified as defense-in-depth protection from load drops. A review of plant drawings and procedures was performed to evaluate the heavy haul route from the laydown area to the turbine staging area. The review concluded that the only structure of concern for the proposed heavy haul route was the underground intake cooling water (ICW) pipes, located just south of the Unit 4 laydown area. This section of piping was previously evaluated for the haul route of a main station transformer and a turbine-generator rotor assembly. It was concluded that the movement of the HP rotor assembly was encompassed by the two previous haul route evaluations, and that the existing utilities along the proposed route would not he adversely affected by the subject activity.
Revision 1. addressed comments from the Plant Nuclear Safety Committee and other administrative items.
Safet Evaluation:
The safety evaluation addressed the various failure modes and effects of handling heavy loads over safety related equipment.
Strict administrative controls and compensatory measures were imposed to ensure that safety related equi.pment under the crane travel path would not be adversely affected by a heavy load drop.
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 not required perform the planned turbine overhaul activities.
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SAPETY EVALUATION PTN-ENG-SENS-96-068 REVIS ION 0 UNIT 3 Ee 4 APPROVAL DATE 12/05/96 SAPETY EVALUATION RELATED TO REPOSITIONING OP VALVE HV-7 ON THE POST ACCIDENT CONTAINMENT VENTILATION SYSTEM
~Summa This safety evaluation was prepared to assess the acceptability of maintaining valve HV-7 on the post accident containment ventilation (PACV) system normally closed during plant operation. Maintaining this valve in a normally closed position reduces the potential to overpressurize the PACV filters and allows greater operational flexibilityfor the post accident hydrogen monitoring (PAHM) system The potential to and post accident sampling system (PASS) .
overpressurize the PACV filters with the current valve alignment was documented in a condition report and involved cross connecting the "A" and "B" PASS/PAHM system sensing lines in an alternate sample alignment. To demonstrate the acceptability of. leaving valve HV-7 in a normally closed position, a detailed dose assessment of operator action to re-open the valve was performed based on the time when operation of the PACV system is predicted
'(17 days post-accident).
Safet Evaluation:
Operation with valve HV-7 closed eliminates a potential failure mode'or the PACV filter housings. The evaluation demonstrated that operator action to re-open the valve for PACV operation would not result in acceptable dose consequences. Since no new failure modes were created by the change in valve position, it was concluded that the actions and procedure changes identified in this safety evaluation did not constitute'n 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.
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SAFETY EVALUATION PTN-ENG-SEMS-96-078 REVISION 0 UNIT 3 APPROVAL DATE 12/09/96 SAFETY EVALUATION RELATED TO TSA No. 03-96-20-09 FOR REPAIR OF MOV-3-832
~SUEUIIR This safety evaluation assessed the impact on'lant safety and operation associated with a proposed temporary alteration of the primary water makeup and component cooling water (CCW) systems repair of the MOV-3-832,valve internals. Valve MOV-3-832. to'ermit isolates the primary water makeup line to the component cooling water (CCW) system-and provides the code boundary break between the Quality Group C (CCW) and Quality Group D (primary makeup water) piping. This evaluation addressed a) the use of a blind flange to temporarily maintain the primary water makeup system pressure boundary, and b) the temporary relocation of the Quality Group C code boundary to valves 3-711A (or 3-737C 3-711B. The evaluation addressed if 3-711A is closed) and the design and material qualifications of the blind bonnet flange for MOV-3-382. It also addressed the impact on plant operation associated with local manual operation of valves 3-711A (or 3-737C) and 3-711B for CCW makeup, versus control room operation of MOV-3-382. The evaluation concluded that the use of a field operator to initiate CCW makeup was acceptable given the maximum permissible leak rate for the system and the normal inventory stored in the surge tank. To satisfy the safety related isolation function, the evaluation
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required that a dedicated operator be stationed at the valves during the makeup process so that they could be quickly closed required by the control room.
if Safet Evaluation:
This evaluation addressed the temporary plant conditions and restrictions imposed on operation of the CCW system while valve MOV-3-382 was out of service. The evaluation also addressed issues, such as, seismic effects, failure modes and effects, technical specification requirements, isolation boundaries, and CCW makeup capabilities. The proposed actions did not adversely affect any safety related functions. The evaluation concluded that the maintenance could be performed, and that the temporary system alterations 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.
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SAFETY EVALUATION P TN- ENG- SEES - 9 6 - 0 8 1 REVIS ION 0 UNIT APPROVAL DATE : 12/13/96 SAFETY EVALUATION FOR TEMPORARY SYSTEM ALTERATION TSA NO. 4"96-003-16 FOR CONTAINMENT INSTRUMENT AIR BLEED VALVE CV-4-2826 CIRCUIT MODIFICATION
~Summa This safety evaluation assessed the impact on plant safety and operation associated with .a temporary modification of the containment instrument air bleed valve control circuit, to restore valve operability. The containment instrument air bleed valve (CV-4-2826) is part of- the containment isolation feature for penetration No. 63 and is normally open during plant operation to prevent containment 'ressurization due to the accumulation o f instrument air exhausting (i.e., bleeding) from pneumatically operated components inside the building. The temporary circuit modification was required to circumvent a faulty power cable in the CV-4-2826 solenoid valve control circuit. The modification consisted of abandoning the faulty cable in place and utilizing spare conductors in an adjacent cable associated with CV-4-2826 to provide the power feed. Since valve Cv-4-2826 is located in the pipe and valve room, all of the circuit modifications were performed in the auxiliary building. The evaluation considered the affects of postulated high energy line breaks, cable separation, and post-modification testing.
Safet Evaluation:
This evaluation examined the potential for new failure modes. It.
was concluded that the circuit changes did not alter the operation of the valve, its actuation logic, or interlocks. Design basis issues such as cable separation and the environmental qualification of cable splices were also addressed. The act'ions or plant changes in procedures, design documents, and/or hardware 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.
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tl SAPETY EVALUATION PTN-ENG-SEMS-97-006 REVISION 0 UNIT 3 & 4 APPROVAL DATE : 02/11/97 SAPETY EVALUATION TO ABANDON AUXILIARY STEAM IN THE AUXILIARYAND RADWASTE BUILDINGS
~Summa The Turkey Point auxiliary steam system was included as part of the original plant design to supply low pressure saturated steam to the auxiliary feedwater pump turbines, boric acid evaporators, waste disposal evaporators, boric acid batching tank, gas strippers, and control building unit heating. By the mid- 19 8 0s, much o f this equipment was no longer being used because the boric acid recycle and waste disposal processes proved too costly to operate and maintain. Engineering Packages were prepared to address the in-place abandonment of the boric acid recycle and waste disposal equipment, including isolation of the auxiliary steam system valves (see PC/Ms94-141 and 95-072 in Section I). This safety evaluation provides the justification to further abandon the auxiliary steam system desuperheater stations; condensate recovery transfer pumps, and remaining auxiliary steam components inside the auxiliary building and radwaste building. It was prepared to support Minor Engineering Package 95-081 which provides the implementing instructions necessary to mechanically and electrically isolate the above equipment. It primarily addresses the abandonment of auxiliary steam to the boric acid batching tank and the ability. to satisfactorily mix sodium tetraborate decahydrate (i.e., Borax) at temperatures below 55 F for post-LOCA chemical inj ection.
concludes that the technical specification requirement to maintain It the boric acid storage tank room temperature above 55 'F does not apply to Borax batching in the batching tank and provided a basis to reduce the minimum water temperature for batching post-LOCA chemicals to 39 'F.
The supply of auxiliary steam to the AFW pump turbines was not affected by the proposed changes.
Safet Evaluation:
The safety evaluation demonstrated that the auxiliary steam system did not perform any safety related functions and was not required to support safe shutdown of the plant. The in-place abandonment of this equipment had no adverse impact on plant safety or plant operations. Consequently, the actions or plant changes in procedures, design documents, and/or hardware 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.
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'APETY EVALUATION PTN-ENQ- SEIS 013 REVISION 0 UNIT 3 & 4 APPROVAL DATE : 02/27/97 SAPETY EVALUATION POR THE ABANDONMENT REPAIR OR RESTORATION OP RVLMS SENSORS
~Smnma This evaluation provided the basis for the acceptability of using Engineering Specification SPEC-IC-007 in the reactor vessel level monitoring system (RVLMS) sensor maintenance process, in lieu of the current practice, which requires that a Plant Change Mod'ification (PC/M) package be issued and implemented to evaluate the work scope and provide any'ecessary drawing revisions.
also demonstrated that the specif ication met all technical and licensing requirements for the Turkey Point Nuclear Units.
Engineering Specification SPEC'-IC-007 was developed to streamline the repair and abandonment process for defective sensors outside of the PC/M process. It allows Maintenance to abandon or repair defective sensors as well as restore presently abandoned or modified sensors upon installation of new RVLMS probes using the "Specification Clarification" process. Sensor abandonment is permitted if the technical specification requirement of, four or more operational sensors per channel exist before and after the activity. Sensor repair is permitted .
specification requirement will be met if following the technical the repair activity. The only repair activity permitted by the specification involves a.failed unheated thermocouple. In these cases, repair is accomplished by jumpering the failed unheated thermocouple to the nearest operable unheated thermocouple. This .repair .process is justified on the basis that there is not a substantial temperature difference between adjacent sensors. The specification also permits the wiring restoration of a given channel upon installation of a new RVLMS probe.
Safet Evaluation:
The RVLMS sensors provide a means for acquiring information about the core, but do not perform any type of control function. Since the number of remaining operable sensors continues to satis y
'o f technical specification requirements, there was safety significance associated with the abandonment, repair, or restoration process described in the generic specification. The provisions of the generic specification 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 engineering provisions of the generic specification identified within this evaluation.
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il SAFETY EVALUATION PTN- ENG - SEMS - 9 7 - 0 14 REVISION 1 UNIT 3 Ec 4 APPROVAL DATE 05/29/97 SAFETY EVALUATXON TO ADDRESS THE OPERATXON OF THE CONDENSATE POLXSHXNG DEMXNERALXZER SYSTEM
~summa This safety evaluation was prepared to update the UFSAR so that accurately reflects the current operational requirements of the it condensate polishing demineralizer system. The condensate polishing demineralizer system was originally designed to purify condensate from the condenser hotwell by filtration and demineralization to provide high quality feedwater to the steam generators. Operation and control of the system is independent from the existing condensate and feedwater system. The system was taken out of service to address a unit reliability issue. On two previous occasions, during unit,startup, while manipulating valves to place the system in service, a valve malfunction caused condensate to be diverted to the canal which resulted in a steam generator feedwater pump trip (low suction pressure), and the subsequent automatic initiation of auxiliary feedwater. As a result of these events, FPL has decided that the condensate polishing demineralizer system will only be used. prior to unit startups. The current FSAR description indicated that the system was still operation.
a normal part of the feedwater flow path during unit Revision 1 addressed PNSC comments related to the 50. 59 safety evaluation and proposed FSAR changes.
Safet Evaluation:
The effects of using the condensate polishing demineralizer system during unit startup, and isolating the ,system during plant operation, .was evaluated and determined not to have an adverse effect on plant safety or operation. The evaluation examined the potential impact on steam generator secondary side degradation, auxiliary feedwater system performance, and offsite doses due to steam generator tube rupture events. The proposed change in operating philosophy and resulting UFSAR 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.
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SAFETY EVALUATION aTPN- PTN- SEMS - 7 " 0 1 6 REVIS ION 0 UNIT 3 APPROVAL DATE : 03/18/97 SAFETY EVALUATION FOR INSTALLATION OF AN OIL DRAIN LINE ON THE 3B RCP MOTOR
~Summa This safety evaluation assessed the impact on plant safety and operation associated with the temporary installation of an oil drain line on the 3B reactor coolant pump (RCP) motor. The drain line was being installed to mitigate the effects of an oil leak at the flywheel seal on the upper reservoir. According to Westinghouse, the leak was most likely caused by air becoming entrained in the oil. It was speculated that the entrained air was being released under the flywheel seal area and forcing an oil mist or foam up through the seal. The .new drain line was intended to provide an additional vent to the upper, reservoir to relieve any pressure buildup under the flywheel seal. It would also direct any leaking oil to the oil collection system, away from any hot reactor coolant system piping. The drain line installation required 12 feet of 3/8-inch diameter stainless steel tubing.
Safet Evaluation:
The temporary drain line was designed to withstand all applicable loads, including seismic loads. The installation did not modify or actively interact with any plant equipment important to safety. An engineering review demonstrated that the applicable containment analyses were not affected by the materials of construction (i.e.,
metal) due to the small amount of material involved. The actions or 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.
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SAPETY EVALUATION PTN-ENG-SEIS-97-017 REVISION 0 UNZT 3 & 4 APPROVAL DATE 05/20/97 SAPETY EVALUATION RELATED TO THE ABANDONMENT 'OP MONITORING FUNCTION OP PARTICULATE DETECTORS OF THE SPING EPFLUENT MONITORS
~SllMS This safety evaluation assessed the acceptability of abandoning the on-line particulate monitoring function of the Special Particulate, Iodine and Noble Gas (SPING) monitors due to the high number of failures associated with Channel 1 (beta particulate detector) and, Channel 2 (alpha particulate detector) of the steam jet air ejector (SJAE) units. The SPING monitoring system was installed to satisfy the post-TMI regulatory requirements (NUREG 0737 and Regulatory Guide 1.97) for high range radioactive gas monitoring and iodine and particulate monitoring. In addition, the system provides a backup to the process radiation monitoring system and provides plant release data for Technical Specification Section 6.8. Plant operating history has shown that the SJAE SPING channel failures were primarily due to the effects of secondary system chemistry in the sample stream. The Channel 1 a' detectors are located immediately behind the particulate filter on the sample skid and are exposed to high concentrations of ammonia. These high concentrations of, ammonia cause corrosion and damage the mylar window of the scintillation detector. The SPING vendor, Eberline, does not currently make a detector that can survive the SJAE ammonia environment.
In accordance with plant technical specifications and the Off Site Dose Calculation Manual (ODCM), the required SPING channels are the noble gas channels (Channels 5, 7; and 9, and Channel 10 for the plant vent SPING). There are no requirements to maintain the particulate monitoring capabilities of Channels 1 8 2. Particulate monitoring is accomplished by collecting samples on filter media and analyzing them in the onsite chemistry laboratory.
Safet Evaluation:
The SPZNG monitors do not perform a safety related function. It was demonstrated that abandonment of Channels 1 8 2 would not adversely affect plant operation, the plant technical specifications, ODCM, or any regulatory commitments made pursuant to NUREG 0737 or Regulatory Guide 1.97. Consequently, the actions or plant changes in procedures, design documents, and/or, hardware 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.
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SAFETY EVALUATION PTN" ENG-SENS- 7-023 REVISION 0 UNIT 3 Zc 4 APPROVAL'ATE 06/13/97 SAPETY EVALUATION RELATED TO SEVERE ACCIDENT MANAGEMENT GUIDELINE IMPLEhQ"NTATION
~SlglM This safety evaluation was prepared to document FPL Engineering's review of the Turkey Point Severe Accident Guidelines to demonstrate consistency with the Westinghouse Owner's Group Severe Accident Management Guidance (SAMG) document prepared for member utilities. The need to addressstrategies, plant response to a severe for dealing with these accident, and develop appropriate "beyond design basis" events, was first addressed by the NRC in a commission position paper and later linked to the resolution of NRC Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities." In addition to the development of Probabalistic Safety Assessments for individual plants, the NRC desired the development of accident management strategies for severe accidents. In response to this aspect of the severe accident issue, the Westinghouse Owner's.Group developed the SAMG for use at member utilities. FPL subsequently used the SAMG to develop a set of Severe Accident Guidelines (SAGs) that are specific to Turkey Point. Transition steps from the Operating Procedures (EOPs) to the new SAGs were also existing'mergency developed and included in the Engineering review. The Engineering review concluded that the SAGs were consistent with the Westinghouse guidance document and that no changes to the UFSAR or DBDs were required. The evaluation identified several additional EOPs that were required to have transition points to the new SAGs.
Safet Evaluation:
Overall implementation of the SAGs was evaluated against the criteria of 10 CFR 50.59 due to its interfaces with the EOPs. The proposed EOP changes did not alter any of the strategies for coping with design basis events. The EOP transition points to the SAGs only occur after all required emergency actions have been completed and are unsuccessful. Consequently, the actions or 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 for implementation of the actions or changes identified within this evaluation.
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SAFETY EVALUATXON PTN-ENG-SENS-97-027 REVISION 0 UNIT 3 & 4 APPROVAL DATE: 02/28/97 SAFETY EVALUATION RELATED TO TEMPORARY ALTERATXON OF THE " C" GAS DECAY TANK SAMPLE 'ATH
~StlSEE This safety evaluation assessed the impact on plant safety and operation associated with a temporary alteration of the'C" gas decay tank sample path. A temporary alteration of the sample path.
was proposed by TSA No. 00-97-061-009 because a failure of valve PCV-1038B prevented the normal sample path from being used. The Turkey Point UFSAR requires that the contents of a gas decay tank sampled prior to release through the monitored plant vent stack.
'e The proposed sample path was from the inlet of PT-1038 to the upstream side of PCV1073B, which discharges to the waste gas analyzer sample header near the normal sample discharge point.
This sample path provided a direct indication of the stored activity in the "C" gas decay tank and was consistent with the normal sampling procedure. To prevent the accidental release of the gas decay tank contents in the proposed configuration, an additional valve was installed in the piping'between the temporary tie-in point at the inlet of PCV-1073B, and waste gas release valve RCV-014. The new valve was administratively -locked closed in compliance with UFSAR requirements. The tubing run was located within the shielded gas decay tank ~alve gallery so plant personnel working in the auxiliary building would not be sub j ected to abnormal dose levels during the temporary system alteration.
Safet Evaluation:
The gas decay tanks and associated discharge piping do not perform any safety related functions. Inadvertent release of the contents of a single tank is bounded by the UFSAR analysis of a gas decay tank rupture. In addition, the UFSAR commitment that two valves remain closed in series to provide a positive means of preventing an inadvertent release of a gas decay tank contents du'ring sampling remains intact during the temporary'ystem alteration.
Consequently, the actions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant'echnical specifications. Therefore, prior NRC approval was not required for implementation of =the actions or changes identified within this evaluation.
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SAFETY EVALUATION PTN-ENG- SEF J 03 0 REVISION 0 UNIT 3 R 4 APPROVAL DATE 09/02/97 10 QFR 50. 59 SAFETY EVALUATION " EVALUATION FOR PERFORMING THE ROD DROP TEST FROM ALL RODS OUT CONDITION
~Summa This safety evaluation assess the acceptability of performing the hot rod drop test from an all rods out (ARO) condition. This would reduce the amount of time spent performing low power physics tests after a refueling outage. The current test procedure measured the rod.drop times of two banks at a .time..Since there are a total of six banks in the Turkey Point, core, the test is repeated three times to gather data for all of the control rod banks. The proposed method of testing allows all six control rod banks to be tested at the same time. A review of the UFSAR and plant technical specifications indicated that there were no licensing commitments associated with a particular test sequence or particular test condition. To accommodate the proposed test procedure changes, contingency measures were established to ensure that adequate shutdown margin would be maintained during the ARO condition.
These included:
a) a requirement to borate the core to a boron concentration corresponding to a Keff of < 0.99 with uncertainty allowance; and b) a requirement to perform a 1/M plot at the time the control rods are being withdrawn to the ARO position during the test.
The safety evaluation examined the impact of proposed procedure changes on the plant safety analyses and provided a disposition for each UFSAR Chapter 14 event.
Safet Evaluation:
The safety evaluation demonstrated that the proposed changes in rod drop time testing did not impact the plant technical specifications or any UFSAR safety analyses. Restrictions were imposed to ensure that the minimum required technical specification shutdown margin is maintained during rod withdrawal to the ARO position. Based on the precautions identified, the evaluation concluded that the rod drop test could be performed prior to criticality with Tavg a 541 F (considered as Mode 3), 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 activities identified within this evaluation.
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0'l SAFETY EVALUATION PTN" ENG- SEMS 033 REVISION 0 UNIT 3 APPROVAL DATE : 03/28/97 SAFETY EVALUATION FOR UNIT 3 REACTOR CORE FOREIGN OBJECT
~StBtlER This safety evaluation was prepared by Westinghouse Electric Corporation to assess the impact on plant safety and operation associated with a loose nylon cable tie in the reactor coolant system. The cable tie came apart while restraining temporary underwater'amera and light cables during the refueling outage.
The cable tie originally fell to the top of a fuel assembly, but was subsequently lost during retrieval efforts. It was assumed that the missing cable tie was made of white nylon 6/6 based on the material of adjacent cable ties on the camera and light rig. The safety evaluation provided an assessment of the potential impact of the unrecovered cable tie 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.
Safet Evaluation:
The nylon cable tie is expected to soften during heatup and completely melt prior to reaching the normal RCS operating temperature. It is further expected that the melted nylon material will be dispersed into the RCS and deposited on cooler surfaces under low flow conditions. The safety evaluation demonstrated that this phenomenon would not affect the operability or integrity of the reactor fuel, RCS, or the interfacing auxiliary systems. In addition, the decomposition of the nylon 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 single nylon cable tie 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.
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SAFETY EVALUATION PTN-ENG-SENS- 7-038 REV1SION 0 UNIT 3 & 4 APPROVAL DATE 04/07/97 APETY EVALUATION FOR RREEZE SEALS ON SEAL TABLE THIMBLE GUIDE TUBE
~summa This safety evaluati'on was prepared to assess the performance and use of freeze seals when conducting repairs on the bottom mounted in-core thimble guide tubes associated with the plant seal table..
It was generated in response to an identified through-wall leak in the thimble guide tube associated with core location H-1. To avoid plant operation in a reduced inventory .condition, it was proposed that two freeze seals be placed on the thimble guide tube to be repaired to allow reactor coolant system (RCS) inventory to remain in the range of the pressurizer. The .activity of placing a freeze seal on a guide tube has previously been evaluated for Turkey Point and found acceptable by Westinghouse Electric Corporation. The intent of this safety evaluation was to review that evaluation and reconfirm its validity for this application. Although the review concluded that the activities proposed for guide tube H-1 were essentially identical to those previously evaluated, implementation of the proposed freeze seals was re-evaluated under the criteria of 10 CFR 50..59.
Safet Evaluation:
This evaluation addressed the consequences of a freeze plug failure (both seals) and demonstrated that sufficient makeup capability could be provided to= ensure accomplishment of the decay heat removal function. It was also concluded .that adequate precautions have been included in the repair procedure to preclude vertical movement of the thimble tube while the freeze plug is intact.
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.
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SAP ETY EVALUATION P TN"ENG - SENS " 9 7 - 04 0 REVISION 0 UNIT 3 Ec 4 APPROVAL DATE : 04/29/97 SAPETY EVALUATION RELATED TO STEAM GENERATOR BLOWDOWN CONTAINMENT ISOLATION PEATURES
~Summa This safety evaluation was prepared to clarify the function of the steam generator blowdown isolation valves and blowdown sample valves as described in the UFSAR relative to containment was prepared in response to an operator discovery that the Unit, isolation.'t 3 steam generator blowdown system interlock bypass switches were left in the drain/fill position when the unit entered Mode 4 from a recent refueling outage (LER 250/97-003). The drain/fill position blocks Phase A, main steam isolation, and auxiliary feedwater actuation signals to the main blowdown isolation valves CV-3-6275 A, B, and C, and blowdown sampling valves MOV-3-1425, 1426, and 1427, such that these valves are open and will not close automatically. The evaluation demonstrated that the valves were not containment isolation valves and not subject to the containment isolation technical specification requirements. It was established that the primary function of these valves was to close to support operation of the auxiliary feedwater system, which is required in Modes 1 through 3. The design basis clarification was supported, with statements, taken from original Westinghouse Electric Corporation design information, and a detailed analysis of the overall secondary system isolation philosophy implemented at Turkey Point. The proposed UFSAR changes were provided as Attachment 1 to the evaluation Safet Evaluation:
The safety evaluation addressed the impact of the mispositioned keylock switches on the plant safety analyses for Mode 4 events.
It was demonstrated that sufficient contingency actions were included in the emergency operating procedures to ensure that manual closure of 'the blowdown valves would occur associated keylock switches were in the "override" position. The if their evaluation also demonstrated that leaving the subject valves open during plant operation would not invalidate the UFSAR single active failure criterion for containment isolation. It was concluded that
'he change in interpretation would not reduce the level of protection provided against the release of radioactivity to the outside environment because the subject valves continue to receive an automatic phase A closure signal. Consequently, the changes identified in this safety evaluation did not constitute an safety question or require changes to the plant 'nreviewed technical specifications. Therefore, prior NRC approval was not required for implementation. of the proposed document changes.
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SAFETY EVALUATION PTN-ENG-SEIS-97-041 REVISION 0 UNIT 3 & 4 APPROVAL DATE 06/27/97 SAFETY EVALUATION FOR CORE EXIT THERMOCOUPLE ABANDONMENT OR REPLACEMENT
~Summa This safety evaluation was developed to establish a generic set of guidelines that could be used by Maintenance personnel to abandon or restore core exit thermocouples (CETs), vi'a the plant abandoned equipment program. This .would allow inoperable CETs to be abandoned in place (or restored) without performing repetitive engineering evaluations. The CETs are part of the inadequate'core cooling system (ICCS) instrumentation which is 'designed to yield information on fuel assembly outlet temperatures at selected core locations. This information is used to confirm that reactor core design parameters are within analyzed limits. The number of operable CETs per core quadrant is governed by plant technical specifications. In keeping .with the applicable requirements, this safety evaluation permitted the abandonment of a given CET if, after the abandonment, at least 2 operable CETs per channel per core quadrant (4 total per quadrant) remained in service. Any abandonment activity that would result in fewer'han 2 operable CETs on a given channel and core quadrant was not permitted by evaluation, even though a minimum of 1 operable CET on a the'afety given channel and core quadrant was permitted by plant technical specifications.
Safet Evaluation:
The CETs provide a means for acquiring information about the core, but do not perform any type of control function. Since the number of remaining operable CETs continued to satisfy technical specification requirements, there was no safety significance to the proposed abandonment guidelines'he actions or plant conditions identified in this safety evaluation did not constitute an unreviewed safety question or require changes to the plant technical specifications. Therefore, prior NRC approval was not required for implementation of the, actions or conditions identified within this evaluation.
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SAPETY EVALUATION PTN-ENG-SECS-97-051 REVISION 0 UNIT 4 APPROVAL DATE 04/25/97 10 CPR '0. 59 SAPETY EVALUATION POR THE PLACEMENT OP L'EAD SHIELDING ON VALVE LCV-4-460 AND SHIELD WALL BETWEEN VALVE LCV-4-460 AND REGEN HEAT EXCHANGER
~Summa This safety evaluation addressed the temporary installation of lead shielding inside containment to permit in-situ repair of the regenerative heat exchanger level control valve, LCV-4-460. The purpose of the evaluation was to demonstrate that the temporary
,shielding would not adversely affect plant safety or operation erected in Mode 3 or below. The evaluation examined the potential if for adverse seismic interactions between the lead shielding and adjacent safety related equipment, the increased pipe stresses caused by the placement of lead blankets on valve LCV-4-460, the potential impact on containment isolation for penetration No. 19, and the potential impact on containment sump operation during postulated reactor coolant system pipe rupture events.
Safet Evaluation:
The temporary shielding covered by the scope of this safety evaluation was designed to withstand all applicable loads, including seismic loads. The lead blankets and temporary support structure did not modify or actively interact'ith any plant equipment important to safety. An engineering review demonstrated that containment isolation and containment sump operation would not be affected by the shielding material or the scaffolding support structure. The 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 to implement the actions identified within this evaluation.
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SAFETY EVALUATION PTN'-ENG- SENS-97-054 REVIS ION 0 UNIT 3 APPROVAL DATE : 05/09/97 SAFETY EVALUATION RELATED TO TEMPORARY REMOVAL OF THE 3A CCW HEAT EXCHANGER CHANNEL HEADS AND TEMPORARY SUPPORT INSTALLATION
~Sl13DIBK This safety evaluation assessed the impact on plant safety and operation associated with the temporary removal of the 3A component cooling water (CCW) heat exchanger channel heads for maintenance access to the tube sheets. Access to the inlet and outlet tube sheets was necessary for maintenance personnel to retube the heat exchanger. The CCW heat exchanger inlet and outlet channel heads provide an anchor point for the seismic analysis of the ICW system.
A stress analysis of the system without these anchor points determined that the ICW/CCW system would remain operable provided that a temporary support was added to the 3B CCW heat exchanger support pedestal. Removing the channel heads from the heat exchanger shell required that a heavy load lift be performed in a restricted area. To minimize the impact on neighboring equipment, provisions were included in the safety evaluation to lower the channel heads straight down to the 18'levation where there was no safety related equipment in the vicinity. A review of the Turkey Point UFSAR and plant technical specifications indicated that the CCW system is capable of accomplishing its safety related function with two of the three heat exchangers in service. Thus, plant operation with one CCW heat exchanger out of service for maintenance has been previously reviewed and evaluated.
Safet Evaluation:
The ability to remove one of the three CCW heat exchangers from service .for repair or replacement is part of the system design basis. The requirement to install supplemental restraints on the 3A support pedestal preserved the seismic qualification of the system during the maintenance evolution, and ensured continued operability of the 3B and 3C heat exchangers. Consequently, the actions and 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 and plant changes identified in this safety evaluation.
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SAFETY EVALUATION PTN-ENG-SENS-97-058 REVISION 0 UNIT 3 & 4 APPROVAL DATE 07/24/97 SAFETY EVALUATION FOR REPEALING THE F IRE BARRIER RE UIRE2KEKTS ASS XGNED TO PENETRATION No. 041E-E001 IN THE EAST WALL OF THE BORIC ACID STORAGE TANK ROOM FIRE AREA M 41
~Sl1EGR This safety evaluation addressed the 10 CFR 50.59 .criteria for repealing the fire barrier requirements assigned to penetration No.
041E-E001 in the east wall of the Units 3&4 boric acid storage tank room'. The'ast wall is an external wall which separates the boric acid batching equipment from the outside radiation control area yard and the Unit 3 component cooling water (CCW) equipment area.
The 4-inch diameter wall penetration is routinely used during refueling outages to refill the boric acid storage tanks from an external supply. Plant. drawings and field walkdowns were used to demonstrate that adequate spatial separation existed between the boric acid storage tanks and the r'efueling water storage tanks (the redundant water source for. safe shutdown) to ensure that one borated water source would be available for cooldown during a fire.
Administrative controls also existed to limit. the presence on intervening combustibles between the two water sources. Several contingency actions were imposed by the safety evaluation to preclude spurious actuation of the 3C/4C ICW and CCW pumps due to.
a fire in the boric acid storage tank room. Cables associated with these pumps are routed in close proximity to the exterior wall penetration. The repeal of the fire barrier requirement permits the wall penetration to be used without having to install a temporary fire seal, or having to post a fire watch.
eliminates the need to perform periodic fire barrier related It also inspections of the penetration sleeve during plant operation. The proposed UFSAR changes were. provided in Attachment 1 of the safety evaluation.
Safet Evaluation:
The proposed fire 'barrier changes did not have any safety significance because safe shutdown during fire scenarios is not a safety related function. Consequently, the actions and 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 and document changes identified within this evaluation.
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SAFETY EVALUATION PTN-ENG-SEIS-97-059 REVISION 0 UNIT 4 APPROVAL DATE 07/07/97 SAFETY EVALUATION- FOR TEMPORARY INSTALLATION OF A RECORDER ON RCS FLOW LOOP F-4-426 TSA 04-97-049-3
~Summa This safety evaluation assessed the impact on plant safety and operation associated with the temporary installation of a recorder on reactor coolant system (RCS) flow loop F-4-426 for channel troubleshooting purposes. The recorder was installed to monitor the flow comparator input and'output signals and the loop power supply, and identify any abnormalities. The recorder was placed on a rubber mat on the floor of the control room, adjacent to rack 4QR15. The evaluation addressed the potential for both*electrical and seismic interactions with other safety systems. The recorder was evaluated as a potential missile during a seismic event and determined not represent a hazard for other safety systems. The following electrical failures were also considered: a) loss of power to the recorder, b) shorting of any pair of input leads to each other, c) shorting of any single input lead to any other, d) grounding of any input lead, e) opening of any input lead, and f) incorrect hookup to any in proximity terminal. The failure modes and effects analysis concluded that the temporary installation would not introduce any new failure modes for the flow channel being monitored. Installation of the recorder was limited to a specific period in time after which requirements of the evaluation.
it was removed under the Safet Evaluation:
The evaluation concluded that the installation of the temporary monitoring recorder would have no adverse impact on plant safety or operation, and would not have compromised the safety or licensing requirements for Unit 4. Its installation was also limited to a specific period in time and was to be removed after data was obtained. Consequently, installation of this temporary monitoring recorder, as. discussed 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 installation and use of the temporary monitoring recorder.
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SAFETY EVALUATION PTN-ENG-SECS-97-061 REVISION 0 UNIT 3 Ec 4 APPROVAL DATE 09/04/97
'AFETY EVALUATION FOR CONTAINMhKT POLAR CRANE MAINTENANCE INSPECTION PROCEDURE
~Summa This safety evaluation assessed the acceptability of spreading the required containment polar crane maintenance inspections over the course an outage to reduce critical path time. The of performing all of the required crane inspections at the. current'ractice beginning of each outage delays many of the critical path activities. As a basis for establishing the new i'nspection 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 c'onsidered 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.
Safet 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, pri'or NRC approval was not requ'ired for impleme'ntation of the phased polar crane inspection plan.
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SAFETY EVALUATION PTN-ENG- EES-97-062 REVISION 0 UNIT 3 & 4 APPROVAL DATE : .08/06/97
- SAFETY EVALUATION FOR SURVEILLANCE RE UIRlKENTS FOR APPENDIX R CIRCUITS AT 4 . 16 KV SWITCHGEAR AND 480 V LOADCENTER BUSSES
~summa This safety evaluation assessed the impact on plant safety and operation associated with extending the surveillance requirement for Appendix R circuits at the 4160 V switchgear and 480 V load centers from every refueling outage to every other refueling outage. The Appendix R circuits at the switchgear and load centers consist of transfer switches, breaker control test switches, isolation switches, and redundant fuses. The transfer switches, test switches, and isolation switches used at Turkey Point are either General Electric Type SB-1 switches or Electro-Switch Type 24 switches. A review of the failure history of these switches was performed over the period between July 1990 and July 1997. The review demonstrated that the switches are not subject to any time dependent failure modes. In addition, the subject switches are maintained in a controlled environment which minimizes the probability of contact corrosion and dust accumulation.
Safet Evaluation=
The circuits, switches, and fuses associated with Appendix R do not have any safety significance because safe shutdown during fire scenarios is not a safety related function. Changing the surveillance test frequency of these devices did not adversely impact operation of the plant safety related buses. Consequently, the 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 for implementation of the new surveillance test frequency.
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SAPETY EVALUATION PTN- ENG - SEMS 0 65 REVISION -0 UNIT 3 Ec 4 APPROVAL DATE 07/29/97 APETY EVALUATION TO DOCUMENT THE CHANGE POR AUXILIARYPEEDWATER SYSTEM STEAM'RAPS ST-33 ST-34 AND ST-35 DRAINING INTO THE 'ADJACENT TROUGH IN LIEU OP THE EXISTING CONDENSER DRAIN-PATH
~Sl15UIIR This safety evaluation was prepared to assess the acceptability of rerouting the ST-33, ST-34, and ST-35 steam trap discharge piping from the existing condenser drain path to an adjacent drain trough.
The proposed change was implemented to eliminate potential air in-leakage to the condenser through the steam trap seats. The safety evaluation demonstrated that adequate drain 'flow would be maintained by the modified piping arrangement. The required UFSAR changes were included as an attachment to the safety evaluation.
Safet Evaluation:
The design change did not adversely impact .operation of the auxiliary feedwater system (AFW), access to AFW components, or habitability in the AFW equipment area. Consequently, the plant changes and 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.
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SAFETY EVALUATION PTN-ENG" SENS- 7-066 REVZS ION 1 UNIT 3 Sc 4 APPROVAL DATE 08/21/97 SAFETY EVALUATION FOR AUGN3MTED RCP OZL COLLECTION SYSTEM RE UIREMENTS
~SUIQKK This safety evaluation addressed the 10 CFR 50.59 criteria for the proposed reactor coolant pump (RCP) oil collection system changes inside containment. The proposed changes extended the coverage of the existing oil collection system to include a) the upper and lower oil reservoir level switch assemblies and associated piping flanges and dra'ins, b) the oil cooler piping drain, c) the upper oil reservoir drain, and d) the two flexible oil connections currently being installed by PC/M 97-016.
fill line hose These changes were needed to comply with the 10 CFR 50 Appendix R requirement that all potential pressurized and unpressurized leakage sites be provided with collection facilities. Only those potential leak sites located below the normal reservoir,oil level were considered within the scope of the modification. The fill safety evaluation required that new drip pans and drain lines be installed to collect all dripping and stream- flow type leaks postulated to occur from the identified sources. The new oil collection facilities were required to be seismically installed to prevent .adverse interactions with adjacent safety related equipment. Zn addition, drip pans were required to be mounted to the RCP motor casings at locations that did not provide a fluid or lube oil pressure boundary function; did not house sensitive motor components; and did not compromise electrical insulation requirements of the motor housing conf iguration. The proposed UFSAR changes were provided as Attachment 1.
Revision 1 documented the Westinghouse position that the main oil reservoir flange on the motor housing was not considered to be a potential source of lube oil leakage.
Safet Evaluation:
The RCP oil collection function is a plant fire protection feature and has no safety significapce. The additional drip pans and drain lines required to augment the existing oil collection facilities were determined to have a negligible impact on the containment heat sink analysis. Consequently, the actions and documentation 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 and document changes identified within this evaluation.
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SAFETY EVALUATION PTN-ENG-SENS-97-076 REVIS ION 0 UNIT APPROVAL DATE : 09/08/97 SAFETY EVALUATION FOR TEMPORARY ALTERATION OF CONTAINMENT PURGE VALVE POV-4-2600 CONTROL AND INDICATION CIRCUIT WIRING
~SllSCOS This safety evaluation assessed the impact on plant safety and operation associated with a temporary wiring change in the containment purge valve POV-4-2600 control and indication circuit,.
to restore valve operability. The temporary wiring change was necessary to circumvent a damaged cable in the valve control circuitry and permit operation of the containment purge system in preparation for an impending unit shutdown and refueling outage.
The modification consisted of abandoning the faulty cable in place and rewiring valve POV-4-2600 to the control circuit of valve POV-4-2602. Valve POV-4-2600 is the outboard containment isolation valve for the purge supply line. Valve POV-4-2602 is the outboard containment isolation valve for the purge exhaust line. The circuits were rewired such that both valves could be operated with the POV-4-2602 control switch. The 'open and close position indications for POV-4-2602 were also rewired'n series with POV 2600 to maintain position indication for POV-4-2600 in compliance with Regulatory Guide 1. 97 commitments. In keeping with these changes, the indication for POV-4-2602 represented the position of both valves, POV-4-2600 and POV-4-2602. The evaluation addressed the'appli'cable containment isolation design bases, the potential for new failure modes, the continuous load rating of the POV-4-2602 control switch, and the impact of the indicating lamp changes on IST testing.
A tabulation of the various indicating light combinations relative to valve POV-4-2600 and POV-4-2602 positions was provided as an attachment to the evaluation for'perator training purposes.
C Safet Evaluation:
This evaluation examined the potential for new failure modes. It was concluded that the circuit changes did not alter the method of isolating the containment purge lines during an accident, or retard the closing speed of the containment purge valves.
. The UFSAR commitment that containment isolation be established assuming a single active failure remained intact with the modified design.
Consequently, the 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 circuit changes.
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SAFETY EVALUATION PTN-ENG- SECS 077 REVISIONS 0 AND 1 UNIT APPROVAL DATE : Rev.0 10/02/97 Rev.1 10/10/97 EVALUATION FOR STORAGE OF TOOLS AND E UIPMENT IN CONTAINMENT DURING ALL MODES OF OPERATION
~Summa 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 within evaluation. The purpose of containment following leaving these tools and equipment'the refueling outages was to reduce 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 addressed in the evaluation were safely stored during plant operation, both generic and specific actions and restrictions were identified for implementation within the evaluation.
Revision 1 addressed the storage of some additional items inside containment.
Safet 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 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.
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SAFETY EVALUATION PTN-ENG-SEMS- 97- 078 REVISION 0 UNIT 3 & 4 APPROVAL DATE 10/14/97 SAFETY EVALUATION FOR CONTINUOUS PIRE WATCH
~Summa This safety evaluation was prepared to assess the acceptability of performing a continuous fire watch via a 15 minute roving patrol, and to incorporate this definition of a continuous fire watch into.
the UFSAR so that it accurately reflects the Turkey Point fire protection program. Currently, there is no regulatory definition of a fire watch and none of the NRC fire protection'uidance documents provide'n explanation of acceptable implementation of a continuous fire watch. The definition established in the safety evaluation requires that a trained individual be in the specified area at all times, that the specified area contain no impediment to restrict the movements of the continuous fire watch, and that each location within a specified area be patrolled at least once every 15 minutes with a margin of 5 minutes. In keeping with these restrictions, the safety evaluation juptifies that the implementation of a continuous fire watch via 15 minute roving patrols is an effect'ive utilization of manpower, provides a level of protection that is commensurate with the impaired protective feature, and is. capable of detecting a fire before it beyond the incipient stage sufficient to cause damage which might develops affect the ability to achieve and maintain safe shutdown conditions. The proposed UFSAR changes were provided as an attachment to the safety evaluation.
Safet Evaluation:
The implementation of a continuous fire watch does not have any safety significance because safe shutdown during fire scenarios i;s not a safety related function. "
Consequently, the actions and 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 and changes identified within this evaluation..-
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SAFETY EVALUATION PTN" ENG-SENS"97-084 REVISION 0 UNIT 3 R 4 APPROVAL DATE . 10/02/97 SAFETY EVALUATION FOR CONTAINMENT SUMP SCREEN DESIGN RE UIRI2KENTS
~Summa This safety evaluation was prepared to augment the documentation of design basis and performance requirements for the containment sump screens, to provide a more mechanistic basis for assessing sump screen operability, to quantify the ECCS heat sink and hydrogen generating material bulk inventory contribution of the sump screens and to provide for clarifying the UFSAR accordingly. This effort was prompted by the discovery of a hole and various gaps in the containment sump screens during the Unit 4 Cycle 17 refueling outage inspection activities, and during a Unit 3 containment entry at power. The conditions were determined to be outside the Turkey Point design basis as stated in the UFSAR and immediately reported to the NRC (LER 250/97-008);
.The evaluation developed a mechanistic basis for assessing Turkey Point sump screen operability by determining the types and quantities of debris likely to be generated and transported to the screens as a result of a. design basis accident. The methodology determined the areas of exposure to a credible pipe break and characterized the potential debris based on reviewing design and procurement documents and inspecting as-built installations in the field. Appropriate revisions to the UFSAR were provided in to the safety evaluation. Attachment 2 provided an analysis of sump screen debris and emergency core cooling system performance relative to the design of the debris resistant fuel assemblies install'ed in the Turkey point cores. Attachment 3 provided an independent assessment of the likelihood of debris generated by a pipe break in the Turkey Point containment bypassing the as-found containment sump screens.
Safet Evaluation:
The evaluation and supporting analyses demonstrated that the extent of screen clogging described in the UFSAR is non-mechanistic, not credible and extremely conservative based on the type and quantity of debris likely to be generated. It was concluded that the proposed UFSAR'larification, and associated augmentation of screen inspection requirements, did not alter the design or operation of any safety related equipment. 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.
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SAFETY EVALUATION SECL-97"200 REVISION 0 UNIT 4 APPROVAL DATE 09/25/97 REACTOR COOLANT SYSTEM LOOSE PARTS EVALUATION
~Summa This safety evaluation was prepared by Westinghouse Electric Corporation to assess the impact on plant safety and 'operation associated with two loose cap screws and locking cups in the reactor coolant system. The cap screws and locking cups were found missing during an inspection of the 4B reactor coolant pump (RCp) diffuser adapter plate during the last Unit 4 refueling outage.
For the purposes of analysis, it was assumed that each RCP lost two cap screws and locking cups in the reactor coolant system. The safety evaluation provided an assessment of the potential impact of the unrecovered parts on the operability and integrity of the fuel, reactor vessel, reactor internals, pressurizer, reactor coolant pump, steam generator, auxiliary equipment, or thermowells during future operating cycles.
Safet Evaluation=
An evaluation of equipment important to safety was performed with the loose parts .present in the RCS and it continued operation of the plant was acceptable.
was determined that All credible scenarios of migration, lodging, or impacting of the loose objects were considered. It was also demonstrated that continued operation of an RCP with two diffuser adaptor cap screws, missing would not adversely impact the design or performance of the RCP. Based on this assessment, the evaluation determined that the presence of up to six cap screws and six locking cups 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.
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SECTXON 3 RELOAD SAPETY EVALUATXONS 92
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PLANT CHANGE MODIFICATION 96-024 UNIT 3 TURN OVER DATE : 11/13/96 TURKEY POINT UNIT 3 CYCLE 15 RELOAD DESIGN THERMAL POWER UPRATE
~Summa This Engineering Package provided the safety evaluation, instructions, and data necessary to operate the Unit 3 Cycle 15 core up to 2300 MWt, as part of the thermal power uprate project.
The only design change for the Cycle 15 core was the change in thermal design power from 2200 MWt to 2300 MWt. The normal Tavg remained at 574.2 'F. The data and instructions provided included:
a) INCORE-3D detector constants needed to monitor the incore power distribution for compliance with the limits specified in the Core Operating Limits Report (COLR);.
b) operational data support'ing expected plant evolutions throughout the cycle, .including spent fuel pool heatup rates and reactor core decay heat generation; and c) COLR parameters for the ax'ial flux difference operating band, K(z) curve, rod insertion limits, and peaking factor Limits.
Key safety parameters associated with Unit 3 Cycle 15 thermal uprate were provided as an attachment to the Engineering Package.
Associated UFSAR changes were provided as part of PC/M 96-022, Revision 0.
Safet Evaluation:
The Unit 3 Cycle 15 thermal uprate core design was evaluated by FPL and by the fuel supplier, Westinghouse Electric Corporation, using NRC approved methodology. The uprated core met all applicable design criteria and.pertinent licensing bases. The reload analyses demonstrated that operation at the higher power level would not exceed any core design criteria, nor cause the core to operate in excess of pertinent design basis operating limits for the key safety parameters. Demonstrated adherence to applicable standards and acceptance criteria precludes new risks to components and systems. Since provisions for power escalation to uprate conditions and associated documentation changes have no adverse affect .on plant safety, security, or ,operation, the changes addressed by this Engineering Package did not involve an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the Cycle 15 (thermal uprate) core reload.
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PLANT CHANGE MODZFZCATZON 9 6- 02 5 UNIT ~
4 TURN OVER DATE : 11/13/96 TURKEY POZNT UNZT 4 CYCLE 16 RELOAD DESZGN THERMAL POWER UPRATE
~summa This Engineering Package provided the safety evaluation, instructions, and data necessary to operate the Unit 4 Cycle 16 core up to 2300 MWt, as part of the thermal power uprate project.
The only design change for the Cycle 16 core was the change in thermal design power from 2200 MWt to 2300 MWt. The normal Tavg remained at 574.2 'F. The data and instructions provided included:
a) INCORE-3D detector constants needed to monitor the incore power distribution for compliance with the limits specified in the Core Operating Limits Report (COLR);
b) operational data supporting expected plant evolutions throughout the cycle, including Spent Fuel Pool heatup rates and reactor core decay heat generation; and c) COLR parameters for the axial flux difference operating band, K(z) curve,'od insertion limits, and peaking factor limits.
Key safety parameters associated with Unit 4 Cycle 16 thermal uprate were provided as an, attachment the Engineering Package.
Associated UFSAR changes were provided as part of PC/M 96-022, Revision 0.
Safet Evaluation:
The Unit 4 Cycle 16 thermal uprate core design was evaluated by FPL and by the fuel supplier, Westinghouse Electric Corporation, using NRC approved methodology. The uprated core met all applicable design criteria and pertinent licensing bases. The reload analyses demonstrated that operation at. the higher power level would not exceed any core desi'gn criteria, nor cause the core to operate in excess of pertinent design basis operating limits for the key safety parameters. Demonstrated adherence to applicable standards and acceptance criteria precludes new risks to components and systems. Since provisions 'or power escalation to uprate conditions 'and associated documentation changes have no adverse affect on plant. safety, security, or operation, the changes addressed by this Engineering Package did not involve an unreviewed safety question or require changes to plant technical specifications. Therefore, prior NRC approval was not required for implementation of the Cycle 16 (thermal uprate) core reload:
94
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PLANT CHANGE MODIFICATION 96 - 07 I UNIT 3 TURN OVER DATE 05/07/97 TURKEY POINT UNIT 3 CYCLE 16 RELOAD DESIGN
~SlMIR This Engineering Package provided the reload core design for the Turkey Point Unit 3 Cycle 16 reload. The pr'imary design change,to.
the core for Cycle 16 was the replacement of 60 .irradiated assemblies with 60 fresh Optimized Fuel Assembly (OFA) Region assemblies. Similar to past reloads, these fresh assemblies 18'uel were all Debris Resistant Fuel Assemblies (DRFA) and all contained a nominal 6-inch axial blanket of natural U02 pellets at both the top and bottom of the fuel stack. The Cycle 16 core was designed to operate at the uprated power level of 2300 MWt. The maximum fuel enrichment was 4.2 w/o which was the first use of fuel with an enrichment above 4.0 w/o at Turkey Point.
Region 18 used the same Debris Resistant Fuel Assembly (DRFA) design as the prior Region 17, except that a Composite Top Nozzle (CAST) assembly was used on the new bundles .to reduce component parts, and grooved end plugs were used on the new fuel rods to improve end plug welding. Neither of these -manufacturing-related design changes had any impact on fuel performance.
Cross core fuel bundle shuffles were utilized in the Cycle 16 loading pattern; these shuffles were adequate to minimize potential power asymmetries. The fuel was.arranged in a low leakage pattern with no significant differences between the 'Cycle 15 and Cycle 16 loading patterns.
Safet Evaluation:
The Unit 3 Cycle 16 reload core design was evaluated by FPL and by the fuel supplier, Westinghouse Electric Corporation. The Cycle 16 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 16 core reload did not have any adverse effect on plant safety or plant operations. Consequently, the Cycle 16 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.
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PLANT CHANGE MODIFICATION 97 - 014 UNIT ~ ~
4 TURN OVER DATE 10/22/97 TURKEY POINT UNIT 4 CYCLE 17'ELOAD DESIGN
~SllEER This Engineering Package provided the Turkey Point Unit 4 Cycle 17 reload core design. The primary design change for Cycle 17 was the replacement of 56 irradiated assemblies with 56 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 UO2 pellets at both the top and bottom of the fuel stack. The Cycle 17 core was designed to operate at the uprated power level of 2300 MWt. The maximum fuel enrichment was 4.2 w/o which was the first use,of fuel with an enrichment above 4.0 w/o in the Unit 4 core. A Composite Top Nozzle (CAST) assembly was included on the new region 19 bundles to reduce component parts, and a .standardized top grid bulge location was established to reduce fuel manufacturing costs. Grooved end plugs were also used on the new fuel rods to improve end plug welding.
Cross core fuel bundle shuffles were utilized in the Cycle 17 loading pattern; these shuffles were. adequate 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 loading patterns. The plant safety analyses for Cycle 17 incorporated a) the pressure losses in the steam piping between the steam generators and the main steam safety valves, and b) an increased time delay in the turbine pressure signal to the automatic rod control system.
Safet Evaluation:
The Unit 4 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 accident reanalysis demonstrated that the safety limits continue to be met for Cycle 17. 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 of the Cycle 16 core reload.
96
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SECTION 4 REPORT OP POWER OPERATED RELIEF VALVE (PORV) ACTUATIONS 97
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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 ' letter of May 7, 1980 (Five Additional TMI-2 Related Requirements for Operating Reactors).
Two power operated relief valve (PORV) challenges occurred for the Turkey Point Plant Units 3 and 4 during the period from April 8, 1996 through October 13, 1997.
Unit 3 actuation occurred when the 3A reactor coolant pump was A PORV started for a one minute run during the reactor coolant system and vent process at the end of the Cycle 16 refueling outage.
fill The PORVs" were operating in the overpresure mitigating system (OMS) mode when the actuation occurred. The OMS operated as designed with PORV (or PORVs) lifting at about 415 psig. A special report for this event was submitted to the NRC Regional Administrator (Region II) under FPL letter L-97-102.
Unit 4 A PORV actuation occurred on April 23, 1997 during a reactor trip from 100% power. Two PORVs lifted during the transient at 2335 pslg.
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SECTION 5 STEAM GENERATOR TUBE INSPECTIONS POR TURKEY POINT 99
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CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT P' 09/97 COMPONENT : S/G A Page Date : 09/24 Examination Dates: 09/18/97 thru 09/22/97 Total Number of Tubes Znspected .....: 3198 Total Indications Between 204 and 394 0 Greater than or equal to 404 0 Total Corrosion type Indications "VOL" 0 Total Tubes Plugged 'as Preventive Maint 0 Total Tubes Plugged 0.
Location 'Of Indications 204 to 1004 & "VOL" Hot Leg Cold Leg TSH .5 to 01H -2.1 : 0 TSC .5 to 01C -2.1 0 01H -2.0 to 06H +2.,0 : 0 01C -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: 0 101
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CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT I 4 COMPONENT : S/G B 09/97'age Date : 09/
Examination Dates : 09/18/97 thru 09/22/97 Total Number of Tubes Inspected 3206 Total Indications Between 204 and 394 2
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Greater than or equal to 404 0 Total Corrosion type Indications "VOL" 0 Total Tubes Plugged as Preventive Maint -: 0 Total Tubes Plugged 0 Location Of Indications 204 to 1004 & "VOL" Hot Leg Cold Leg TSH ..5 to 01H -2.1 : 0 TSC -;5 to 01C -2.1 0 01H -2.0 to 06H +2.0 : 0 01C -2.0 to .06C +2.0 0 06H +2.1 to AV1 -3.1 : 0 '06C +2. 1 to AV4 -3. 1
- '2 0
AV1 -3.0 to AV4 -3.0 102
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TXOrCATiOXS/TREHOrXC REPORT PTX-C OUTACE : 09/97 CCHPOXENT : S/0 B Page :
OESCRIPTlOX t 20K to 39X indications Oate : 10/]0/97 Tiae t OBt29.5T I Extent I 09/97 N/A I
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-I--I Xceber of RECORDS Selected fraa Current Outage ~
2 xuaber of TUBES Selected froa Current Outage: 2 103
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CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT P 4 09/97 COMPONENT: S/G C Page Date : 09/2q Examination Dates : 09/18/87 thru 09/22/97 Total Number of Tubes Inspected ---..- 3205 Total Indications Between 204 and .394 Greater than or ecpxal to 404 0 Total Corrosion type Indications "VOL" 0 Total Tubes Plugged as Preventive Maint 0 Total Tubes Plugged 0 Location Of Indications 204 to 1004,6 "VOL" Hot Leg Cold Leg TSH .5 to 01H -2.1 TSC .5 to 01C -2.1 0 01H -2.0 to 06H +2.0 : 0 01C -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: 4 104
1NDLCAT1OMS/TREMDlNC REPORT PTN-4 OUTAGE t 09/97 COMPONENT . S/0 C Page DESCRLPTlOM t 20X to 39X indications Date t 10/10/97 Tice t OQ:30.42 Extent I 09/97 I I N/A I
(Tat/Mote( Reel Probe Location (Volts(Des(Ch X fDiff( 'Location ts(Deg I'Vol (Roe(Col(LeRI I- I ( I
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I I I Nudx.r. of RECORDS Selected froa Current Outage: 4
.Murkier of TUSES Selected fran Current OutaDe: 4 105
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CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT P 3 03/97 COMPONENT: S/G A Page .: 1 Date 05/g Examination Dates: 03/13/97 thru 03/19/97 Total Number of Tubes Inspected .....- --3197 Total Indications Between 204 and 394 ~ ~ ~ ~ ~ ~ 6 (1).
Greater than or equal to 40: 0 Total Corrosion type. Indications "VOL" 2 Total Tubes Plugged, as Preventive Maint 1 Total Tubes Plugged 3 Location. Of Indications 204 to 1004 & "VOL"
'Hot Leg Cold Leg TSH -.5 to 01H -2.1 TSC .5 to 01C '-2'.1 0 01H -2.0 to 06H +2.0 01C -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 - 6 (1) 107
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INOlCATIONS/TRENOlNC REPORT PTN-3 OU'TACE : 03/97 COMPONENT : S/0 A Page :
OESCRlPTlON : Al.L ZOX TO 39 lkglCATlONS Oete; 5/12/97 T ime: 10 00'00 4 \ ~ ~ ~
I Extent ) f 03/97 I I M/A IRou(col(Leg) (1st/Mote( Reel Probe '( Lactation (volts(geg(ch JOiff( Loca ion (Volts)Oeg(ch I I I -I
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N~er of RECOROS Selected from Current Outage : , 6 Number of TUSES Selected from Current Outege : 5 108
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1 MO 1 CAT IOMS/TREMO lMG REPGR1 PTM.3 OUTAGE: 03/97 Page ':
COMPONENT : 'S/G A OESCRlPTlOM : ALL CO to 100X, VOL S PTP iMDICATlOMS Oate 5/12/97 T!me: 10-00.0 0 ~
Extent I 03/97 H/A I
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Hunber of TUBES Selected from Current Outage": 3 109
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CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT P 3 03/97 COMPONENT : S/G B Page: 1 Date: 05/12 Examination Dates: 03/13/97 thru 03/19/97 Total, Number of Tubes Inspected 3196 Total Indications Betveen 204 and 394 (1)
Greater than or equal to 404 1 Total Corrosion type Indications <<VOL<< (2)
Total Tubes Plugged as Preventive Maint 2 Total Tubes Plugged 9 (2)
Location Of Indications 204 to 1004 & <<VOL<<
Hot Leg Cold Leg TSH .5 to 01H -2.1 TSC .5 to 01C'2.1 : 0 01H -2.0 to OGH +2.'0 01C -2. 0. to OGC +2. 0: 1 06H +2.1 to AV1 -3.1 e 0 06C +2.1 to AV4 -3.1,: '
AV1 -3.0 to AV4 -3..0: 7 (1) 110
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iXOicATiOXS/TREXOINt REPORT Ptx 3 ovracE 1 03/97 COMPONENT : S/0 8 Page : 1 OESCRIPTiox : ALL ZOX to 39X INoiCAtiONS Oate 5/12/97 time: 10:00 00
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CUMULATIVE DISTRIBUTION
SUMMARY
TURKEY POINT UNIT 4' 03/97 COMPONENT: S/G C Page Date: .05/1 Examination Dates: 03/13/97 thru 03/19/97 Total Number of Tubes Inspected 3181 Total .Indications Between 204 and 394 20 (1)
Greater than or equal to 404 1 Total Corrosion type. Indications "VOL" 0 Total Tubes, Plugged as Preventive Maint 1 Total Tubes Plugged 2 Location Of Indications 204 to 1004 & "VOL"
'Hot Leg Cold Leg TSH .5 to 01H -2. 1 TSC .5 to 01C -2.1 0
,01H -2. 0 to 06H +2. 0 0 01C -2.0 to 06C +2.0 0 06H +2.1 to AV1 -3.1 0 06C +2.1 to AV4 -3.1 ~
0 AVl -3.0 to AV4 -3.0 20 (1) 113
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IMOICAT IOKS/TREND INC REPORt
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POINT. 8 ~f7 SITE VICE PRESICENT R. J. HOVEY OUAUTY MANAGER T. V. A88ATIELLO
~ PROTECTION 8USINESS LICENSING PIANT GENERAL MATERIALS TRAINlhlC SERVICES SYS TENS ENGINEERING MAhNGER MANACER MANAGER MAhNGER ~ '. I MANAGER MANACER G. E. HOLUNGER J. E. KIRKPATRICK W. E. GWINN O. E. JERIIIGAN M. P. HUCA M. L LACAL E. A. THOMPSON HVMAN RESOURCES MANAGER MAINTENANCE OPERA TIONS NQRK TBERMOCAC J. h MARCO MANACER CONTROLS PROIECT MANAGER hNhN GER ~ MAIN GER M. O. PEARCE R. G. WEST R. E. ROSE A T. TIELONKA HEALTH PHYSICS '
OPERATIONS CHEMISTRY SVPERVISOR SUPERVISOR T. O. JONES J. R TREJO
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Performance Highlights e Strong nuclear safety performance e Efficient refueling outages o Safe and reliable unit operation
~ Continued reductions in station operating cost
~ Continued reduction in personnel exposure e Industrial safety No lost time accidents in 1996/1997 Injury rate 0.00 Safety 2000 Initiative
+ 2.8 million manhours without a lost time accident
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P 'l.,~q-.~7@~! P~P~~7i WANG Weighted Overall Performance C3 Unit 3 ml Unit 4 indUstry Median 96.4 96.5 100 93.7 95 88.6 94.8 86.7 90 6.5 84.8 80 70 60 50 40 30 20.
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PLANT SUPPORT Health Physics Collective Radiation Exposure (3-year running average per unit) 400 350 350 332 II person-rem 300 II WANO goal 256 R. 250 E
200 179 145 135 150 110 100 50 0
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PLANT SUPPORT PolNT su Training STRENGTHS' Workforce Development Programs
~ + Operations Programs National Nuclear Accrediting Board noted Turkey Point as industry model for operations training programs 5 of 5 SRO candidates received NRC license
+ Maintenance Programs General Maintenance Leader training Valve maintenance training (focus team)
+ Radiation Protection Program Enhanced radiation worker and radioactive material control training (focus team}
+ Engineering Support Program Oral boards administered by licensed Senior Reactor Operator,
.. engineering supervisor, and training instructor
+ Enhanced leadership training
+ Sitewide human error reduction training 58
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ISi CLOSlgG REMARKS R. J. Hovey 62
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