Text

Technical Specifications Appendix a to License No. DPR-67 and Appendix B Environmental Technical Specifications
	ML18096B437
Person / Time
Site:	Saint Lucie
Issue date:	03/01/1976
From:	; Florida Power & Light Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML18096B437 (450)
	v • d • e

ST. LUCIE PLANT UNIT 1 TECHNICAL SPECIFICATIONS APPENDIX "A" TO LICENSE NO. DPR-67

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS D efined Terms.......'...................................

T hermal Power ..............................................

Rated Thermal Power..................:.....................

J 0 perational Mode...........................................

Actiono ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

Operable - Operability.....

Reportable Occurrence......................................

Containment Vessel Integrity............................... 1-2 Channel Calibration.... 1-2 Channel Check............. . 1-2 Channel Functional Test......;..........................;.. 1-3 C ore Alteration.................................. " ........

J 1-3 S hutdown Margin..................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-3 Identified Leakage.... 1-3 Uni den ti fied Leakage............. 1-4 Pressure Boundary Leakage... . 1-4 C ontrolled Leakage ........ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-4 Azimuthal Power Tilt...... 1-4 Dose Equivalent I-131................... 1-4 E - Average Disintegration Energy ...... 1-4 Staggered Test Basis............... 1-5 Frequency Notation...................... 1-5 Axial Shape Index....................... ~ ~ ~ ~

P Unrodded Planar Radial Peaking Factor Fr" 1-5 Shield Building Integrity.............................. ~ ~ ~ ~ 1-5 Reactor Trip System Response Time....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6 Engineered Safety Feature Response Time. 1-6 P hysics Tests.............................................. 1-6 ST. LUCIE - UNIT 1

INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS R eactor Core ...............................,............ ... 2-1 Reactor Coolant System Pressure ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2 1 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints ..... .. ~ ~ ~ ~ ~ 2 3 BASES SECTION PAGE 2.1 SAFETY LIMITS R eactor Core........... ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 2 1 Reactor Coolant System Pressure ... .. .. .. . ............. B 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints ... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t B 2 4 ST. LUCIE - UNIT 1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.0 APPLICABILITY........................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS II I,

3/4.1.1" BORATION CONTROL...........................,........., . 3/4 1-1 Shutdown Margin - T avg

> 200'F...... 3/4 1-1 Shutdown Margin -' avg- 200'F... .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-3 1

M Boron Dilution 3/4 1-4 I

Moderator Temperature Coefficient .. 3/4 1-5 Minimum Temperature for Criticality 3/4 1-7 3/4.1.2 BORATION SYSTEMS........'.......................... 3/4 1-8 Flow Paths - Shutdown.................'............... 3/4 1-8 Flow Paths. - Operating............ 3/4 1-10 Charging Pump - Shutdown.......... 3/4 1-12 Charging Pumps - Operating........ 3/4 1-13 Boric Acid Pumps - Shutdown....... 3/4 1-14 Boric Acid Pumps - Operating...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-15 Borated Water Sources - Shutdown.. 3/4 1-16 Borated Water Sources - Operating........... ~ ........ 3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES........................... 3/4 1-20 Full Length CEA Position........ 3/4 1-20 Part Length CEA Insertion Limits 3/4 1-23 Position Indicator Channels..... 3/4 1-24 CEA Drop Time................... 3/4 1-26 Shutdown CEA Insertion Limit.... 3/4 1-27 Regulating CEA Insertion Limits...................... 3/4 1-28 ST. LUCIE - UNIT 1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE................. 3/4 2-1 3/4.2.2 TOTAL RADIAL PEAKING FACTOR - F. 3/4 2-6 3/4.2.3 AZIMUTHAL POWER TI LT Tq o ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-9 3/4.2.4 FUEL RESIDENCE TIME,.'............ 3/4 2-11 3/4.2.5 DNB PARAMETERS.............................,.. 3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION;...................................... 3/4 3-9 3/4.3. 3 MONITORING INSTRUMENTATION.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-21 Radiation Monitoring......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-21 Incore Detectors..................................... 3/4 3-25 Seismic Instrumentation..............;............... 3/4 3-27 Meteorological Instrumentation... 3/4 3-30 Remote Shutdown Instrumentation.. ~ ~ ~ '

~ ~ ~ 3/4 3-33 Chlorine Detection Systems .......................... 3/4 3-36 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS................" . .. ...... 3/4 4-1 3/4.4.2 SAFETY VAL'VES - SHUTDOWN............................. 3/4 4-2 3/4.4.3 SAFETY VALVES - OPERATING............................ 3/4 4-3 ST. LUCIE -,UNIT 1 IV

INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4.4.4 PRESSURIZERo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-4 3/4.4.5 STEAM GENERATORS...................................... 3/4 4-5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE........................ 3/4 4-12 Leakage Detection Systems............................. 3/4 4-12 Reactor Coolant System Leakage........................ 3/4 4-14 3/4.4.7 C HEMISTRY.............................................

3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY..................................... 3/4 4-17 3/4.4.9 PRESSURE/TEMPERATURE LIMITS............................ 3/4 4-21 Reactor Coolant System............'.................... 3/4 4-21 P ressurlzer........................................... 3/4 4-25 3/4.4.10 STRUCTURAL INTEGRITY.................................. 3/4 4-26 Safety Class 1 Components..... 3/4 4-26 Safety Class 2 Components..... 3/4 4-37 0 3/4.4.11 Safety Class CORE BARREL 3 Components.....

MOVEMENT..........

3/4 3/4 4-53 4-56 3/4.5 EMERGENCY CORE COOLING SYSTEMS ECCS 3/4.5.1 SAFETY INJECTION TANKS ............................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS T > 300'F............ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ 3/4 5 3 3/4.5.3 ECCS SUBSYSTEMS - T ( 300'F. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 5 7 3/4.5.4 REFUELING WATER TANK...................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 5 8 ST. LUCIE - UNIT 1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3'/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT VESSEL.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 6-1 Containment Vessel Integrity. 3/4 6-1 Containment Leakage.. ~ ~ 3/4 6-2 Containme'nt Air Locks......... 3/4 6-10 Internal Pressure . 3/4 6-12 Air Temperature................... 3/4 6-13 Containment Vessel Structural Integrity...... 3/4 6-14 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS......... 3/4 6-15 Containment Spray System......... 3/4 6-15 Containment Cooling System........................... 3/4 6-17 3/4.6.3 CONTAINMENT ISOLATION VALVES....... 3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL.............................. 3/4 6-23 Hydrogen Analyzers......;............. 3/4 6-23 Electric Hydrogen Recombiners W .... 3/4 6-24 3/4.6.5 VACUUM RELIEF, VALVES....; 3/4 6-26 3/4.6.6 SECONDARY CONTAINMENT........... 3/4 6-27 Shield Building Ventilation System.... 3/4 6-27 Shield Building Integrity........... 3/4 6-30 Shield Building Structural Integrity.. 3/4 6.-31 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ...... 3/4 7-1 Safety Valves.... .....

~ 3/4 7-1 Auxiliary Feedwater System..... 3/4 7-4 Condensate Storage Tank. 3/4 7-6 A ctsvity .................... 3/4 7-7 Main Steam Line Isolation Valves 3/4 7-9 Secondary Water Chemistry............................ 3/4 7-10 ST. LUCIE - UNIT 1 VI

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION........ 3/4 7-13 3/4.7.3 COMPONENT COOLING WATER SYSTEM ........................ 3/4 7-14 .

3/4.7.4 INTAKE COOLING WATER SYSTEM.......................:."... 3/4 7-16 3/4.7.5 ULTIMATE HEAT SINK ...'................................. 3/4 7-18 3/4.7.6 FLOOD PROTECTION....................................... 3/4 7-19 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ............. 3/4 7-20 3/4.7.8 ECCS AREA VENTILATION SYSTEM........................... 3/4 7-24 3/4.7.9 SEALED SOURCE CONTAMINATION......'...................... 3/4 7-27 3/4.7.10 HYDRAULIC SNUBBERS................."........... 3/4 7-29 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES.........:........... 3/4 8-1 Operating......... o ~ ~ ~ ~ ~ ~ 3/4 8-1 S hutdown............. ~ ........... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-7 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS 3/4 8-8 A.C. Distribution - Operating.... 3/4 8-8 A.C. Distribution - Shutdown..... 3/4 8-9 D.C. Distribution - Operating.... 3/4 8-10 D.C. Distribution - Shutdown..... 3/4 8-13 3/4. 9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION......................... 3/4 9-1 3/4.9.2 INSTRUMENTATION............................ ~ 3/4 9-2 3/4.9.3 D E CAY T I ME ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

't ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4 9-3 3/4.9.4 CONTAINMENT PENETRATIONS............................... 3/4 9-4 3/4.9.5 COMMUNICATIONS......................................... 3/4 9-5 3/4.9.6 MANIPULATOR CRANE OPERABILITY.......;....... 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING

~ ~ ~ ~ ~ ~ ~ ~ 3/4 9-7 3/4.9.8 COOLANT CIRCULATION.................................... 3/4 9-8.

3/4.9.9 CONTAINMENT ISOLATION SYSTEM................;.......... 3/4 9-9 ST. LUCIE - UNIT 1 VII

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.9.10 WATER LEVEL - REACTOR VESSEL......................... 3/4 9-10 3/4.9.11 STORAGE POOL WATER LEVEL............................. 3/4 9-11 3/4.9.12 FUEL POOL VENTILATION SYSTEM - FUEL S TORAGE............................................ 3/4 9 12 3/4.9.13 SPENT FUEL CASK CRANE................................ 3/4 9-15 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGINe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION L IMITS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ 3/4 10-2 3/4.10.3 PRESSURE/TEMPERATURE LIMITATION - REACTOR CRITICALITY.....................:........... 3/4 10-3 3/4.10.4 PHYSICS TESTS.............................,... 3/4 10-4 3/4.10.5 CENTER CEA MISALIGNMENT.........,.......;............ 3/4 10-5 ST. LUCIE - UNIT 1 VIII

INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY..........,................................ B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL................................. B 3/4 1-1 3/4.1.2 BORATION SYSTEMS..................................... B 3/4 1-2 3/4.1. 3 MOVABLE CONTROL ASSEMBLIES...................... ~ .... B 3/4 1-3

/

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE ~ .................................... B 3/4 2-1 3/4.2.2 TOTAL RADIAL PEAKING FACTOR.......................... B 3/4 2-1 3/4.2.3 AZIMUTHAL POWER TILT ................... ~ ............ B 3/4 2-1 3/4.2.4 FUEL RESIDENCE TIME..... ........ .. ..... ......

~ ~ ~ ~ ~ ~ ~ ~ B 3/4 2-2 3/4.2.5 DNB PARAMETERSo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 2-2 3 4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION........................... B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION............ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION........................... B 3/4 3-1 ST. LUG IE - UNIT 1 IX

INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.2 and 3/4.4.3 SAFETY VALVES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-1 3/4.4.4 PRESSURIZER.............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-2 3/4.4.5 STEAM GENERATORS............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-4 3/4.4.7 CHEMISTRY...................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY.............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 'o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY........... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-12 3/4.4.11 CORE BARREL MOVEMENT........... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 4-13 3/4.5 EMERGENCY CORE COOLING SYSTEMS ECCS 3/4.5.1 SAFETY INJECTION TANKS. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK (RWST).. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT VESSEL.................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 6-3 3/4.6.5 VACUUM RELIEF VALVES.................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 6-3 3/4.6.6 SECONDARY CONTAINMENT................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 6-4 ST. LUG IE - UNIT 1

INDEX SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................ .. ...

~ ~ B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION. B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM.................. B 3/4 7-4 3/4.7.4 INTAKE COOLING WATER SYSTEM..................... B 3/4 7-4 3/4.7.5 ULTIMATE HEAT SINK................. ~..... ~..... ~ B 3/4 7-4 3/4.7.6 FLOOD PROTECTION................................ B 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM....... B 3/4 7-4 3/4.7.8 ECCS AREA VENTILATION SYSTEM..................... B.3/4 7-5 .

3/4.7.9 SEALED SOURCE CONTAMINATION...............,.... ~ B 3/4 7-5 3/4.7.10 HYDRAULIC SNUBBERS.............................. B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS.....,........................ B 3/4 8-1 3/4. 9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....,.....,..i'............ B 3/4 9-1 3/4.9. 2 INSTRUMENTATION...............-. ~........... B 3/4 9-1 3/4.9.3 DECAY TIME............... B 3/4 9-1 3/4.9.4 CONTAINMENT PENETRATIONS.......... .......

~ B 3/4 9-1 3/4.9.5 COMMUNICATIONS............................ B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE OPERABILITY............. B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING B 3/4 9-2 3/4.9. 8 COOLANT CIRCULATION.. B 3/4 9-2 ST. LUCIE UNIT 1 XI

INDEX BASES SECTION PAGE 3/4.9.9 CONTAINMENT ISOLATION SYSTEM............. B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL WATER LEVEL................... B 3/4 9-2 3/4.9.12 FUEL POOL VENTILATION SYSTEM-FUEL STORAGE ............................ B 3/4 9-3 3/4.9.13 SPENT FUEL CASK CRANE................................ B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN...... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 10-1 3/4.10.3 PRESSURE/TEMPERATURE LIMITATION - REACTOR CRITICALITYe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 10-1 3/4.10.4 PHYSICS TESTS...... B 3/4 10-1 3/4.10.5 CENTER CEA MISALIGNMENT.............................. B 3/4 10-1 ST. LUG IE - UNIT 1 XII

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE E xcluslon Area............................................ 5-1 Low Population Zone........ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-1 Flood Control...... ~ ~ ~ ~ 0 I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-1

5. 2 CONTAINMENT f gura tl on C on I 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ 5-1 Design Pressure and Temperature........................... 5-4 P enetratlons.............................................. 5-4 5.3 REACTOR CORE F uel Assemblies........................................... 5-4 Control Element Assemblies................................ 5-5 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature. ...-.... ................. 5-5 V olume.................... ..... ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-5 5.5 EMERGENCY CORE COOLING SYSTEMS.. .... .. ............;.... 5-5 5.6 FUEL STORAGE C rl t' ti cal tg....... ..

J 1 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-5 Drainage................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-6 5.7 SEISMIC CLASSIFICATION.................................... 5-6 5.8 METEOROLOGICAL TONER LOCATION............................ 5-6 5.9 COMPONENT CYCLE OR TRANSIENT LIMITS..................... 5-6 ST. LUG IE - UNIT 1 XI I I

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY............ ~ ~ ~ 6 1 6.2 .ORGANIZATION 0 ffsite. J 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-1 Facility Staff.... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 1 6.3 FACILITY STAFF UALIFICATIONS.............'................ 6-5 6~4 TRAINING e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6.5 REVIEW AND AUDIT 6.5.1 FACILITY REVIEW GROUP F unct>on........................ 6'-5 C omposltlon ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ 6-5 Alternates....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 6 Meeting Frequency........ 6-6 uorum................ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-6 Responsibilities.... 6-6 Authority......... 6-7 R ecords.................... 6-8 6.5.2 COMPANY NUCLEAR REVIEW BOARD Function............. 6-8 Composition...... 6-9 Alter nates....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 9 Consultants................... ~ ~ ~ ~ 6 9 M t ecting Frequency........................................ 6-9 Q uorum........... 6-9 Review... 6-10 Audits. 6-11 Authority..... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-11 R ecords ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6-12 ST. LUCIE - UNIT 1 XIV

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE OCCURRENCE ACTION......."........................ 6-12 6.7 SAFETY LIMIT VIOLATION..................................... 6-13 6 .8 PROCEDURES................................................. 6-13 6.9 REPORTING RE UIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES............... 6-14 6.9.2 SPECIAL REPORTS................................ . 6-14 6.10 RECORD RETENTION......................... ~ ~ ~ .. . .

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-15 6.11 RADIATION PROTECTION PROGRAM.............................. 6-16 6.12 RESPIRATORY PROTECTION PROGRAM 6.12.1 ALLOWANCE............................................... 6-16 6.12.2 PROTECTION PROGRAM...................................... 6-17 6.12.3 REVOCATION................. ~ ............................ 6-19 6.13 HIGH RADIATION AREA....................... ~ . ~ . ~..... 6-19 ST. LUCIE - UNIT 1 XV

SECTION 1.0 DEFINITIONS

C

1. 0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications.

THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2560 MWt.

OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the specifications.

OPERABLE - OPERABILITY 1.6 A system, subsystem, train, component 'or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified definition shall be the assumption that function(s). Implicit in this all necessary attendant instrumentation, controls, electric power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, "subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

REPORTABLE OCCURRENCE 1.7 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Revision 4 of Regulatory Guide 1.16, "Reporting of Operating Information - Appendix "A" Technical Specifications."

ST. LUCIE - UNIT 1

. DEFINITIONS CONTAINMENT 'VESSEL 'NTEGRITY 1.8 CONTAINMENT VESSEL INTEGRITY shall exist when:

1.8.1 All containment vessel penetrations required to be closed during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed position except as provided in Table 3.6-2 of Specification 3.6.3.1, t

1.8.2 All containment vessel equipment hatches are closed and

,. sealed, 1.8.3 Each containment vessel airlock is OPERABLE pursuant to Specification 3.6.1.3, and 1.8.4 The containment leakage rates are within the limits of Specification 3.6.1.2.

CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequen-tial, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

ST. LUCIE - UNIT 1 1-2

DEFINITIONS 1.11 A CHANNEL FUNCTIONAL TEST shall be the injection. of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is or would be subcritical from its present condition assuming:

a ~ All full length control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn, and

b. No change in part length control element assembly position.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or 2

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. 'eactor'oolant system 'leakage through a steam generator to the secondary system.

ST. LUCI E - UNIT 1 1-3

DEFINITIONS UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be the water flow from the reactor coolant pump seals.

AZIMUTHAL POWER TILT - T 1.18 AZIMUTHAL POWER TILT shall be the maximum difference between the power gen'crated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the cor e.

Power in an core uadrant u er or lower Average power of all quadrants upper or lower)

DOSE E UIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (pCi/gram) which al.one would produce the same dose as the quantity and isotopic mixture of I-131, I-'32, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844,.

E - AVERAGE DISINTEGRATION ENERGY 1.20 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95/ of the total non-iodine activity in the coolant.

ST. LUCI E - UNIT 1 1-4

DEFINITIONS STAGGERED TEST BASIS 1.21 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components .obtained by dividing the specified test interval into n equal subintervals, and

b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

FREQUENCY NOTATION 1.22 The FRE(UENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

AXIAL SHAPE INDEX 1.23 'he AXIAL SHAPE INDEX (Y ) is the power level detected by the lower excore'nuclear instrumen( detectors (L) less the power level detected by the upper excore nuclear instrument detectors (U) divided by the sum of these power levels. The AXIAL SHAPE INDEX (Y ) used for the trip and pretrip signals in the reactor protection system is the above value (Y ) modified by an appropriate multiplier (A) and a constant (B) to deterkiine the true core axial power distribution for that channel.

L-U Y YI AYE + B E L+U UNRODDED PLANAR RADIAL PEAKING FACTOR - F 1.24 The UNRODDED PLANAR RADIAL PEAKING FACTOR is the maximum ratio of the peak to average power density of the individual fuel rods in any of the unrodded horizontal planes, excluding tilt.

SHIELD BUILDING INTEGRITY 1.25 SHIELD BUILDING INTEGRITY shall exist when:

1.25.1 Each door is closed except when the access opening is being used for normal transit entry and exit, and 1.25.2 The shield building ventilation system is OPERABLE.

ST. LUCIE - UNIT 1 1-5

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trio setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.27 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge. pressures reach their required values, etc.). Times shall include diesel generator star ting and sequence loading delays where applicable.

PHYSICS TESTS 1.28 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) .otherwise approved by the Commission.

ST. LUCIE - UNIT 1 1-6

. TABLE 1.1 OPERATIONAL MODES REACTIVITY %RATED AVERAGE COOLANT MODE CONDITION, K THERMAL POWER* TEMPERATURE

1. "

POWER OPERATION > 0.99 . > 5'A > 300'F

2. STARTUP > 0.99 < SC > 300'F

3. HOT STANDBY < 0.99 > 300'F

4. HOT SHUTDOWN < 0.99 0 300 F> T

> 200'F

5. COLD SHUTDOWN < 0.99 200'F

6. REFUELING** .< 0.95 140'F 0

Excluding decay heat.

Reactor vessel head unbolted or removed and fuel in the vessel.

ST. LUCI E - UNIT 1 1-7

TABLE "1.2 FREtRUEIICY NOiAiiOII'OTATION

~RBRUENCY At least once per 12 hours.

At least once per 24 hours.

At least once per 7 days.

At least once per 31 days.

At least once per 92 days.

SA At least onc'e per 6 months.

At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

ST,. LUCIE - UNIT 1 1-8

SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and maxi-mum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of maximum cold leg temper-ature and THERMAL POWER has exceeded the appropriate'ressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant 'System pressure to within its limit within 5 minutes.

ST. LUCIE - UNIT 1 2-1

600 ~ 4 ~

iiif tt. jftf fttf lf t ift} !i'f ii!I ifi! fi! fiii iiii if:-! i!if iii.'ll! ffff I!ii lfi'. tiff i=fj itI i!ii!:.Ii:-i i!i!i if.! jf'f ?f:,l i!if Fi) ~ ~

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i'I I

UNACCEPTABLE OPERATION ~

i l I?Ii !i!!

580 REACTOR OPERATION LIMITED TO LE SS ~ I fli! ';;;! It!*,!".I It:t THAN 5800F BY ACTUATION OF THE I;! :(. :.'.I '.Ifl itt, i'tll MAIN STEAM LINE SAFETY VALVES. I,UNACCEPTABLE; .It;;

-i VESSEL FLOW LESS MEASUREMENT OP E RATION j

- ~

560 ;; jUNCERTAINTIES~370 000. GPM I

I

~

'Ii

.':".i FOR PRE CLAD COLLAPSE ~

I I y I'

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Iit j I ~

I- OPERATION ONLY :. t "! " 'I !i:: i I:,I!I t if LIMITS CONTAIN NO ALLOWANCE I' i!i! -'JI 540 f!'OR

~

INSTRUMENT ERROR OR tI ll ill; :,.

FLUCTUATIONS I,'!

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jf I f~ I pg+ + Iftt I:Ii, 'I. 'f f!if f"i'tll U VALID FOR AXIALSHAPES AND I 'ttt I, ffi ht.

fif! f INTEGRATED ROD RADIAL PEAKING ltl f!I ~

~

i 'tff !If! ll 'I FACTORS L'ESS THAN OR EQUAL TO i O 520 " ii lj II 'tii lt II!

0 fE THOSE ON FIGURE B 2.1-1 I

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!f !I.l !f!ii!:.'i:f!i:i.i!! f!!! t ?;,'!f. I:l..-:. l; "ii:.;If .fi i:,'ji 5PP + 9 PI tf iitIACCEPTABLE;I; ?i.l !i)i,-' ::.L I':!! i n-'

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l? It ;"If:I i: it',I i!Ii ;:;.'!li ii'j f I '

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Ii !II! tI!li iit! .li. i::i iift ilL ii:'! I I~ I

.': i I

I I',II '.!"tf l,:)t lilt it!i I!It ahff jilt I 460 i!ij ! - ~~ '.! '.: II I I 0 0.20 0.40 0.60 0.80 1.00 1.20 1.40 1.60 1.80 2.00 FRACTION OF RATED THERMAL POWER Figure 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT FOUR REACTOR COOLANT PUMPS OPERATING

SAFETY LIMITS,AND LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: AS SHOWN FOR EACH CHANNEL IN TABLE 3.3-1.

ACTION:

With-a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement require-ment of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

ST. LUCIE - UNIT 1 2-3

TABLE'2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP'SETPOINT LIMITS I

n FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES m

1. Manual Reactor Trip Not Applicable Not Applicable

2. Power Level - High (1)

Four Reactor Coolant Pumps < 9.61$ above THERMAL POWER, < 9.61% above THERMAL POWER, and Operating with a minimum setpoint of 15Ã a minimum setpoint of 15K of RATED of RATED THERMAL POWER, and a THERMAL POWER and a maximum of maximum of < 106.5% of RATED < 106.5X of RATED THERMAL POWER.

THERMAL POWER.

3. Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps > 95K of design reactor coolant > 95K of design reactor coolant Operating flow with 4 pumps operating* flow with 4 pumps operating*

4. Pressurizer Pressure - High - < 2400,psia < 2400 psia

5. Containment Pressure - High < 3.9 psig < 3.9 osig

6. Steam Generator Pressure - Low (2) > 485 psig > 485 psig

7. Steam Generator Water Level -Low > 36.3X Water Level - each > 36.3Ã Water Level - each

8. Local Power Density - High (3) steam generator Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2

'rip steam generator set ooint adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2.

Design reactor coolant flow with 4 pumps operating is 390,000 gpm.

TABLE 2.2-1 Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

9. Thermal Margin/Low Pressure (1)

Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted to not Operating exceed the limit lines of exceed the limit lines of Figures 2.2-3 and 2.2-4. Figures 2.2-3 and 2.2-4.

10. Loss of'Turbine Hydraulic > 800 psig > 800 psig Fluid Pressure - Low (3) ll. Rate of Change of Power - High (4) < 2.49 decades per minute < 2.49 decades per minute TABLE NOTATION (1) Trip may be bypassed below 5/ of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 5/ of RATED THERMAL POWER.

(2) Trip may be manually bypassed. below 585 psig; bypass shall be automatically removed at or above 585 psig.

(3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15Ã of RATED THERMAL POWER.

(4) Trip may be bypassed below 10 X and above -15Ã of RATED THERMAL POWER.

1.0 0.8 0.6 0 1.0 QR~

0.4 0.2 0

0 0.2 OA 0.6 0.8 FRACTION OF RATED THERMAL POWER FIGURE 2.2-1 Local Power Density High Trip Setpoint Part 1 (Fraction of RATED THERMAL POWER Versus QR2)

ST. LUCIE - UNIT 1 2-6

1.2 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION YI TRIP IPR Rl QR2 ACCEP A OP ER A ION 0.2 0.6 0.4 0.2 0 ~

0.2 -0.4 -0.6 AXIALSHAPE INDEX, YI Figure 2.2-2 LOCAL POWER DENSITY HIGH TRIP SETPOINT PART 2 (QR2 VERSUS Yi)

1.30 WHERE: ~ A, x QR I QoNe AND 1388 x QoNe + 1 2 5 x TIN 6250 m 1.25 PyAR 0 RE INL p 0 AI 1.15 04 M

I oo 1.10 3YI +0 61 1.05 1.00

-0.5 ~ 0.4 ~ 0.3 ~ 0.2 -0.1 0.1 0.2 0.3 0.4 0.5 AXIALSHAPE INDEX YI FIGURE 2.2-3 Thermal Margin/Low Pressure Trip Setpoint Part 1(YI Versus At)

WHERE: A1 x QR1 QoNa AND VAR 1388 x QONB + 125 TIN - 6250 1.2 1.0 0.8 QR1 0.6 0.4 0.32 0.2 0

0 0.2 0.4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER FIGURE 2.2-4 Thermal Margin/Low Pressure Trip Setpoint Part 2 (Fraction of RATED THERMAL POWER Versus QR1)

ST. LUCIE - UNIT 1 2-9

BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cia'dding perforation which would result in the re-lease of, fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear hear rate at or less than 21 kw/ft. Centerline fuel melting will not occur for this peak linear heat rate. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the claddin'g surface temperature's slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the W-3 correlation.

The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux dis-tributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.

This value corresponds to a 95 percent probability at a 95 percent con-fidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curve's of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the minimum DNBR is no less than 1.30 for the family of axial shapes and corresponding radial peaks shown in Figure B 2.1-1. The limits in Figure 2.1-1 were cal-culated for reactor coolant inlet temperatures less than or equal to 580'F. The dashed line at 580'F coolant inlet temperature is not a safety limit; however, operation above 580'F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POWER levels higher than,ll2Ã of RATED THERMAL POWER is 'pro-hibited by the high power level trip setpoint specified in Table 2.1-1.

The area of safe operation is below and to-the left of these lines.

ST. LUG IE - UNIT 1 B 2-1

2.0 1.6 z0 I

D 1.2 K

I-0 0.

ROD 0.6 RADIAL PEAK 1.44 X

1.44 1.51 0.4 1.56 1.56 10 20 30 40 50 60 70 80 90 100 PERCENT OF ACTIVE CORE LENGTH FROM BOTTOM Figure B2.1-1 Axial Power Distribution for Thermal Margin Safety Limits

SAFETY LIMITS BASES The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.

The reactor protective'system in combination with, the Limiting Condi-tions for Operation, is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and thermal power level that would result in a DNBR of less than 1.30 and preclude the existence of flow instabilities.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant components which permits a maximum transient pressure of 1105 (2750 psia) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, Class I which permits a maximum transient pressure of 110Ã (2750 psia) of component design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code'equirements.

The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.

ST. LUCI E - UNIT 1 8 2-3

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Values have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

Manual Reactor Tri The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Level-Hi h The Power Level-High trip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin/Low Pressure trip.

The Power Level-High trip setpoint is operator adjustable and can be set no higher than 9.61% above the indicated THERMAL POWER level.

Operator action is required to increase the trip setpoint as THERMAL POWER is increased. The trip setpoint is automatically decreased as THERMAL POWER decreases. The trip setpoint has a maximum value of 106.5'A of RATED THERMAL POWER and a minimum setpoint of 15% of RATED THERMAL POWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip would be actuated is 1125 of RATED THERMAL POWER, which is the value used in the safety analyses.

Reactor Coolant Flow-Low The Reactor Coolant Flow-Low trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reactor protective system to permit operation of the reactor at reduced power if one or two ST. LUCI E - UNIT 1 B 2-4

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow-Low (Continued) reactor coolant pumps are taken out of service. The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.30 under normal operation and expected transients. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip setpoints, the Power Level-High trip setpoints, and the Thermal Margin/Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two- or three-pump position. Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.30 during normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating.

Pressurizer Pressure-Hi h The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip s setpoint is 100 psi below the hominal setting (2500 psia) of the pressurizer code safety valves and its 'ift concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.

Containment Pressure-Hi h The Containment Pressure-High trip provides assurance that a reactor trip in initiated concurrently with a .safety injection.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and sub-sequent cooldown of the reactor coolant. The setting of 485 psig is sufficiently below the full-load operating point of 800 psig so as not ST. LUCIE - UNIT 1 B 2-5

LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Pressure-Low (Continued) to interfere with normal operation, but still high enough to provide the required 'protection in the event of excessively high steam flow. This setting was'sed with an uncer'tainty factor of + 22 psi in the accident analyses.

Steam Generator Water Level The Steam Generator Water Level-Low trip pr'ovides core protection by, preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded. The specified setpoint provides allowance that there will be sufficient water.

inventory in the steam generators at the time of trip to provide a margin of more than 10 minutes before auxiliary feedwater is required.

Local Power Densit -Hi h The local Power Density-High trip is provided to prevent the peak local power density in the fuel from exceeding 21 kw/ft during steady state operation thereby assuring that the melting point of the UOz fuel will not be reached. A value of 21 kw/ft is well below the value corresponding to fuel centerline melting.

A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2. The AXIAL SHAPE INDEX is cal-culated from the upper and lower ex-core neutron detector channels. The calculated setpoints are generated as a function of THERMAL POWER level with'he CEA group position being inferred from the THERMAL POWER level.

The trip is automatically bypassed below 15 percent power .

The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment per-mitted for continuous operation are assumed in generation of the set-points. In addition, CEA group sequencing in accordance with the Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

ST. LUCIE - UNIT 1 B 2-6

LIMITING SAFETY SYSTEM SETTINGS BASES Thermal Mar in Low Pressure The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than 1.30,'or when a void fraction limit is exceeded which could result in local flow instability.

The trip is initiated whenever the reactor coolant system pressure signal drops below either 1750 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of aT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function. In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

The Thermal Margin/Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error.

A safety margin is provided which includes: an allowance of 55 of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 2'F to compensate for potential temperature measure-ment uncertainty; and a further allowance of 47 psia to compensate for pressure measurement error, trip system processing error, and time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. The 47 psia allowance is made up of a 22 psia pressure measurement allow-ance, a 5 psia trip system processing allowance and a 20 psia time delay allowance.

Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15K of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuring transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

ST. LUCIE - UNIT 1 B 2-7

LIMITING SAFETY SYSTEM SETTINGS BASES Rate of Chan e of Power-Hi h The Rate of Change of Power-High trip is provided to protect the core during startup operations and its use serves as a backup to the.

administratively enforced star tup rate limit. Its trip setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

ST. LUCIE - UNIT 1 8 2-8

SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS 3/4. 0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for each specification.

3.0.2 Adherence to the requirements of the Limiting Condition for Operation and/or associated ACTION within the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, completion of the ACTION statement is not required.

3.0.3 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months unless corrective measures are completed that permit operation under the permissible ACTION statements for the specified time interval as measured from initial discovery. Exceptions to these requirements shall be stated in the individual specifications.

3.0.4 Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the conditions of the Limiting Con-dition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION statements.

SURVEILLANCE RE UIREMENTS 4.0.1 Surveillance Requirements shall be applicable during the OPERA-TIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:

a. A maximum allowable extension not to exceed 25K of the test interval, and ST. LUCIE - UNIT 1 3/4 0-1

APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)

b. A total maximum combined interval time for any 3 consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.

4.0.3 Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with OPERABILITY requirements for a Limiting Condition for Operation and associated ACTION statements unless otherwise required by the specification.

4.0.4 Entry. into an OPERATIONAL NODE or other specified applicability condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.

ST. LUCIE - UNIT 1 3/4 0-2

3/4.1 REACTIVITY CONTROL SYSTEMS 3 4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T > 200'F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 2.45K Ak/k.

APPLICABILITY: MODES 1, 2*, 3 and 4.

ac7rOV:

With the SHUTDOWN MARGIN < 2.45%%d hk/k, immediately initiate and continue boration at > 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 2.45K hk/k:

a ~ Within one hour after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable.

If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or un-trippable CEA(s).

b. When in MODES or 2 , at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying 1

that CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3.1.3.6.

co 'Jhen in MODE 2 , at least once during CEA withdrawal and at least once per hour thereafter until the reactor is critical.

d. Prior to-initial operation above 5X RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the CEA groups at the Power Dependent Insertion Limits of Specification 3.1.3.6.

See Special Test Exception 3.10.1.

¹ With

~

K > 1 e0 ~

¹¹ With K ff +10 eff ~

EST LUCIE - UNIT 1 3/4 1-1

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued

e. When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by con=

sideration of the following factors:

l. Reactor coolant system boron concentration,

2. CEA position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.05 hk/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4. l. l. l. l.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

ST. LUCIE UNIT 1 3/4 1-2

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN T < 200'F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be > 1.05 ak/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN < 1.0/ ak/k, immediately initiate and continue boration at > 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be > 1.0% Ak/k:

a. Within one hour after detection of an inoperable CEA(.s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable.

If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or un-trippable CEA(s).

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

l. Reactor coolant system boron concentration,

2. CEA position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

ST. LUCIE - UNIT 1 3/4 1-3

REACTIVITY CONTROL SYSTEMS BORON DILUTION L'IMITING:CONDITION FOR OPERATION 0

3.1.1.3 The flow rate of reactor coolant to the reactor pressure vessel shall be > 3000 gpm whenever a reduction in Reactor Coolant System boron concentat7on is being made.

APPLICABILITY: ALL MODES.

ACTION:

With the flow rate of reactor coolant to the reactor pressure vessel

( 3000 gpm, immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System.

SURVEILLANCE; RE UIREMENTS 4.1.1.3 The flow rate of reactor coolant to the reactor pressure vessel shall be determined to be > 3000 gpm within one hour prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either:

a. Verifying at least one reactor coolant pump is in operation, or

b. Verifying that at least one low pressure safety injection pump is in operation and supplying > 3000 gpm to the reactor pressure vessel.

ST. LUCIE - UNIT 1 3/4 1-4

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The, moderator temperature coefficient (MTC) shall be:

a ~ Less positive than 0.5 x 10 hk/k/'F whenever THERMAL POWER is < 70/o of RATED THERMAL POWER,

b. Less positive than 0.2 x 10 a,k/k/ F whenever THERMAL POWER is > 705 of RATED THERMAL POWER, and c~ Less negative than -2.5 x 10 hk/k/'F at RATED THERMAL POWER.

APPLICABILITY: MODES 1 and 2*¹ ACTION:

With the moderator temperature'oefficient outside any one of the above limits, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

With K ) 1.0.

¹See Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 1-5

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:

a. Prior to initial operation above 5X of RATED THERMAL POWER, after each refueling.

b. At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 520 ppm.

c. At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.

ST. LUCIE - UNIT 1 3/4 1-6

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System lowest operating loop temperature (T

v )

shall be > 515'F when the reactor is critical.

APPLICABILITY: MODES 1 and 2*8.

ACTION'ith a Reactor Coolant System operating loop temperature (T ) < 515'F, restore T to within its limit within 15 minutes or be in MT STANDBY within th3 3ext 15 minutes.

SURVEILLANCE RE UIREMENTS 4.1.1.5 The Reactor Coolant System temperature (T shall be determined to be > 515'F. avg )

a. Within 15 minutes prior to achieving reactor criticality, and

b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System temperature (T ) is < 525'F.

avg See Special Test Exception 3.10.3 With K > 1.0.

ST. LUCIE - UNIT 1 3/4 1-7

REACTIVITY"CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING.,CONDITION;FOR.OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:

a. A flow path from the boric acid makeup tank via either a boric acid pump or a gravity feed connection and charging pump to the Reactor Coolant System if only the boric acid makeup tank in Specification 3.1.2.7a is OPERABLE, or

b. The flow path from the refueling water tank via either a charging pump or a high pressure safety injection pump to the Reactor Coolant System if only the refueling water tank in Specification 3.1.2.7b is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status.

SURVEILLANCE.RE UIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demon-strated OPERABLE:

a ~ At least once per 7 days by:

1. Cycling each testable oower operated or automatic valve in the flow path required for boron injection through at least one complete cycle of full travel, and

2. Verifying that the temperature of the heat traced portion of the flow path is above the temperature limit line shown on Figure 3.1-1 when a flow path from the boric acid makeup tanks is used.

ST. LUCIE - UNIT 1

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS

b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position,-is in its correct position.

ST. LUCIE - UNIT 1 3/4 1-9

REACTIVITY CONTROL 'YSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:

a ~ Two flow paths from the boric acid makeup tanks via either a boric acid pump or a gravity feed connection, and a charging pump to the Reactor Coolant System, and

b. The flow path from the refueling water tank via a charging pump to the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron i,njection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or make the reactor subcritical within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and borate to a SHUTDOWN MARGIN equivalent to at least 1/ hk/k at 200 F; restore at least two fl.ow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.

ST. LUCIE - UNIT 1 3/4 1-10

REACTIVITY CONTROL SYSTEMS SURVEILLANCE.REQUIREMENTS Continued)

2. Verifying that the temperature of the heat traced portion of the flow path from the boric acid makeup tanks is above the temperature limit line shown on Figure 3.1-1.

b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c~ At least once per 18 months during shutdown by:

1. Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least once complete cycle of full travel.

2. Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection Actuation signal.

ST. LUCI E - UNIT 1 3/4 1-11

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump or one high pressure safety injection pump in the boron injection flow path required OPERABLE pur-suant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no charging pump or high pressure safety injection pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one of the required pumps is restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.1.2.3 At least the above required charging pump or high pressure safety injection pump shall be demonstrated OPERABLE at least once per 31 days by:

a. Starting (unless already operating) the pump from the control room,

b. Verifying pump operation for at least 15 minutes, and

c. Verifying that the pump is aligned to receive electrical power from an OPERABLE emergency bus.

T. LUCIE - UNIT 1 3/4 1-12

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE at least once per 31 days on a STAGGERED TEST BASIS by:

a. Starting (unless already operating) each pump from the control room, and

b. Verifying that each pump operates for at least 15 minutes.

ST. LUCIE - UNIT 1 3/4 1-13

'EACTIVITY 'ONTROL'YSTEMS BORIC ACID PUMPS - SHUTDOWN LIMITING CONDITION 'OR OPERATION, 3.1.2.5 At least one boric acid pump shall be OPERABLE if only the flow path through the boric acid pump in Specification 3.1.2.la above, is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no boric acid pump OPERABLE as required'to complete the flow oath of Specification 3.1.2.la, suspend all operations involving CORE ALTERA-TIONS or positive reactivity changes until at least one boric acid pump is restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.1.2.5 At least the above required boric acid pump shall be demonstrated OPERABLE at least once per 7 days by:

a ~ Starting (unless already operating) the pump from the control room,

b. Verifying, that on recirculation flow, the pump develops a discharge pressure of > 75 psig, and c~ Verifying pump operation for at least 15 minutes.

ST. LUCIE - UNIT 1 3/4 1-14

REACTS'VITY'CONTROLSYSTEMS BORIC ACID'PUMPS - OPERATING LIMITING CONDITION. FOR OPERATION 3.1.2.6 At least the boric acid pump(s) in the boron injection flow path(s) required OPERABLE pursuant to Specification 3.1.2.2a,.shall be OPERABLE if the flow path through the boric acid pump in Specification 3.1.2.2a is OPERABLE.

.P APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one boric acid pump required for the boron inje'ction flow path(s) pursuant to Specification 3.1.2.2a inoperable, restore the boric acid pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4.1.2.6 At least the above required boric acid pump(s) shall be demonstrated OPERABLE at least once per 7 days by:

a 4 Starting (unless already operating) the pump from the control room,

b. Verifying, that on recirculation flow, the pump develops a discharge pressure of > 75 psig, and c ~ Verifying pump operation for at least 15 minutes.

ST. LUCIE - UNIT 1 3/4 1-15

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE:

a. One boric acid makeup tank and one associated heat tracing circuit with the tank contents in accordance with Figure 3.1-1.

b. The refueling water tank with:

1. A minimum contained volume of 125,000 gallons,

2. A minimum boron concentration of 1720'pm, and

3. A minimum solution temperature of 40'F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water sources OPERABLE, suspend all operations involving positive reactivity changes until at least one borated water source is restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

l.

Verifying the boron concentration of the water,

2. Verifying the water level of the tank, and

3. Verifying the boric acid makeup tank solution temperature when it is the source of borated water.

b. At least once per'4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by verifying the RWT temperature when it is the source of borated water and the site ambient air temperature is < 40'F.

ST. LUCIE - UNIT 1 3/4 1-16

9,500 200 180 9,000 160 140 R

0 U

- 8,500 120 D

0 R

100 I P V

I Q 8,000 80 0

O CC O

m z

7,500 40 20 7,000 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13j STORED BORIC ACID CONCENTRATION (WT%)

Figure 3.1-1 Minimum Boric Acid Makeup Tank Volume and Temperature as a Function of Stored Boric Acid Concentration ST. LUCIE - UNIT 1 3/4 1:-17

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES LIMITING CONDITION OPERATING FOR OPERATION 0

3.1.2.8 At least two of the following three borated water sources shall be OPERABLE:

a., Two boric acid makeup tanks 'and one associated heat tracing circuit with the, contents of the tanks in accordance with Figure 3.1-1, and

b. The refueling water tank with:

l.. A minimum contained volume of 371,800. gallons of water,

.2. A minimum boron concentration of 1720 ppm,

3. A maximum solution temperature of 100'F,

4. A minimum solution temperature of 55'F when in MODES 1 and 2, and

5. A minimum solution temperature of 40'F when in MODES 3 and 4,

APPLICABILITY: MODES 1, 2, 3.and 4.

ACTION:

With only one borated water source OPERABLE, restore at least two borated water sources to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or make the reactor subcritical within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and borate to a SHUTDOWN MARGIN equivalent to at least 1/ hk/k at 200'F; restore at least two borated water sources to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.1.2.8 At least two borated water sources shall be demonstrated OPERABLE:

a. At least one per 7 days by:

l. Verifying the boron concentration in each water source, ST. LUCI E - UNIT 1 3/4 1-18

REACTIVITY CONTROL SYSTEMS

2. Verifying the water level in each water source, and

3. Verifying the boric acid makeup tank solution temperature.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWT temperature.'T.

LUCIE UNIT 1 3/4 1-19

REACTIVITY CONTROL'YSTEMS 3/4.'1.3 MOVABLE CONTROL ASSEMBLIES FULL LENGTH CEA POSITION LIMITING CONDITION FOR OPERATION 3.1.3.1 The CEA Block Circuit and all full length (shutdown and regulating)

CEAs shall be OPERABLE with each CEA of a given group positioned within 7.5 inches (indicated position) of all other CEAs in its group.

APPLICABILITY: MODES 1* and 2*.

ACTION:

a ~ With one or more full length CEAs inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, be in HOT STANDBY within 6 hours.

With the CEA Block Circuit inoperable, within 6 hours either:

Restore the CEA Block Circuit to OPERABLE status, or Place and maintain the CEA-drive system mode switch in either the "Manual" or "Off" position and fully with-draw all CEAs in groups 3, 4, 5 and 6 and withdraw the CEAs in group 7 to less than 5X insertion, or

3. Be in at least HOT STANDBY.

c ~ With one full length CEA inoperable (unless imovable as a result of excessive friction or mechanical interference or

, known to be untrippable) but within its above specified alignment requirements, operation in MODES 1 and 2 may con-tinue for up to 7 days per occurrence with a total accumulated time of < 14 days per calendar year.

d. With one or more full length CEAs misaligned from any other CEAs in its group by more than 7.5 inches but less than 15 inches, operation in MODES 1 and 2 may continue, provided that-within one hour the misaligned CEA(s) is either:

1. Restored to OPERABLE status within its above specified alignment requirements, or Special Test Exceptions 3.10.2 and 3.10.5.

4 See ST. LUCIE - UNIT 1 3/4 1-20

REACTIVITY CONTROL'YSTEMS .

FUL'L LENGTH CEA POSITION (Continued)

L'IMITING.CONDITION FOR, OPERATION (Continued

2. Declared inoperable. After declaring the CEA inoperable, operation in MODES 1 and 2 may continue for up to 7 days per occurrence with a total accumulated time of < 14 days per calendar year provided all of the following conditions are met:

a) The THERMAL POWER level shall be reduced to < 70K of the maximum allowable THERMAL POWER level for the

'xisting Reactor Coolant Pump combination within one hour;,.if negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used.

b) Within one hour after reducing the THERMAL POWER as required by a) above, the remainder of the CEAs in the group with the inoperable CEA shall be aligned to within 7.5 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1-2; the THERMAL POWER level, shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

1

e. With one full length CEA misaligned from any other CEA in by 15 inches or more, reduce THERMAL POWER to < 705 of its'roup the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination within one hour; if negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used. Within one hour after reducing THERMAL POWER as required above, either:

1. Restore the CEA to within the above specified alignment requirements, or

2. Declare the CEA inoperable. After declaring the CEA inoperable, POWER OPERATION may continue for up to 7 days per occurrence with a total accumulated time of < 14 days per calendar year provided the remainder of the CEAs in the group with the inoperable CEA are aligned to within 7..5 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

ST. LUCI E - UNIT 1 3/4 1-21

REACTIVITY CONTROL SYSTEMS FULL LENGTH CEA POSITION (Continued LIMITING CONDITION FOR OPERATION (Continued

f. With more than one full length CEA inoperable or misaligned from any other CEA in its group by 15 inches (indicated position) or more, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE.RE UIREMENTS 4.1.3.1.1 The position of each full length CEA shall be determined to be within 7.5 inches (indicated position) of all other CEAs in its group at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Deviation Circuit and/or CEA Block Circuit are inoperable, then verify the individual CEA positions at, least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full length CEA not fully inserted shall be determined to be OPERABLE by movement of at least 7.5 inches in any one direction at least once per 31 days.

4.1.3.1.3 The CEA Block once oer 31 days by a Circuit shall be demonstrated OPERABLE at least functional test which verifies that the circuit maintains the CEA overlap and sequencing requirements of Specification 0

3.1.3.6 and that the circuit prevents the regulating CEAs from being inserted beyond the Power Dependent Insertion Limit of Figure 3.1-2.

E ST. LUCIE - UNIT 1 3/4 1-22

REACTIVITY 'ONTROL SYSTEMS PART LENGTH CEA'NSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.2 All part length CEAs shall be withdrawn to at least 132.0 inches.

APPLICABILITY: MODES 1* and 2*.

ACTION:

With a maximum of one PLCEA withdrawn to less than 132.0 inches, either:

a. Withdraw the PLCEA to at least 132.0 inches within one hour, or

b. Be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE.RE UIREMENTS 4.1.3.2 Each part length CEA shall be determined withdrawn to at least 132.0 inches by:

a. Verifying the positions of the PLCEAs prior to increasing THERMAL POWER above 5A of RATED THERMAL POWER, and

b. Verifying, at least once per 31 days, that electric power has been disconnected from its drive mechanism by physical removal of a breaker from the circuit.

See Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 1-23

REACTIVITY CONTROL 'SYSTEMS POSITION INDICATOR CHANNELS LIMITING.CONDITION FOR OPERATION.

3.1.3.3 All shutdown, regulating and part length CEA reed switch posi-tion indicator channels and CEA pulse counting position indicator chan-nels shall be OPERABLE and capable of determining the absolute CEA positions within + 2.25 inches.

APPLICABILITY: MODES 1 and 2.

ACTION:

a ~ With one or more PLCEA reed switch or pulse counting position indicator channels inoperable and the applicable PLCEA fully withdrawn and electric power to its drive mechanism dis-connected, operation may continue provided the applicable PLCEA is verified imnediately and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter to be fully withdrawn by i'ts "Full Out" limit.

~ b. With a maximum of one reed switch position indicator channel per group or one (except as permitted by ACTION item d. below)

'ulse counting position indicator channel per group inoperable and the CEA(s) with the inoperable position indicator channel partially inserted, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

l. Restore the inoperable position indicator channel to OPERABLE status, or

2. Be in HOT STANDBY, or

3. Reduce THERMAL POWER to < 70Ã of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination; if negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used. Operation at or below this reduced THERMAL POWER level may continue provided that within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a) The CEA group(s) with the inoperable position indi-cator is fully withdrawn while maintaining the withdrawal sequence required by Specification 3.1.3.6 and when this CEA group reaches its fully withdrawn position, the "Full Out" limit of the CEA with the .

inoperable position indicator is actuated and verifies this CEA to be fully withdrawn. Subsequent to fully withdrawing this CEA group(s), the THERMAL POWER level may be returned to a level consistent with all other applicable specifications; or ST. LUCIE - UNIT 1 3/4 1-24

REACTIVITY CONTROL SYSTEMS 0 POSITION INDICATOR CHANNELS LIMITING CONDITION Continued FOR OPERATION b) The CEA group(s) with the inoperable position indi-cator is fully inserted, and subsequently maintained fully inserted, while maintaining the withdrawal sequence and THERMAL .POWER level required by Speci-fication 3.1.3.6 and when this CEA group reaches its fully inserted position, the "Full In" limit of the CEA with the inoperable position indicator is actuated and verifies this CEA to be fully inserted. Subsequent operation shall be within the limits of Specification 3.1.3.6.

c. With a maximum of one reed switch position indicator channel per group or one pulse counting position indicator channel per group inoperable and the CEA(s) with the inoperable position indicator channel at either its fully inserted position or fully withdrawn position, operation may continue provided:

1. The position of this CEA is verified immediately and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter by its "Full In" or "Full Out" limit (as applicable),

2. The fully inserted CEA group(s) containing the inoperable position indicator channel is subsequently, maintained fully inserted, and

3. Subsequent operation is within the -l.imits of Specifica-tion 3.1.3.6.

d. With one or more pulse counting position indicator, channels inoperable, operation in MODES 1 and 2 may continue for'up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided all of the reed switch position indicator channels are OPERABLE.

SURVEILLANCE .RE UIREMENTS 4.1.3.3 Each position indicator channel shall be determined to be OPERABLE by verifying the pulse counting position indicator channels and the reed switch position indicator channels agree within 4.5 inches at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Deviation circuit is inoperable, then compare the pulse counting position indicator reed switch position indicator channels at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

'nd ST. LUCIE - UNIT 1 3/4 1-25

REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 0

3.1.3.4 The individual full length (shutdown and control) CEA drop time, from a fully withdrawn position, shall be ( 3.3 seconds from when electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:

a. T 515 F, and

b. All r'eactor coolant pumps operating.

APPLICABILITY: MODE 3 ~

ACTION:

a. With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.

SURVEILLANCE RE UIREMENTS 4.1.3.4- The CEA drop time of full length CEAs shall be demonstrated through measurement prior to reactor criticality:

a ~ For all CEAs following each removal of the reactor vessel head,

b. For specifically affected individual CEAs following any main-tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and c ~ At least once per 18 months.

ST. LUCIE - UNIT 1 3/4 1-26

REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION. FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to at least 132.0 inches.

APPLICABILITY:'ODES 1 and 2*¹.

ACTION:

With a maximum of one shutdown CEA withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, to less than 132.0 inches, within one hour either:

a. Withdraw the CEA to at least 132.0 inches, or

b. Declare the CEA inoperable and apply Specification 3.1.3.1.

SURVEILLANCE,RE UIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to at least 132.0 inches:

a. Within 15 minutes prior to withdrawal of any CEAs in regulating groups during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

See Special Test Exception 3.10.2.

¹-Wzth K > 1.0.

ST. LUCI E - UNIT 1 3/4 1-27

REACTIVITY CONTROL'YSTEMS REGULATING CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits shown on Figure 3;1-.2 with CEA insertion between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits restricted to:

a. < 4 hours per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval,

b. < 5 Effective Full Power Days per 30 Effective Full Power Day interval, and

c. < 14 Effective Full Power Days per calendar year.

APPLICABILITY: MODES 1* and 2*5'.

ACTION:

a ~ With the regulating CEA groups inserted beyond the Power Dependent Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either:

1. Restore the regulating CEA groups to within the limits, or-- ..

2. Reduce THERMAL'POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position using the above figure.

b. With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits for intervals > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, except during operation pursuant to the provisions of ACTION items

c. and d. of Specification 3.1.3.1, operation may proceed pro-vided either:

The Short Term Steady State Insertion Limits of Figure 3.1-2 are not exceeded, or F

2. Any subsequent increase in THERMAL POWER is restricted to < 5% of RATED THERMAL POWER per hour.

See Special Test Exceptions 3.10.2 and 3.10.5.

With K > 1.0.

ST. LUCIE - UNIT 1 3/4 1-28

REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS Continued)

LIMITING CONDITION FOR OPERATION c ~ With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Power Dependent Inser-tion Limits for intervals > 5 EFPD per 30 EFPD interval or

> 14 EFPD per calendar year, except during'perations pursuant to the provisions of ACTION items c. and d. of Specification 3.1.3.1, either:

1. Restore the'egulating groups to within the Long Term Steady State Insertion Limits within two hours, or

2. Be in HOT STANDBY within 6 hours.

SURVEILLANCE RE(jUIREMENTS 4.1.3.6 The position of each regulating CEA group shall be determined to be within the Power Dependent Insertion Limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the PDIL Auctioneer Alarm Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The accumulated times during which the regu-lating CEA groups are inserted between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ST. LUCIE - UNIT 1 3/4 1-29

1.00 0.90 0.80 LONG TERM STEADY STATE INSERTION

a. 0.70 WE D PE Ml

'IMIT NS ION cL' 0.60 Q 0.50 0.40 0

zO SHORT TERM 030 STEADY STATE O INSERTION LIMIT

u. 0.20 0.10 0

GROUPS 0 27 55 82 109 137 0 27 55 82 109 137 0 27 55 82 109 137 6 4' 27 55 82 109 137 0 27 55 82 109 137 CEA INSERTION(INCHES)

Figure 3.1-2 CEA'nsertion Limits vs THERMAL POWER with 4 Reactor Coolant.-

Pumps Operating

3 4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall. not exceed the limits shown on Figure 3.2-1.

APPLICABILITY: MODE 1 .

ACTION:

With the linear heat rate ex'ceeding its limits, as indicated by four or more coincident incore channels or by the AXIAL SHAPE INDEX 'outside of the power dependent limits on the Power Ratio Recorder, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:

a. Restore the linear heat rate to within its limits within one hour, or

b. Be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.1.1 The provisions of Specification 4.0.4 are not applicable.

V 4.2.1.2 The linear heat rate shall be determined to be within its limits

'by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring system.

4.2.1.3 Excore Detector Monitorin S stem - The excore detector moni-toring system may be used for monitoring the core power distribution by:

a. Verifying at least once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limits shown on Figure 3.2-2.

b. Verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2, where 100 percent of maxi-mum allowable power represents the maximum THERMAL POWER allowed by the determination made in Specification 4.2.1.3.c, and ST. LUCIE - UNIT 1 3/4 2-1

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued c~ Verifying at least once per 31 days that the THERMAL POWER does not exceed the value determined by the following relationship:

L x M 17.85 where:

1. L is the maximum allowable linear heat rate as determined from Figure 3.2-1 and is based on the core average burnup at the time of the la'test incore flux map.

2. M is the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination.

4.2.1.4 Incore Detector Monitorin S stem - The incore detector moni-toring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:

a. Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 days.

b. Have their alarm setpoint adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alarms:

1. Flux peaking augmentation factors as shown in Figure 4.2-1,

2. A measurement-calculational uncertainty factor of 1.10,

3. An engineering uncertainty factor of 1.03,

4. A linear heat rate uncertainty factor of 1.01 due to axial fuel densification and thermal expansion,

5. A THERMAL POWER measurement uncertainty factor of 1.02, and

6. A rod bow penalty factor of 1.05.

ST. LUCIE - UNIT 1 3/4 2-2

16 UNACCEPTABLE OPERATION 15.8 I- ACCEPTABLE OPERATION llJ

~o~

K I- ~K

~o Z0 15 4+

~o Z

~0

+

LV ~

CL D

JQ)

LU lL 14 O

13 100 200 300 400 500 EFFECTIVE FULL POWER DAYS Figure 3.2-1 Allowable Peak Linear Heat Rate vs Burnup

1.0 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION I 09

~ >>t

.," f!r SH 0.8

~4 ffj! i!

t>> >>>> >>

O ACCEPTABLE 0z 0.T OPERATION

~

I i>> ~>>\

g 0.6

'i ft'ijf

~

ii t

I!I t>> >>>>

o I

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Stt 1 St

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>> SSL ~1 St off f! >>

0.5

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t >>I jt sr, ir.

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0.4 0.8 ~ 0.6 ~ 0.4 0.2 0 0.2 0.4 0.6 0.8 PERIPHERAL AXIALSHAPE INDEX Figure 3.2.2 AXIALSHAPE INDEX vs Fraction of Maximum Allowable Power Level per Specification 4.2.1.3 ST. LUCIE - UNIT 1 3/4 2-4

1.10 1.08 1.06 Z

0

'1.04 1.02 1.00 0 14 28 42 56 70 84 98 112 126 140 DISTANCE FROM BOTTOM OF CORE, INCHES Figure 4.2-1 Augmentation Factors vs Distance from Bottom of Core

POWER DISTRIBUTION LIMITS TOTAL RADIAL PEAKING FACTOR - F LIMITING CONDITION FOR OPERATION T

3.2.2 The calculated value of F, r' defined as F = F r (1+Tq ), shall be to < 1.36. 'imited APPLICABILITY: MODE 1*.

ACTION:

With F r > 1.36, within 6 hours either:

a ~ Reduce THERMAL POWER to bring the combination of THERMAL POWER and F to within the limits of Figure 3.2-3, fully withdraw the PLCEAF and withdraw the full length CEAs to or beyond the Long Term Steady State Inser tion Limits of Specification 3.1.3.6; or b.

c.

Reduce the Local Power Pressure Calculator by Be in trip HOT STANDBY.

a > r, Density - High and Thermal Margin/Low setpoints and the setpoints on the Power Ratio factor equivalent to FT

~36 or SURVEILLANCE RE UIREMENTS 4.2.2.1 The provision's of Specification 4.0.4 are,not applicable.

4.2.2.2 F shall be calculated by the expression F = F (1+Tq) and F T shall be determined to be within its limit at the following intervals:

a. Prior to operation abdve 70 percent of RATED THERMAL POWER after each fuel loading,

b. At least once per 31 days of accumulated operation in MODE 1, and

/

1 p 1 1 1 2 p 1 2.12.2.

ST. LUCIE - UNIT 1 3/4 2-6

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued

c. Within four hours if the AZIMUTHAL POWER TILT (T )

q is > 0.02.

4.2.2.3 F shall be determined each time a calculation of F, is required by using the incore detectors to obtain a power distributionrmap with no part length CEAs inserted and with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant-Pump combination. This determination shall be limited to core planes be-tween l5$ and 85Ã of full core height and shall exclude regions influenced by grid effects.

4.2.2.4 T shall be determined each tiye a calculation of F is required and the value of T used to determine F r .shall be the measured value of T .

q H

ST. LUCIE - UNIT 1 3/4 2-7

0.90 0.80 UNACCEPTABLE OP E RATION 0.70 0.60 0.50 ACCEPTABLE OPERATION 0.40'.30 0.20 0.10 1.30 1.34 1.38 1.42 1.46 1.50 1.54 1.58 1.62 MEASURED FT FIGURE 3.2-3 Allowable Combinations of THERMAL POWER and Ff

POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - T LIMITING CONDITION FOR OPERATION 3.2.3 The AZIMUTHAL POWER TILT (T ) shall not exceed 0.02.

APPLICABILITY: MODES 1* and 2*.

ACTION:

a With the indicated AZIMUTHAL POWER TILT determined to be

~

> 0.02 but < 0.10, 'either correct the power tilt within two hours or determine within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at least once per subsequent 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , that the TOTAL RADIAL PEAKING FACTOR (F ) is within the limit of Specification 3.2.2.

r

b. With the indicated AZIMUTHAL POWER TILT determined to be

> 0.10, operation may proceed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided F$

or the combination of Fr and THERMAL POWER is maintained within the limit of Specification 3.2.2. Subsequent operation for the purpose of measurement and to identify the cause of the tilt is allowable provided:

1. The THERMAL POWER level is restricted to < 20% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination, and

2. The Local Power Density-High and Thermal Margin/Low Pressure trip setpoints and Power, Ratio Calculator set-Fr meas .

points are reduced by a factor equivalent to >

SURVEILLANCE RE UIREMENT 4.2.3.1 The provisions of Specification 4.0.4 are not'applicable.

4.2.3.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit by:

a ~ Calculating'he tilt at least once per 7 days when the Subchannel Deviation Alarm is OPERABLE, See Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 2-9

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued

b. Calculating the Subchannel tilt at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Deviation Alarm is inoperable, and when the

c. Using the incore detectors to determine the AZIMUTHAL POWER TILT at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when one excore safety channel is inoperable and THERMAL POWER is > 755 of RATED THERMAL POWER.

ST. LUCIE - UNIT 1 3/4 2-10

'POWER'DISTRIBUTION'L'IMITS FUEL RESIDENCE 'TIME LIMITING CONDITION FOR. OPERATION 3.2.4 The core average fuel burnup shall be limited to < 500 Effective Full Power Days during the initial fuel cycle.

APPLICABILITY: MODE 1.

ACTION:

With the core average fuel burnup determined to exceed 500 Effective Full Power Days, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE;RE UIREMENTS 4.2.4 The core average fuel burnup, based on gross thermal energy generation, shall be determined by calculation at least once per 31 days.

ST. LUCIE - UNIT 1 3/4 2-11

POWER DISTRIBUTION "LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The follow'ing DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a. Cold Leg Temperature

b. Pressurizer Pressure

c. Reactor Coolant System Total Flow Rate

d. AXIAL SHAPE INDEX APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to '< 5/ of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REOUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits by instrument readout at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

ST. LUCIE - UNIT 1 3/4 2-12 0

TABLE 3.2-1 DNB MARGIN LIMITS Four Reactor Coolant Pumps Parameter ~0eratin Cold Leg Temperature < 542'F Pressurizer Pressure > 2225 psia*

Reactor Coolant Flow Rate > 370 000 gpm AXIAL SHAPE INDEX Figure 3.2-4

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER or a THERMAL POWER step increase of greater than lOX of RATED THERMAL POWER.

ST. LUG IE UNIT 1 3/4 2-13

1.2 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION 1.0 ACCEPTABLE OPERATION Z

I 0.8 I

0 R

O 0.6 0.8 0.6 0.2 0 4I.2 0.4 Yi FIGURE 3.2-4 AXIALSHAPE INDEX Operating Limits with 4 Reactor Coolant Pumps Operating ST. LUCIE - UNIT '1 3/4 2-14

3 4. 3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION. FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of 'Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE RE UIREMENTS 4.'3.1.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1..

4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1:

ST. LUG IE - UNIT 1 3/4 3-1

TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 1, 2 and *

2. Power Level - High 2(a) 3(f) 1 2

3. Reactor Coolant Flow - Low 4/SG 2(a)/SG 3/SG

4. Pressurizer Pressure - High 1, 2

5. Containment Pressure - High 4 1, 2

6. Steam Generator Pressure - Low 4/SG 2(b)/SG 3/SG 1, 2

7. Steam Generator Water Level - Low 4/SG 2/SG 3/SG 1, 2

8. Local Power Density - High 2(c)

9. Thermal Margin/Low Pressure 4 2(a) 1, 2 (e)

10. Loss of Turbine--Hydraulic Fluid Pressure - Low 2(c)

TABLE 3.3-1 Continued l REACTOR PROTECTIVE INSTRUMENTATION foal MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION ll. Wide Range Logarithmic Neutron Flux Monitor

a. Startup and Operating--

Rate of Change of Power-High 4 2(d) 1, 2 and *

b. Shutdown 4 0 3, 4, 5

12. Reactor Protection System 2 1 2*

Logic

13. Reactor Trip Breakers 1, 2*

TABLE 3.3-1 Continued TABLE NOTATION With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 5% of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 585 psig; bypass shall be automatically removed at or above 585 psig.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15% of RATED THERMAL POWER.

(d) Trip may be bypassed below 10 4% and above 15% of RATED THERMAL POWER.

(e) .Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3.

(f) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Monitoring Channels.

ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:

a ~ < 5% of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour and restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after increasing THERMAL POWER above 5% of RATED THERMAL POWER; otherwise, reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. > 5% of RATED THERMAL POWER, operation may continue provided all of the following conditions are satisfied:

ST. LUCIE - UNIT 1 3/4 3-4

TABLE 3.3-1 Continued ACTION STATEMENTS

1. The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2. All functional units receiving an input from the tripped channel are also placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

3. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1 provided one of the inoperable channels is placed in the tripped condition.

ACTION 3 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:

a ~ < 55 of RATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour and restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after increasing THERMAL POWER above 5X of RATED THERMAL POWER; otherwise, reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. > 5X of RATED THERMAL POWER, POWER OPERATION may continue provided all of the following conditions are satisfied:

1. , The inoperable channel is placed in the tripped condition within one hour.

2. The Minimum Channels OPERABLE requirements is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1 provided'ne of the inoperable channels is placed in the tripped condition.

ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 5 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ST. LUCIE - UNIT 1 3/4 3-5

TABLE 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Tri p Not 'Appl i cable

2. Power Level - High < 0.40 seconds*

3. Reactor Coolant Flow - Low < 0.65 seconds

4. Pressurizer Pressure - High < 0.90 seconds

5. Containment Pressure - High < 1.40 seconds

6. Steam Generator Pressure - Low < 0.90 seconds

7. Steam Generator Water Level - Low < 0.90 seconds

8. Local Power Density - High < 0.40 seconds*

9. Thermal Margin/Low Pressure < 0.90 seconds*

10. Loss of Turbine--Hydraulic Fluid Pressure Low Not Applicable ll. Wide Range Logarithmic Neutron Flux Monitor Not Applicable

Neutron detectors are exempt from response time testing. Response time shall be measured from detector output or input of first electronic component in channel.

TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED

1. Manual Reactor Trip N.A. N.A. S/U(1) N.A.

2. Power Level - High

a. Nuclear Power D(2), M(3),Q(5) M 1 2

b. aT Power D(4), q 1

3. Reactor Coolant Flow - Low R M 1, 2

4. Pressurizer Pressure - High R M 1 ~ 2

5. Containment Pressure - High R M 1, 2

6. Steam Generator Pressure - Low R M 1, 2

7. Steam Generator Water Level - Low 1, 2

8. Local Power Density - High 1

9. Thermal Margin/Low Pressure 1, 2

10. Loss of Turbine--Hydraulic Fluid Pressure - Low N.A. N.A. S/U(1) N.A.

ll. Wide Range Logarithmic Neutron N.A. S/U(1) 1, 2, 3, 4, Flux Monitor 5 and

12. Reactor Protection System Logic N.A. N.A. M and S/U(l) 1, 2

13. Reactor Trip Breakers N.A. N.A. M 1, 2 and

TABLE 4.3-1 Continued TABLE NOTATION With reactor trip breaker closed.

(1) - If not performed in previous 7 days.

Heat balance only, above 15K of RATED THERMAL POWER; adjust "Nuclear Power Calibrate" potentiometer to null "Nuclear Pwr - tT Pwr." During PHYSICS TESTS, these daily calibrations of nuclear power and aT power may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

(3) Above 15% of RATER THERMAL POWER, recalibrate the excore detectors which monitor the AXIAL SHAPE INDEX by using the incore detectors or restrict THERMAL POWER during subsequent operations to < 90Ã of the maximum allowed THERMAL POWER level with the existing Reactor Coolant Pump combination.

(4) - Adjust "hT Pwr Calibrate" potentiometers to make aT power signals agree with calorimetric calculation.

(5) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

ST. LUCIE - UNIT 1 3/4 3-8

INSTRUMENTATION 3/4.'3.2 ENGINEERED'SAFETY'FEATURE'ACTUATION SYSTEM INSTRUMENTATION L IMITING." CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instru-mentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a. With an ESFAS instrumentation channel trip setpoint less conservative than the value 'shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the appl.icable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEIL'LANCE 'E UIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of. each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the "Total No. of Channels" Column of Table 3~3 3 ~

ST. LUCIE - UNIT 1 3/4 3-9

TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION SAFETY, INJECTION (SIAS)

a. Manual (Trip Buttons) 1, 2, 3, 4

b. -

Containment Pressure-Hi gh 1, 2, 3

c. Pressurizer Pressure-Low 1, 2(d), 3(a) 9

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) 1, 2, 3, 4

b. Containment Pressure--

High - High 2(b) 1, 2, 3 10

3. CONTAINMENT ISOLATION (CIS)

a. Manual (Trip Buttons) 2 1, 2, 3, 4

b. Containment Pressure-High 4 1, 2, 3

c. Containment Radiation-High 4 1, 2, 3, 4

4. MAIN STEAM LINE ISOLATION (MSIS)

a. Manual (Trip Buttons) 2/steam 1/steam 2/operating 1, 2, 3, 4 generator generator steam generator

b. Steam Generator 4/steam 2/steam 3/steam 1, 2, 3(c)

Pressure - Low generator generator generator

TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. CONTAINMENT SUMP RECIRCULATION (RAS)

a. Manual RAS (Trip 8 Buttons) 1, 2, 3, 4

b. Refueling Water Tank-Low 1, 2, 3

6. LOSS OF POWER 4.16 kv Emergency Bus Undervoltage (Under-voltage relays) 1/Bus 1/Bus 1/Bus 1>> 2. 3

TABLE 3.3-3 Continued TABLE NOTATION (a) Trip function may be bypassed in this MODE when pressurizer pressure is < 1725 psia; bypass shall be automatically removed when pressurizer pressure is > 1725 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed in this MODE below 585 psig; bypass shall be automatically removed at or above 585 psig.

(d) Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3.

ACTION STATEMENTS ACTION 8 With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within'8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD'HUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 9 - With the number of OPERABLE channels one less than the Total Number of Channels and with the pressurizer pressure:

a. < 1725 psia; place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after increasing the pressurizer pressure above 1725 psia; otherwise, be in at least HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. > 1725 psia, operation may continue provided all of the following conditions are satisfied:

1. The inoperable channel is placed in the tripped condition within 1 hour.

2. All functional units receiving an input from the tripped channel are also placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

3. The-Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided one of the inoperable channels is placed in the tripped condition.

ST. LUCIE - UNIT 1 3/4 3-12

TABLE 3.3-3 Continued TABLE NOTATION ACTION 10 - With the number of OPERABLE channels one less than the Total Number of Channels and with the pressurizer pressure:

a. ( 1725 psia; place the inoperable channel in the condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and restore the bypassed inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after increasing the pressurizer pressure above 1725 psia; otherwise, be in at least HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. > 1725 psia, demonstrate that the Minimum Channels OPERABLE requirement is met within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; operation may continue with the inoperable channel bypassed and one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

ST. LUCIE UNIT 1 3/4 3-13

TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES.

1. SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Containment Pressure - High < 5 psig < 5 psig

c. Pressurizer Pressure - Low > 1600 psia > 1600 psia

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Containment Pressure -- High-High < 10 psig < 10 psig

3. CONTAINMENT ISOLATION (CIS)

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Containment Pressure - High < 5 psig < 5 psig

c. Containment Radiation - High < 10 R/hr < 10 R/hr

4. MAIN STEAM LINE ISOLATION (MS IS)

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Steam Generator Pressure - Low > 485 psig > 485 psig

5. CONTAINMENT SUMP RECIRCULATION (RAS)

a. Manual RAS (Trip Buttons) Not Applicable Not Applicable

b. Refueling Water Tank - Low 48 inches above 48 inches above tank bottom tank bottom

TABLE 3.3-4 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ALLOWABLE FUNCTIONAL UNIT TRIP VALUE VALUES

6. LOSS OF POWER 4.16 kv Emergency Bus Undervoltage (Undervoltage relays) > 3307 volts > 3307 volts

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual

a. SIAS Safety Injection (ECCS) Not Applicable Containment Fan Coolers Not Applicable Feedwater Isolation Not Applicable

b. CSAS Containment Spray Not Applicable

c. CIS Containment Isolation Not Applicable Shield Building Ventilation System Not Applicable

d. RAS e.

Containment MSIS Sump Recirculation Isolation Not Applicable Not Applicable 0

Main Steam

2. Pressurizer Pressure-Low

a. Safety Injection (ECCS) < 30.0*/19.5**

b. Containment Fan Coolers < 30 '*/17.0**

c. Feedwater Isolation < 60.0

3. Containment Pressure-Hi h

a. Safety Injection (ECCS) < 30.0*/19.5**

b. Containment Isolation < 30.5*/20.5**

c. Shield Building Ventilation System < 30.0*/14.0**

d. Containment Fan Coolers < 30.0*/17.0+*

e. Feedwater Isolation < 60.0 ST. LUCIE - UNIT 1 3/4 3-16

TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

4. Containment Pressure--Hi h-Hi h

a. Containment Spray < 30.0*/18.5**

5. Containment Radiation-Hi h

a. Containment Isolation < 30.5*/20.5**

b. Shield Building Ventilation System < 30.0*/14.0*~

6. Steam Generator Pressure-Low

a. Main Steam Isolation < 6.9

7. Refuelin Water Stora e Tank-Low.

8. ~ii a.

a.

Containment Sump Recirculation Feedwater Flow Reduction to 5A 91.5 60.0 TABLE NOTATION Diesel generator starting and sequence loading delays included.

Diesel generator starting

~

and sequence loading delays not included.

Offsite pow'er available.

ST. LUCIE - UNIT 1 3/4 3-17

TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) N.A. N.A. R N.A.

b. Containment Pressure - High S R M 1, 2, 3

c. Pressurizer Pressure - Low S R M 1, 2,.3

d. Automatic Actuation Logic N.A. N.A. M(1) 1,2,3

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) N.A. N.A. N.A ~

b. Containment Pressure High - High S R M 1 ~ 2 3 c, Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3

3. CONTAINMENT ISOLATION (CIS)

a. Manual (Trip Buttons) N.A. N.A. R N.A.

b. Containment Pressure - High S R 1, 2, 3

c. Containment Radiation - High S R 1, 2, 3, 4

d. Automatic Actuation Logic N.A. N.A. M(1) 1 2 3

4. MAIN STEAM LINE ISOLATION (MSIS)

a. Manual (Trip Buttons) N.A. N.A. R N.A.

b. Steam Generator Pressure - Low S R 1, 2, 3

c. Automatic Actuation Logic N.A. N.A. M(1) 1 2 3

5. CONTAINMENT SUMP RECIRCULATION (RAS)

a. Manual RAS Trip Buttons) N.A. N.A. N.A.

b. Refueling Water Storage Tank - Low S R 1,2,3

c. Automatic Actuation Logic N.A. N.A. M(1) 1 y 2$ 3

TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST RE UIRED FUNCTIONAL UNIT

6. LOSS OF POWER 4.16 kv Emergency Bus Undervoltage (Undervoltage relays) 1, 2, 3

TABLE 4.3-2 Continued TABLE NOTATION 0

(1) The logic circuits shall be tested manually at least once per 31 days.

ST. LUGIE - UNIT 1 3/4 3-20

INSTRUMENTATION 3/4.3. 3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a. With a radiation monitoring channel alarm setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK,'HANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3.

ST. LUCIE - UNIT 1 3/4 3-21

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

l. AREA MONITORS 10 - 10 mR/hr 13

a. Fuel Storage Pool Area < 15 mR/hr

b. Containment (CIS) < 90 mR/hr 1 - 10 5

mR/hr 16

2. PROCESS MONITORS

a. Containment Gaseous Activity -

RCS Leakage Detection 1 1, 2, 3 & 4 Not Applicable 10 10 cpm 14 ii. Particulate Activity RCS Leakage Detection 1 1, 2, 3 & 4 Not Applicable 10 - 10 cpm/hr 14

With fuel in the storage pool or building

TABLE 3.3-6 (Continued TABLE NOTATION ACTION 13- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 14- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 16- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.

ST. LUCIE

~

- UNIT 1 3/4 3-23

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST RE UIRED

1. AREA MONITORS

a. Fuel Storage Pool Area M-

b. Containment (CIS)

2. PROCESS MONITORS

b. Containment

i. Gaseous Activity RCS Leakage Detection 1, 2, 3, 5 4 ii. Particulate Activity RCS Leakage Detection 1, 2, 3, & 4

With fuel in the storage pool or building

INSTRUMENTATION 0 INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The incore detection system shall be OPERABLE with:

a ~ At least 75$ of all incore detector locations, and h

b. A minimum of two quadrant symmetric incore detector locations per core quadrant.

An OPERABLE incore detector location shall consist of a fuel assembly containing a fixed detector string with a minimum of three OPERABLE rhodium detectors.

APPLICABILITY: When the incore detection system is used for:

a. Recalibration of the excore axial flux offset detection system,

b. Monitoring the AZIMUTHAL POWER TILT,

c. Calibration of the power level neutron flux channels, or

d. Monitoring the linear heat rate.

ACTION:

With the incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.2 The incore detection system shall be demonstrated OPERABLE:

a ~ By performance of a CHANNEL CHECK within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to its use and at least once per 7 days thereafter when required for:

1. Recalibration of the excore axial flux offset detection system,

2. Monitoring the linear heat rate pursuant to Specification 4.2.1.3, ST. LUCIE - UNIT 1 3/4 3-25

INSTRUMENTATION SURVEILLANCE RE UIREMENTS Continued

3. Monitoring the AZIMUTHAL POWER TILT, or

4. Calibration of the Power Level Neutron Flux Channels.

b. At least once per 18 months by performance of a CHANNEL CALI-BRATION operation which exempts the neutron detectors but in-cludes all electronic components. The neutron detectors shall be calibrated prior to installation in the reactor core.

ST. LUCIE - UNIT 1 3/4 3-26

INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation channels shown in Table 3.3-7 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a. With the number of OPERABLE seismic monitoring channels less than required by Table 3.3-7, restore the inoperable chan-nel(s) to OPERABLE status within 30 days.

b. With one or more seismic monitoring channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the system to OPERABLE status.

c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.3.1 Each of the above seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-4.

4.3.3.3.2 Each of the above seismic monitoring instruments actuated during a seismic event shall be restored to OPERABLE status and a CHANNEL CALIBRATION performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety.

ST. LUCIE - UNIT 1 3/4 3-27

TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT CHANNELS INSTRUMENT CHANNEL SENSOR LOCATION RANGE OPERABLE

1. STRONG MOTION TRIAXIAL ACCELEROGRAPHS a 0 SMR-42-1 R.B. Elev. 23.0'.B.

g

b. SMR-42-2 Elev. 0-1 g C.

d.

SMR-42-3 SMR-42-4 Elev.

62.0'.A.B.

Elev.

43.0'-1

-0.5'.A.B.

0-1 g 0-1 g

2. PEAK RECORDING ACCELEROGRAPHS

a. SMR-42-6 R.B. Piping from S.I.T.lA2-c Elev.

46'0 9/16" 0-2 g

b. SMR-42-7 R.B. Equipment on S. I .T.1A2 0-2 g

c. SMR-42-8 R.A.B.-Sh. Dn. Ht.

XCHR Supports 0-2 g

3. PEAK SHOCK RECORDERS

a. SMR-42-9 R.B. Elev.

b.-

23.0'.B.

SMR-42-10 M.S. Pipe Restraints - S.G.lBl

4. EARTHQUAKE FORCE MONITOR

a. SMI-42-11 Control Room 0-0.2 g

5. SEISMIC SWITCH

a. SMS-42-12 R.B. Elev.

23.0'T.

LUCIE - UNIT 1 3/4 3-28

I TABLE 4.3-4 n

foal SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHANNEL CHECK CALIBRATION TEST

1. STRONG MOTION TRIAXIAL ACCELEROGRAPHS

a. SMR-42-1 SA

b. SMR-42-2 SA

c. SMR-42-3

d. SMR-42-4 'A SA

e. SMR-42-5 SA

2. PEAK RECORDING ACCELEROGRAPHS

a. SMR-42-6 N.A. R N.A.

b. SMR-42-7 N.A. R N.A.

c. SMR-42-8 N.A. R N.A.

3. PEAK SHOCK RECORDERS

a. SMR-42-9 N.A. N.A.

b. SMR-42-10 N.A. N.A.

4. EARTHQUAKE FORCE MONITOR

a. SMI-42-11 SA

5. SEISMIC SWITCH

a. SMS-42-12 N.A. SA

INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 0

3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a ~ With the number of OPERABLE meteorological monitoring chan-nels less than required by Table 3.3-8, suspend all release of gaseous radioactive material from the radwaste gas decay tanks unti 1 the inoperable channel(s) is restored to OPERABLE status.

b. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel(s) to OPERABLE status.

c ~ The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.4 Each meteorological monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHAN-NEL CALIBRATION operations at the frequencies shown in Table 4.3-5.

ST. LUCIE - UNIT 1 3/4 3-30

~ Table 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT MINIMUM MINIMUM CHANNELS INSTRUMENT LOCATION ACCURACY OPERABLE

1. WIND SPEED

a. Nominal Elev. 32.8 ft. + 0.5 mph*

b. Nominal Elev. 190 ft. + 0.5 mph*

2. WIND DIRECTION

a. Nominal Elev. 32.8 ft. 50

b. Nominal Elev. 190 ft. 50 3.

~ AIR TEMPERATURE DELTA T a.~

~

Nominal Elev. 32.8

~ ~ ft.~ + 0.18'F

b. Nominal Elev. 110 ft. + 0.18'F

c. Nominal Elev. 200 ft. + 0.18'F Starting speed of anemometer shall be < 1 mph.

ST. LUCIE - UNIT 1 3/4 3-31

TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK .CALIBRATION

1. WIND SPEED

a. Nominal Elev. 32.8 ft. SA

b. Nominal Elev. 190 ft. SA

2. WIND DIRECTION

a. Nominal Elev. 32.8 ft. SA

b. Nominal Elev. 190 ft. SA

3. AIR TEMPERATURE - DELTA T a; Nominal Elev. 32.8 ft. SA

b. Nominal Elev. 110 ft.

c. Nominal Elev. 200 ft. SA

INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, either:

a ~ Restore the inoperable channel to OPERABLE status within 30 days, or

b. Be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel, shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

ST. LUCIE - UNIT 1 3/4 3-33

TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION MINIMUM READOUT MEASUREMENT CHANNELS INSTRUMENT LOCATION RANGE OPERABLE

1. Reactor Trip Breaker Indication SWGR OPEN-CLOSE 1/trip breaker

2. Pressurizer Pressure Hot. Shutdown Panel 1500-2500 psia

3. Pressurizer Level Hot Shutdown Panel 0-100K

4. Main Steam Pressure Hot Shutdown Panel 0-1200 psig 1/steam'enerator

5. Steam Generator Level Hot Shutdown Panel 0-100K 1/steam generator

6. Cold Leg Temperature Hot Shutdown Panel 0-600'F

TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

l. Reactor Trip Breaker Indication N.A.

2. Pressurizer Pressure

3. Pressurizer Level

4. Steam Generator Level

5. Main Steam Pressure

6. Cold Leg Temperature

INSTRUMENTATION CHLORINE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.3.3.6 Two separate and independent chlorine detection systems, with their alarm/trip setpoints adjusted to actuate at a chlorine concentra-tion of ( 5 ppm, shall be OPERABLE with each chlorine detection system having at least one chlorine detector in each control room outside air intake duct.

APPLICABILITY: ALL MODES ACTION:

a ~ With one chlorine detector inoperable, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either isolate the associated outside air intake or initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation until the inoperable chlorine detector is restored to OPERABLE status.

b. With no chlorine detection system'PERABLE, initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation until two chlorine detection systems are restored to OPERABLE status.

C. The provisions'f Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS.

4.3.3.6 Each chlorine detection system shall be verified energized at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and demonstrated OPERABLE by performance of a CHANNEL CALIBRATION at least once per 18 months.

ST. LUG I E - UNIT 1 3/4 3-36

3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.4.1 Four reactor coolant pumps shall be in operation.

APPLICABILITY: As noted below, but excluding MODE 6.*

ACTION:

MODES 1 and 2:

With less than four reactor coolant pumps in operation, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 3, 4 and 5:

Operation may proceed provided at least once reactor coolant loop is in operation with an associated reactor coolant pump or shutdown cooling pump; however, operation for up to 15 minutes with no pump in operation is permissible to accommodate transition between shutdown cooling pump and reactor coolant pump operation. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.1 The Flow Dependent Selector Switch shall be determined to be in the 4 pump position within 15 minutes prior to making the reactor critical and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

See Special Test Exception 3.10.4.

ST. LUCIE - UNIT 1 3/4 4-1

REACTOR COOLANT SYSTEM SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION lift settingof one 3.4.2 A minimum pressurizer code safety valve shall be OPERABLE with a of 2500 PSIA + 15.

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and'lace an OPERABLE shutdown cooling loop into operation.

SURVEILLANCE RE UIREMENTS 4.4.'2 The pressurizer code safety valve shall be demonstrated OPERABLE per Sur veil lance Reguirement 4.4.3.

ST. LUCIE - UNIT 1 3/4 4-2

REACTOR COOLANT SYSTEM SAFETY VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2500 PSIA + 1/.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.4.3 Each pressurizer code safety valve shall be demonstrated OPERABLE with a lift setting of 2500 PSIA + 1%, in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition.

ST. LUCIE - UNIT 1 3/4 4-3

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a,steam bubble.

APPLICABILITY: MODES 1 and 2.

ACTION:

With the pressurizer inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.4.4 Not applicable.

ST. LUCIE - UNIT 1 3/4 4-4

REACTOR COOLANT SYSTEM STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more steam generators inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing T 'above 200'F ~

avg SURVEILLANCE RE UIREMENTS 4.4.5.1 Steam Generator Sam le Selection and Ins ection Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sam le Selection and Ins ection The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.

The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. Steam generator tubes shall be examined in accordance with Appendix IV of the ASME Boiler and Pressure Vessel Code - Section XI - "Inservice Inspection of Nuclear Power Plant Components" 1974 Edition and Addenda through Summer 1976. The tubes selected for each inservice inspection shall include at least 3X of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 505 of the tubes inspected shall be from these critical areas.

b. The first inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

l. All nonplugged tubes that previously had detectable wall penetrations ()20/)', and

2. Tubes in those areas where experience has indicated potential problems.

ST. LUG I E - UNIT 1 3/4 4-5

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued

c. The second and third inservice 'inspections may be less than a full tube inspection by concentrating (selecting at least 50%

of the tubes to the inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three. categories:

~Cate or Ins ection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Ins ection Fre uencies The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed. at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecu-tive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

ST. LUCI E - UNIT 1 3/4 4-6 0

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued)

b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection whose results fall in Category C-3, the inspection frequency shall be reduced to at least once per 20 months. The reduction in inspection frequency shall apply until a subsequent inspection demonstrates that a third sample inspection is not required.

c~ Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions.

1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,

2. A seismic occurrence greater than the Oper'ating Basis Earthquake,

3. A loss-of-coolant accident requiring actuation of the engineered safeguards, or

4. A main steam line or feedwater line break.

4.4.5.4 Acce tance Criteria a ~ As used in this Specification:

f 4 df 2 d or contour of a tube from that required by fabrication f,f4 f4 drawings or specifications. Eddy-current. testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2. ~di " 4 -fd wear or general corrosion occurring on d . bfg, g either inside or

~ddfbb outside of of the a tube.

ff fp f nominal wall thickness caused by degradation.

4 dbg

4. ~gd thickness 4

affected or 4 p g f 2 removed by degradation.

b ST.~ LUCIE - UNIT 1 3/4 4-7 d

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued

5. Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.-

6. ~61 I 111 <<I fp f I dpd 6 6 which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness.

7. Unserviceable describes df the condition of a tube if it I g gd ff structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3;c, above.

6. ~Tb I I I 6 I 1 1 g tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 ~Re orts a ~ Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Comission within 15 days.

b. The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

ST. LUCI E - UNIT 1 3/4 4-8

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued c ~ Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

ST. LUCIE - UNIT 1 3/4 4-9

TABLE 4.4-1 MINIMUMNUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Two Three Four Two Three Four First Inservice Inspection All One Two Two Second 5 Subsequent Inservice Inspections One" One1 One2 One3 Table Notation P

1. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the'operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circum-stances the sample sequence shall be modified to inspect the most severe conditions.

2. The other steam generator not inspected during the first inservice inspection shall be inspected. The third and subsequent inspections should follow the instructions described in 1 above.

3. Each of the other two steam generators'not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above.

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C 1 None N/A N/A N/A N/A S Tubes per S. G.

C-2 Plug defective tubes C-1 None N/A N/A and inspect additional Plug defective tubes C 1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G.

Perform action for C 3 C-3 result of first sample Perform action for C-3 C-3 result of first N/A N/A sample C-3 Inspect all tubes in All other this S. G., plug de- S. G.s are None N/A N/A fective tubes and C-1 inspect 2S tubes in Some S. G.s Perform action for each other S. G.

N/A N/A C 2 but no C-2 result of second additional sample Prompt notification S. G. are to NRC pursuant C-3 to specification Additional Inspect all tubes in 6.9.1 S. G. is C-3 each S. G. and plug defective tubes.

Prompt notification N/A N/A to NRC pursuant to specification 6.9.1 Where N is the number of steam generators in the unit, and n is the number of steam generators inspected

$ 3 / during an inspection

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a ~ A containment atmosphere particulate, radioactivity monitoring system,

b. The reactor cavity sump level and flow monitoring system, and c ~ A containment atmosphere gaseous radioactivity monitoring system.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a ~ Wi th one of the above r equired radi oacti vi ty moni toring 1 eakage detection systems. inoperable, operations may continue for up to 30 days provided:

1. The other two above required leakage detection systems are OPERABLE, and

2. Appropriate grab samples are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, be i'n at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours.

b. With both the above required radioactivity monitoring leakage detection systems inoperable, operations may continue for up to 30 days provided:

1. The reactor cavity sump level and flow monitoring system is OPERABLE,

2. Appropriate grab samples are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 3; A Reactor Coolant System water inventory balance is per-formed at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during steady state operation except when operating in the shutdown cooling mode; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ST. LUCIE UNIT 1 3/4 4-12

REACTOR COOLANT SYSTEM LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION Continued c ~ With the containment sump level and flow monitoring system in-operable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:

. a. Containment atmosphere gaseous and particulate monitoring systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and

b. Reactor cavity sump level and flow monitoring system-performance of CHANNEL CALIBRATION TEST at least once per 18 months.

ST. LUG I E - UNIT 1 3/4 4-13

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through steam generators, alid

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous and particulate radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

c. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state opera-tion except when operating in the shutdown cooling mode, and

d. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ST. LUG IE - UNIT 1 3/4 4-14

REACTOR COOLANT, SYSTEM CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-1.

APPLICABILITY: ALL NODES.

ACTION'ODES 1, 2, 3 and 4 a ~ With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the parameter to within its Steady State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY wi thin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6 With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of its Transient Limit, reduce the pressurizer pressure to ( 500 psia, if applicable, and perform an analysis to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations prior to increasing the pressurizer pressure above 500 psia or prior to proceeding to MODE 4.

SURVEILLANCE RE UIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3.

ST. LUG IE UNIT 1 3/4 4-15

TABLE 3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT DISSOLVED OXYGEN < 0.10 ppm* < 1.00 ppm*

CHLORIDE < 0.15 ppm < 1.50 ppm FLUORIDE < 0.10 ppm < 1.00 ppm

Limit not applicable with T < 250'F.

avg TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE RE UIREMENTS MINIMUM MAXIMUM TIME PARAMETER SAMPLING FRE UENCIES BETWEEN SAMPLES DISSOLVED OXYGEN 3 times per 7 days* 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months CHLORIDE 3 times per 7 days 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months FLUORIDE 3 times per 7 days 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Not required with Tavg< 250'F.

ST. LUCIE - UNIT 1 3/4 4-16

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a. < 1.0 pCi/gram DOSE EQUIVALENT I-131, and

b. < 100/E pCi/gram.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1, 2 and 3*:

a ~ With the specific activity of the primary coolant > 1.0 pCi/gram DOSE EQUIVALENT I-131 but within the allowable limit {below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months provided that operation under these circumstances shall not exceed 10 percent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

b. With the specific activity of the primary coolant > 1.0 yCi/gram DOSE EQUIVALENT I-131 for more than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months during one con-tinuous time interval or exceeding the limit line shown on Figure 3.4-1, be in KOT STANDBY with T 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

avg C. With the specific activity of the primary coolant > 100/E pCi/gram, be in HOT STANDBY with T v < 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 1, 2, 3, 4 and 5:

d. With the specific activity of the primary coolant > 1.0 pCi/gram DOSE EQUIVALENT I-131 or > 100/E pCi/gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. A REPORTABLE OCCURRENCE shall be prepared and submitted to the Comnission pursuant to Specification 6.9.1. This report shall contain the results of the specific activity analyses together with the following information:

l. Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, With Tav > 500'F ST. LUCIE - UNIT 1 3/4 4-17

REACTOR COOLANT SYSTEM ACTION: Continued

2. Fuel burnup by core region,

3. Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded,

4. History of de-gassing operation, if any, starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, and

5. The time duration when the specific activity of the pri-mary coolant exceeded 1.0 pCi/gram DOSE E(UIVALENT I-131.

SURVEILLANCE RE UIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis pro-gram of Table 4.4-4.

ST. LUCI E -

o UNIT 1 3/4 4-18

TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT MINIMUM MODES IN WHICH SAMPLE AND ANALYSIS FRE(RUENCY AND ANALYSIS RE UIRED

1. Gross Activity Determination 3 times per 7 days with a 1, 2, 3 and 4 maximum time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples

2. Isotopic Analysis for DOSE 1 per 14 days EQUIVALENT I-131 Concentration

3. Radiochemical for E Determination 1 per 6 months

4. Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 1 2 3 4 and 5 Including I-131, I-133, and I-135 whenever the DOSE EQUIVALENT I-131 exceeds 1.0 pCi/gram, and b) One sample between 2 1 2.3 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

Until the specific activity of the primary coolant system is restored within its limits.

After at least 2 EFPD and at least 20 days since the last shutdown of longer than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

E u 250 I

i 0"

i 200 UNACCEPTABLE OPERATION CD CD Ul CO i- 1S0 0

CD 0-CC 100 ACCEPTAB LE OPERATION so 8

30 40 60 60 . 70. 80 90'00 PERCENT OF RATED THERMAL POWER FIGURE 3'.4-'I'OSE EQUIVALENTI-13'I Primary Coolant Specific Activity Limit Versus Percept of RATED THERMAL POWER with the Primary Coolant Specific Activity>'1.0p,Ci/gram Dose Equivalent I-131 ST. LUCIE - UNIT 1 3/4 4-2O 0

REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100'F in any one hour period,

b. A maximum cooldown of 100'F in any one hour period, and

c. A maximum temperature change of O'F in any one hour period, during hydrostatic testing operations above system design pressure.

APPLICABILITY: At all times.*

ACTION:

With any of the above 1 imi ts exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an analysis to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T and pressure to less than 200'F and 500 psia, respectively, within tQ following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

See Special Test Exception 3.10.3.

ST. LUG IE - UNIT 1 3/4 4-21

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS 4.4.9.1

a. The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

b. The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of. the criticality limit line within 15 minutes prior to achieving reactor criticality.

c. The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5. The results of these examinations shall be used to update Figure 3.4-2.

0 ST. LUCIE - UNIT 1 3/4 4-22

3600

,MINIMUMPRESSURE TEMPERATURE FOR:

0 CRITICALOPERATION OF CORE 0% NON CRITICALOPERATION OF CORE 3200 SYSTEM HYDRO LOWEST SERVICE TEMPERATURE 2800 HEATUP"'A%P COOLDOWN"i HEATUP0 COOLDOWNi 2400 K

D 2000 N

K D

a. 1600 O

I O

O a

1200 NOTE 1 REACTOR VESSEL BELTLINE MATERIAL INITIALRTNoT ~ 5 F NOTE 2 REACTOR VESSEL BELTLINE MATERIAL 800 2YEAR RTgoT SHIFT~55 F NOTE 3- REMAINING PRESSURE BOUNDARY MAXIMUMRTNDT ~ 50 F 400 MAXIMUMPRESSURE FOR 6 DC OPE RATION 100 200 300 400 500 INDICATED REACTOR COOLANT TEMPERATURE TK F FIGURE 3A-2a Reactor Coolant System Pressure Temperature Limitations for 0 to 2 Years of Full Power Operation ST. LUCIE - UNIT 1 3/4 4-23a

3600 MINIMUMPRESSURE TEMPERATURE FOR:

3200

"~ CRITICAL OPERATION OF CORE NON-CRITICALOPERATION OF CORE LOWEST SERVICE SYSTEM HYDRO TEMPERATURE 2800 HEATUP~~

COOLDOWNi" HEATUPi COOLDOW 2400 K

D ug 2000 N

lC D

~16OO

~

CL O

I 5

O z

1200 NOTE 1 REACTOR VESSEL BELTLINE MATERIAL 800 INITIALRTNDT ~ 5 F NOTE 2 REACTOR VESSEL BELTLINE MATERIAL 10YEARRTNoTSHIFT 135 F NOTE 3- REMAINING PRESSURE BOUNDARY MAXIMUMRTNDT ~50 F MAXIMUMPRESSURE FOR SDC OPERATION 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE Tc F FIGURE 3.4-2b Reactor Coolant System Pressure Temperature Limitations for 2 to 10 Years of Full Power Operation ST. LUCIE - UNIT 1 3/4 4-23b

3600 NOTE 1 REACTOR VESSEL BELTLINE MATERIAL INITIALRTIIor 6 F NOTE 2- REACTOR VESSEL BELTLINE MATERIAL 40 YEAR RTIIoT SHIFT ~ 225 F 3200 NOTE 3 REMAINING PRESSURE BOUNDARY MAXIMUMRTgoT 50 F LOWEST SERVICE SYSTEM HYDRO TEMPERATURE 2800 HEATUP""

COOLDOWN""

2400 HEA UPi K

g 2000 LDOW Ni CC M

K rc 1600 O

I O

Z 1200 800 MINIMUMPRESSURE TEMPERATURE FOR:

0 CRITICAL OPERATION OF CORE MAXIMUM PRESSURE oo NON.CRITICALOPERATION OF CORE FOR SD C OPERATION 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE T~ F FIGURE 3.4-2c Reactor Coolant System Pressure Temperature Limitations for 10 to 40 Years of Full Power Operation ST. LUCIE - UNIT 1 3/4 4-23c

TABLE 4.4-5 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE SPECIMEN REMOVAL INTERVAL 8 years

2. 16 years

3. 23 years

4. 30 years

5. 35 years

6. 40 years

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

a. A maximum heatup of 100'F in any one hour period,

b. A maximum cooldown of 200'F in any one hour period, and

c. A maximum Reactor Coolant System spray water temperature differential of 350'F.

APPLICABILITY: At al 1 times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an analysis to determine the effects of the out-of-limit con-dition on the fracture toughness properties of the pressurizer; deter-mine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psia within, the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.

The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady state operation.

ST. LUCIE - UNIT 1 3/4 4-25

REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY SAFETY CLASS 1 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of components (except steam generator tubes) identified in Section 3.2.2 of the FSAR as Safety Class 1 com-ponents shall be maintained at a level consistent with the acceptance criteria in Specification 4.4. 10. l.

APPLICABILITY: MODES 1, 2, 3 and 4:

ACTION:

With the structural integrity of any of the above components not conform-ing to the above requirements, restore the structural integrity of the affected component to within its limit or isolated the affected component prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations. The provisions of Specification 3. 0. 4 are not applicable.

SURVEILLANCE, RE UIREMENTS 4.4.10.1 The following inspection program shall be performed during shutdown:

a ~ Inservice Ins ections The structural integrity of the Safety Class 1 components shall be demonstrated by verifying their acceptability when inspected per the applicable requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Winter 1972, as outlined by the inspection program shown in Table 4.4-6.

An initial report of any abnormal degradation of the structural integrity of the Safety Class 1 components detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be submitted pursuant to Specification 6.9. 1 within 90 days after completion of the surveillance requirements of this specification.

ST. LUCIE - UNIT 1 3/4 4-26

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued The Inservice Inspection Program shall be reviewed every 5 years to assure that the equipment, techniques and procedures being utilized are current and applicable. The results of these reviews shall be reported in Special Reports to the Coranission pursuant to Specification 6.9.2 within 90 days of completion.

b. Ins ections Followin Re airs or Re lacements The structural integrity of the reactor coolant system shall be demonstrated after completion of all repairs and/or replacements to the system by verifying the repairs and/or replacements meet the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Winter 1972.

When repairs and/or replacements are made which involve new strength welds on components greater than 2 inch diameter, the new welds shall receive a surface and 100 percent volumetric examination and meet applicable code requirements. When re-pairs and/or replacements are made which involve new strength welds on components 2 inch diameter or smaller, the new welds shall receive a surface examination and meet applicable code requirements.

Co Ins ections Followin S stem 0 enin The structural integrity of the reactor coolant system shall be demonstrated after each closing by performing a leak test, with the system pressurized to at least 2235 psig, in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Winter 1972, and the Pressure/Temperature limits of Specification 3.4.9.1.

ST. LUCIE - UNIT 1 3/4 4-27

TABLE 4.4<

INSERVICE INSPECTION PROGRAM - SAFETY CLASS 1 COMPONENTS SECTION 1. REACTOR VESSEL AND CLOSURE HEAD Tentative Inspection Item Examination ,During 10-year eo. (b) ~Cate or b Examination Area(b) ~Method b Interval Remarks 1.1 A Longitudinal and Cir- Volumetric See Note (a) and Remarks The required examinations will be performed cumferential shell welds at or near the end of the 10-year inspection in core region. interval or when the internals are removed for other reasons. Mechanized Ultrasonic ex-amination will be used.

1.2 Longitudinal and cir- Volumetric See Note (a) and Remarks (1) The required amount of weld lengths will cumferential welds in be examined at or near the end of the shell (other than those 10-year inspection interval or when the of Category A and C). internals are removed for other reasons.

(2) Closure head meridional welds and dome (circumferential) welds which are not accessible inside the head cooling shroud assembly will be excluded from the examination.

(3) A partial examination of the RPV lower head welds can be done from the ID with mechanized techniques.

a) Approximately 25$ of meridional welds.

b) Approximately 75% of circumferential weld (lower head to shell).

c) 100$ of the dome welds (circumferen-tial).

. examination of RPV lower head (4) 100% manual welds will be performed externally during baseline.

TABLE 4.4-6 (Cont'd)

SECTION 1. REACTOR VESSEL AND CLOSURE HEAD Tentative Inspection Item Examination During 10-year No.(b) ~Cate or (b) Examination Area(b) ~Method b Interval Remarks 1.3 Vessel-to-flange cumferentiall and Volumetric See Note (a) and Remarks Both of these welds are available for ex-head-to-flange cir- amination during normal refueling operations.

welds.

1.4 Primary nozzle-to-vessel Volumetric See Note (a) and Remarks The required examinations will be performed welds and nozzle-to-vessel by mechanized techniques as follows:

inside radiused sections.

(1) Hhenever the internals are removed, one (1) outlet and one (1) inlet nozzle during the first two thirds of this interval (not to exceed 2 each). Balance per remark (2) below.

(2) In case the internals are not removed, two (2) outlet nozzles for the first two thirds of the interval and four (4) inlet nozzles near or at the end of the interval.

1.5 E-1 Vessel penetrations, in- Volumetric See Remarks (1) The CROM tubes and in-core instrumentation cluding control rod drive tubes are welded to the upper head with a penetrations and control partial penetration weld. The assemblies rod housing pressure contain an integrally welded thermal sleeve boundary welds. Volumetric examination is not feasible by currently available techniques. These penetrations are included in Category E-2.

(2) Housing pressure boundary welds are in-accessible due to head cooling shroud assembly.

(3) Meets exclusion criteria of paragraph IS-121(a) ref: FSAR 15.4.5.

1.6 E-2 Vessel penetrations, in- Visual See Note (a) cluding control rod drive penetrations and control rod housing pressure boundary welds.

TABLE 4.4-6 (Cont'd)

SECTION 1. REACTOR VESSEL AND CLOSURE HEAD Tentative Inspection Item Examination During 10-year e leo.(b) ~Ceto or (b) Examination Area(b) ~tletbod b Interval Remarks 1.7 F Primary nozzles to safe Visual, surface See Remarks There are no welds in this category.

end welds. and volumetric.

le8 G-1 Closure studs and nuts. Visual. surface See Note (a) and Remarks Surface examination does not apply to and volumetric nuts.

1.9 G-1 Ligaments between Volumetric See Note (a) and Remarks The ligaments will be examined whenever threaded stud holes. the Reactor Vessel Flange Weld of Item 1.3 is examined.

1.10 G-1 Closure washers, bushings. Visual See Note (a) There are no bushings.

G-2 Pressure retaining Visual See Remarks There is no bolting less than 2-inches in bolting. diameter accessible for examination.

Integrally welded Volumetric See Remarks Nozzle-type pad supports are installed and vessel supports. are examined under Item 1.4, Category D.

1.13 Closure head cladding Visual and See Note (a) and Remarks At least six (6) patches (each 36 sq. in.)

surface or of the RPV Closure Head will be examined-volumetric visual and surface.

1.14 Vessel cladding. Visual See Note (a) and Remarks At least six patches (each 36 sq. in.) of the vessel will be examined by remote techniques.

1.15 Interior surfaces and Visual See Note (a) and Remarks (1) The examinations will be done by remote internals and integrally visual techniques.

welded internal supports.

(2) Whenever refueling occurs and internals are removed, a remote visual examination will be performed on the vessel interior.

TABLE 4.4-6 (Cont'd)

SECTION 2. PRESSURIZER Tentative Inspection Item Examination During 10-year

~No. b ~Cote or b Examination Area b ~Nethod b Interval Remarks 2.1 B Longitudinal and Visual and See Note (a)

Circumferential Welds Volumetric 2.2 Nozzle-to-Vessel Welds Volumetric See Note (a) and nozzle-to-vessel inside radius section 2.3 E-1 Heater Connections Visual and See remarks (1) Meet the exclusion criteria of IS-121 Surface (c) and are examined under category E-2.

2.4 E-2 Heater Connections Visual See Note (a) 2e5 G-1 Pressure Retaining Visual and See Remarks (1) Bolting 2-inch in diameter and greater Bolting Volumetric do no exist.

2.6 G-2 Pressure retaining Visual See Note (a) and Remarks The bolting less that 2-inches in diameter will Bolting be visually inspected. either in-place under tension or when they are disassembled.

TABLE 4.4-6 (Cont'd)

SECTION 2. PRESSURIZER Tentative Inspection Item Examination During 10-year

~Method h Interval Remarks 2.7 Integrally welded Visual and See Note (a) vessel supports. Volumetric 2.8 I-2 Vessel cladding. Visual See Note (a) and Remarks At least one (1) patch (36 sp. in.) below the primary manway will be examined.

SECTION 3. STEAN GENERATORS 3.1 Longitudinal and cir- Visual and See Note (a) and Remarks Hechanized examination, no visual, will be cumferential welds, volumetr ic performed on three (3) each staywell circum-including tube sheet- ferential welds due to limited access and undue to-head welds on the radiation exposure to personnel.

primary side.

3.2 Primary nozzle-to- Volumetric See Hot (a) head welds and nozzle-to-head inside radi-used sections.

3.3 Primary nozzle-to-safe Visual, surface See Remarks Hot Applicable.

end welds. and volumetric.

3.4 G-1 pressure retaining Visual and See Remarks Hot Applicable.

bolting. volumetric 3.5 G-2 Pressure retaining Visual See Note (a) and Remarks The bolting less that 2-inches in diameter bolting. will be visually examined in place under tension, or whenever the bolting connection is disassembled.

3.6 Integrally welded Visual and See Note (a) vessel supports. volumetric

TABLE 4.4-6 (Cont'd)

SECTION 3. STEAN GENERATORS Tentative Inspection Item Examination During 10-year No.(b) ~Cate or b Exam(oat)oh Area(b) ~Nethod b Interval Remarks 3.7 I-2 Vessel cladding Visual See Note (a) and Remarks Radiation levels permitting, the examinations will be conducted.

SECTION 4. PIPING PRESSURE BOUNDARY 4.1 Vessel, pump, and valve Visual, sur- See Note (a) and Remarks The pressurizer and pump nozzle-to-safe end Safe ends-to-primary pipe face and volumetric welds are included in this item.

welds and safe ends in branch piping welds.

4.2 G-1 Pressure retaining Visual and See Remarks There is no bolting 2-inches and larger in the bolting. volumetric piping system.

4.3 G-2 Pressure retaining Visual See Note (a) and Remarks All bolting below 2-inches in diameter will be bolting. visually examined, either in-place or whenever the bolted connection is disassembled.

4.4 Circumferential and Visual and See Note (a) An augmented inspection of the following welds longitudinal pipe welds. volumetric shall be performed:

(1) During first inspection interval-1005 of the longitudinal welds from the reactor vessel to approximately the outboard surfaces of the reactor cavity shall be examined at approximately 3 1/3 year intervals in each of the two hot legs of the unit.

(2) During subsequent inspection intervals-provided no defects are found during the first inspection interval exceeding the allowable limits provided in Section XI of the ASHE Code, 100K of these longitudin-al welds in either one of the two hot legs shall be examined during each inspection interval following the first.

TABLE 4.4-6 (Cont'd)

SECTION 4. PiPING PRESSURE BOUNDARY Tentative Inspection Item Examination During 10-year No.(b) ~Cate or (a) Exam(nation Area(t) ~Method a Interval Remarks Branch pipe connection Visual and See Note (a) welds, exceeding 4-in. volumetric nominal pipe size.

4.6 Socket Melds. Visual and See Note (a) surface 4.7 Branch pipe connection Visual and See Note (a) welds, 4-in. nominal surface pipe size and smaller.

4.8 J-2 Circumferential and Visual See Note (a) longitudinal pipe welds and branch pipe connection welds.

4.9 K-1 Integrally welded Visual and See Notes (a) and (d) supports. volumetric 4.10 K-2 Piping supports and Visual See Note (a) hangers.

SECTION 5. PUMP PRESSURE BOUNDARY AND PUMP FLYWHEELS 5.1 Pump casing welds. Visual and See Note (a) and Remarks The only feasible method known to date to volumetric volumetrically examine these pump casing welds is radiography. If radiography is possible, or an alternate examination method is developed, the examination will be performed at the fre-quency indicated. See Note (c).

5.2 L-2 Pump casings. Visual See Note (a) and Remarks Radiation levels permitting, the examination will be performed whenever the pumps are disassembled.

Nozzle-to-safe end welds. Visual and See Note (a) and Remarks This item is considered under Section 4, volumetric Item 4.1.

TABLE 4-4-6 (Cont'd)

SECTION 5. PUMP PRESSURE BOUNDARY AND PUMP FLYWHEELS Tentative Inspection Item Examination During 10-year No.(b) ~Cate or (b) Eaamtaat(oo Area(b) ~lletbod b Interval Remarks 5.4 G-1 Pressure retaining Visual and See Note (a) and Remarks (1) Bolting 2-inches and larger in diameter bolting. volumetric will be examined either in-place under tension, or when the bolting is removed or when the bolting connection is disassembled.

(2) There is no feasible method of examining the flange ligaments of the pumps volumetrically at the present time. See note (c).

5.5 G-2 Pressure retaining Visual See Remarks There is no bolting below 2-inches in diameter.

bolting.

5.6 K-1 Integrally welded Visual and See Note (a) supports. volumetric 5.7 K-2 Supports and hanger s. Visual See Note (a) 5.8 Flywheel Volumetric (1)(2) See Remarks (1) An fn-place ultrasonic volumetric ex-Surface (2) amination of the areas of higher stress concentration at the bore and key way at approximately 3 year intervals, during the refueling or maintenance shutdown coinciding with the inservice inspection schedule.

(2) A surface examination of all exposed sur-faces and complete ultrasonic volumetric examination at approximately 10 year intervals, during the plant shutdown coinciding with the inservice inspection schedule. Removal of the flywheel is not required to perform these examinations.

TABLE 4.4-6 (Cont'd)

SECTION 6. VALVE PRESSURE BOUNDARY Tentative Inspection Item Examination During 10-year No.(b) ~tate or b Examination area(b) ~Method b Interval Remarks 6.1 Valve body welds Visual and Remarks Not Applicable.

volumetric 6.2 N-2 Valve bodies Visual See Note (a) and Remarks Examinations will be conducted provided valves can be disassembed without undue radiation exposure to personnel.

6e3 Valve-to-safe end Visual and See Remarks Not Applicable.

welds. volumetric 6.4 G-1 Pressure retaining Visual and See Remarks There are no valves with bolting 2-inches and bolting. volumetric larger in diameter.

6.5 G-2 Pressure retaining Visual See Note (a) and Remarks All bolting below 2-inches in diameter will be bol'ting. visually examined, either in-place if the bolting connection is not disassembled during the inspection interval, or whenever the bolting connection is disassembled.

6.6 K-1 Integrally welded Visual and See Remarks There are no valves with integrally welded supports volumetric supports.

6.7 K-2 Support and hangers Visual See Note (a)

Note a : The extent an frequency of t e examinations will be at least as much as the guidelines of Table IS-251 of Section XI 1971 Edition through Minter 1972 Addenda.

(b): The item number, category, examination area and examination method are listed in Table IS-261 of Section XI of the ASNE Boiler and Pressure Vessel Code.

(c): Present ultrasonic techniques may not be amenable to examination of pump/valve castings.

(d): Where accessibility to welds prohibits their examination, each weld shall be examined on a case by case basis.

REACTOR COOLANT SYSTEM SAFETY CLASS 2 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.2 The structural integrity of components identified in Section 3.2.2 of the FSAR as Safety Class 2 components shall be maintained at a level consistent-with the acceptance criteria in Specification 4.4.10.2.

APPLICABILITY: NODES 1, 2, 3, and 4.

ACTION:

With the structural integrity of any of the above components not con-forming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component prior to incr easing the Reactor Coolant System above 200'F.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.10.2 The following inspection program shall be performed during shutdown:

a ~ Inservice Ins ections The structural integrity of the Safety Class 2 components shall be demonstrated by verifying their acceptability when inspected per the applicable requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Winter 1972, as outlined by the inspection program shown in Table 4.4-7.

An initial report of any abnormal degradation of the structural integrity of the Safety Class 2 components detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be submitted pursuant to Specification 6.9.1 within 90 days after completion of the surveillance requirements of this specification.

ST. LUG IE - UNIT 1 3/4 4-37

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued The Inservice Inspection Program shall be reviewed every 5 years to assure that the techniques and procedures being uti-lized are current and applicable. The results of these reviews shall be reported in Special Reports to the Commission pursuant to Specification 6.9.2 within 90 days of completion.

b. Ins ections Followin Re airs or Re lacements The structural sntegrity of Safety Class 2 components shall be demonstrated after completion of all 'repairs and/or replacements by veri-fying the repairs and/or replacements meet the requirements of Section XI of the ASME Boiler and Pressure'Vessel Code, 1971 Edition, and Addenda through Winter 1972. When repairs and/or replacements are made which involve new strength welds on com-ponents greater than 2 inch diameter, the new welds shall receive a surface and 100 percent volumetric examination and meet applicable code requirements. When repairs and/or replace-ments are made which involve new strength welds on components 2 inch diameter or smaller, the new welds shall receive a surface examination and meet applicable code requirements.

c. S stem Pressure Tests The structural integrity of Safety Class 2 components shall be demonstrated by performing system pressure tests per the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Winter 1972.

ST. LUCIE - UNIT 1 3/4 4-38

TABLE 4.4-7 INSERVICE INSPECTION PROGRAM - SAFETY CLASS 2 COMPONENTS SECTION Cl. PRESSURE VESSELS Tentative Inspection Item Examination During 10-year tto. ~Cate oe Examination Area Method Interval Remarks Cl . 1 C-A Circumferential Volumetric The components in this category are included (1) The areas of examination include shell and head in the overall total of required examinations shell and head circumferential welds for ISC-261(a) and ISC-261(b) components. A welds which are gross structural cumulative 25K of the overall total of re- discontinuities (see Subparagraph quired examinations will be completed during NE-3213.2 of Section III-the interval. See Note (a). Nuclear Power Plant Components Code) in vessels exceeding 4-inch nominal pipe size.

(2) See Note (f)

(3) See Note (b).

(4) The steam generator staywell dome welds will not be examined due to their configuration, limited access and radiation exposure to personnel.

Cl.2 C-B Nozzle-to-vessel Volumetric The components in this category are included (1) See Note (b).

welds in the overall total of required examinations for ISC-261(a) and ISC-261(b) components. A (2) See Note (f).

cumulative 25% of the overall total of re-quired examinations will be completed during the interval. See Note (a).

TABLE 4.4-7 Cont'd SECTION Cl. PRESSURE VESSELS Tentative Inspection Item Examination During 10-year No. ~Cate or Examination Area Method Interval Remarks C1.3 C-C Integrally-welded Surface The components in this category are included (1) See Note (b).

supports in the overall total of required examinations for ISC-261(a) and ISC-261(b) components. A (2) See Note (f).

cumulative 25K oF the overall total of re-quired examinations will be completed during the interval. See Note (a).

C1.4 C-D Pressure-retaining Visual and The components in this category are included (1) Bolting exceeding 1-inch in diameter bolting either surface in the overall total of required examinations will be examined in-place in the bolted-or volumetric for ISC-261(a) and ISC-261(b) components. A up condition or whenever the bolted cumulative 25% of the overall total of required connection is disassembled. Flange examinations will be completed during the ligaments between threaded stud holes interval. See Note (a). will be included whenever the connection is disassemble, and feasible methods of examination have been developed.

(2) A required examination consists of the following:

(a) Visual examinations will include 100K of the bolts, studs, and nuts in the bolted joint.

(b) Surface examinations will be performed on 10$ of the bolting in each joint, but not less than 2 bolts or studs per joint whenever the connection is disassembled.

(3) See Note (b).

(4) See Note (f).

TABLE 4.4-7 Cont'd SECTION Cl. PRESSURE VESSELS Tentative Inspection Item Examination During 10-year eo. ~cate or Examination Area Nethod Interval 'emarks C1.5 ISC-261(c) Visual Cumulative 100% of the required (1) Examinations will be conducted in accordance with components examinations (see Note (c)). the procedures of IS-211 for evidence of component leakage or structural distress with the system under pressure as specified in ISC-520(b), when the system is undergoing either a periodic system performance test or a system pressure test.

(2) For insulated components, the examinations will be conducted without the removal of the insulation.

(3) Components in letdown portion of CVCS system (from pressure control valves to volume control tank, which normally operates at less than 275 psig 200 F) will be visually examined in accordance with ISC-261(c).

C1.6 ISC-261(d) Visual Cumulative 100% of the required (1) See Note (g).

Components examinations (see Note (c)).

SECTION C2. PIPING C2.1 C-F Circumfer ential Volumetric The components in this category (1) The areas of examination include circumferential C2.1.1 butt welds are included in the overall total butt welds at structural discontinuities in of required examinations for piping exceeding 4-inch nominal pipe size and ISC-261(a) components. circumferential butt welds in piping exceeding A cumulative 255 of the overall 4-inch nominal pipe size within 3 pipe-diameters total of required examinations will of the center line of rigid pipe anchors, or be completed during the interval. anchor s at the penetr ation of the primary con-See Note (a). tainment, or at rigidly anchored components.

(2) See Note (b).

(3) See Note (h).

TABLE 4.4-7 Cont'd SECTION C2. PIPING Tentative Inspection Item Examination During 10-'year aao. ~tate oe Examination Area Method Interval Remarks C2.1 C-G Circumferential Volumetric One-half of the components in this (1) The areas of examination include circumferential C2.1.2 butt welds category will be selected and included butt welds at structural discontinuities in in the overall total of required ex- piping exceeding 4-inch nominal pipe size and aminations for ISC-261(b) components. circumferential butt welds in piping exceeding A cumulative 25% of the overall total 4-inch nominal pipe size within 3 pipe-diameters of required examinations will be of the centerline of rigid pipe anchors, or completed during the interval. See anchors at the penetration of the primary con-Note (a). tainment, or at rigidly anchored components.

(2) A representative sampling among the total number of welds will be selected for examination.

(3) See Note (b).

(4) See Note (i).

C2.2 C-F Longitudinal weld Volumetric The components in this category are in- (1) The areas of examination include the longitudinal C2.2.1 joints in fittings cluded in the overall total of required weld joints in pipe fittings exceeding 4-inch examinations for ISC-261(a) components. nominal pipe size.

A cumulative 25% of the overall required examinations will be completed during (2) See Note (b).

the interval. See Note (a).

(3) See Note (h).

TABLE 4.4-7 Cont'd SECTION C2. PIPING Tentative Inspection Item Examination During 10-year No. ~Cate or Examination Area Hethod Interval Remarks C2.2 C-G Longitudinal weld Volumetric See Remarks. Not applicable for Note (i) systems.

C2.2.2 Joints in fittings C2.3 C-F Branch pipe-to-pipe Volumetric See Remarks. Not applicable for Note (h) systems.

C2.3.1 weld joints C2.3.2 C-G Branch pipe-to-pipe Volumetric One-half of the components in this (1) The areas of examination include branch con-weld joints category will be selected and included nection weld joints exceeding 4-inch nominal in the overall total of required ex- pipe size.

aminations for ISC-261(b) components.

A cumulative 25K of the overall total (2) A representative sampling among the total number of required examinations will be of welds will be selected for examination.

completed during the interval.

See Note (a). (3) See Note (b).

(4) See Note (i).

TABLE 4.4-7 Cont'd SECTION C2. PIPING Tentative Inspection Item Examination During 10-year No. ~tate or Examination Area Method Interval Remarks C2.4 C-C Integrally-welded Suface The components in this category are in- (1) See Note (b).

support-to-pipe cluded in the overall total of required welds examinations for ISC-261(a) and ISC-261(b) (2) See Notes (h) and (i).

components. A cumulative 25K of the over-all total of required examinations will be completed during the interval. See Note (a).

C2.5 C-D Pressure-retaining Visual and The components in this category are in- (1) Bolting exceeding 1-inch in diameter will be bolting either sur- eluded in the overall total of required examined in-place in the bolted-up condition face or examinations for ISC-261(a) and ISC-261(b) or whenever the bolted connection is dis-volumetr ic componetns. A cumulative 25% of the over- assembled. Flange ligaments between threaded all total of required examinations will be 'stud holes will be included whenever the con-completed during the interval. See Note (a). nection is disassembled, and feasible examina-tion methods have been developed.

(2) A required examination consists of the following:

(a) Visual examinations will include 100%

of the bolts, studs, and nuts in the bolted joint.

(b) Surface examinations will be performed on of the bolting in each joint, but not less than 2 bolts or studs per joint whenever the connection is disassembled.

(3) See Note (b).

(4) See Notes (h) and (i) ~

TABLE 4.4-7 Cont'd SECTION C2. PIPING Tentative Inspection Item Examination During 10-year tto. ~tate or Examination Area Method Interval Remar ks C2.6 C-E Supports and Visual The components in this category are included (1) See Note (b).

hangers in the overall total of required examinations for ISC-261(a) and ISC-261(b) components. A cumulative 25$ of the overall total of required examinations will be completed during the inter-val. See Notes (a) and (e).

C2.7 ISC-261(c) Visual Cumualtive 100K of the required examinations. (1) Examinations will be conducted in accord-System See Note (c). ance with the procedure of IS-211 for evidence of component leakage or structural distress with the system under pressure as specified in ISC-520(b), when the system is undergoing either a periodic system per-formance test or a system pressure test.

(2) For insulated components, the examination will be conducted without the removal of the insulation.

I TABLE 4.4-7 Cont'd n

SECTION C2. PIPING I

Tentative Inspection Item Examination During 10-year Io. ~Cate or Ixantnatton Area Method Interval Remarks (3) Components in letdown portion of CVCS system (from pressure control valves to volume control tank, which normally operates at less than 275 will be visually psig'00 F) examined in accordance with ISC-261(c}.

C2.8 -- ISC-261(d) systems Visual Cumulative 10(C of the required (1} See Hot (g).

examinations (see Note (c)).

SECTION C3. PUMPS C3.'1 C-F Pump casing welds Volumetric See Remarks. Hot applicable for Note (j) pumps.

C.3.1.1 C.3.1.2 C-G Pump casing welds Volumetric See Remarks. Not applicable for Hote (j) pumps.

TABLE 4.4-7 Cont'd SECTION C3. PUMPS Tentative Inspection Item Examination During 10-year eo. ~cate or Examination Area Method Interval Remarks C3.2 C-H Pump Casing Visual The components in this category are in- (1) See Note (b).

eluded in the overall total of required examinations for ISC-261(a) and ISC-261(b) (2) See Note (j).

components. A cumulative 25% of the over-all total of required examinations will be completed during the interval. See Note (a).

C3.3 C-D Pressure-,retaining Visual and The components in this category are in- (1) Bolting exceeding 1-inch in diameter will be bolting either surface eluded in the overall total of required examined in-place in the bolted-up condition or volumetric examinations for ISC-261(a) and ISC-261(b) or whenever the bolted connection is dis-components. A cumulative 25K of the over- assembled. Flange ligaments between threaded all total of required examinations will be stud holes will be included whenever the con-completed during the -interval. See Note nection is disassembled, and feasible examina-(a). tion methods have been developed.

(2) A required examination consists of the following:

(a) Visual examinations will include 100% of the bolts, studs, and nuts in the bolted joint.

(b) Surface examinations will be performed on 10$ of the bolting in each joint, but not less than 2 bolts or studs per joint whenever the connection is disassembled.

(3) See Note (b).

(4) See Note (j).

I TABLE 4.4-7 Cont'd tol SECTION C3. PUMPS Tentative Inspection Item Examination During 10-year flo. ~Cate or Examination Area Method Interval Remarks C3.4 C-E Supports and Visual The components in this category are in- (1) See Note (b).

hangers eluded in the overall total of required examinations for ISC-261(a) and ISC-261(b) (2) See Note (j).

components. A cumulative 25% of the over-all total of required examinations will be completed during the interval. See Notes (a) and (e).

C3.5 Integral ly welded Surface See Remarks Not applicable for Note (j) pumps.

supports I C3.6 ISC-261(c) Visual Cumulative 50K of the required examinations (1) Examinations will be conducted in accordance components See Note (c). with the procedures of IS-211 for evidence of component leakage or structura'I distress with the system under pressure as specified in ISC-520(b), when the system is undergoing either a periodic system performance test or a system pressure test.

(2) For insulated components, the examinations will be conducted without the removal of the insulation.

E TABLE 4.4-7 Cont'd n

Etl SECTION C3. PUMPS Tentative Inspection Item Examination Ouring 10-year Eo. ~Cate or Examination Area Method Interval Remarks (3} Components in letdown portion of CVCS system (from pressure control valves to volume control tank, which normally operates at less than 275 psig 200 F) will be visually examined in accordance with ISC-261(c).

C3.7 ISC-261(d) Visual Cumulative 50$ of the required examinations (1) See Note (g).

Components (see Note (c)).

SECTION C4. VALVES E

C4.1 C-F Valve body welds Volumetric See Remarks Not applicable for Notes (h) and (i) systems.

C4.1.1 C4.1.2 C-G Valve body welds Volumetric See Remarks Not applicable for Notes (h) and (i) systems.

I TABLE 4.4-7 Cont'd C

Dl SECTION C4. VALVES Tentative Inspection Item Examination During 10-year Oo. ~Cate or Examination Area Hethod Interval Remarks C4.2 C-H Valve Bodies Visual The components in this category are included (1) See Note (b).

in the overall total of required examinations for ISC-261(a) and ISC-26'l(b) components. A (2) See Notes (h) and (i).

cerulative 25K of the overall total of required examinations will be completed during the interval. See Note (a).

C4.3 C-D Pressure-retain- Visual and The components in this category are included (1) Bolting exceeding inch in diameter will ing bolting either sur- in the overall total oF required examinations be examined in-place in the bolted-up con-face or for ISC-261(a) and ISC-261(b) components. A dition or whenever the bolted connection volumetric cumulative 25K of the overall total of re- is disassembled. Flange ligaments between C) quired examinations will be completed during threaded stud holes will be included when-the interval. See Note (a). ever the connection is disassembled. See Note (d).

(2) A required examination consists of the following:

(a) Visual examinations will include 100$ of the bolts, studs, and nuts in the bolted joint.

(b) Surface examinations will be performed on 10Ã of the bolting in each joint, but not less than 2 bolts or studs per joint whenever the connection is disassembled.

(3) See Note (b).

(4) See Notes (h) and (i).

TABLE 4.4-7 Cont'd SECTION C4. VALVES Tentative Inspection 5 Item Examination During 10-year No. ~Cate oe Examination Area Method Interval Remarks C4.4 C-E Supports and Visual The components in this category are included (1) See Note (b).

hangers in the overall total of required examinations for ISC-261(a) and -ISC-261(b) components. A cumulative 25% of the overall total, of required examinations will be completed during the interval. See Notes (a) and (e).

C4.5 Integrally-welded Surface See Remarks Not applicable for Notes (h) and (i) systems.

supports C4.6 ISC-261(c) Visual Cumulative 50% of the required examinations. (1) Examinations will be conducted in accordance components See Note (c) ~ with the procedures of IS-211 for evidence of component leakage or structural distress with the system under pressure as specified in ISC-520(b), when the system is undergoing either a periodic system performance test or a system pressure test.

(2) For insulated components, the examinations will be conducted without the removal of the insulation.

(3) Components in letdown portion of CVCS system (from pressure control valves to volume con-trol tank, which normally operates at less than 275 psig 200 F) will be visually ex-amined in accordance with ISC-261(c).

TABLE 4.4-7 Cont'd SECTION C4. VALVES Tentative Inspection Item Examination During 10-year nn. ~Cate nr Exantnattnn Area Method Interval Remarks C4.7 --- ISC-261(d) Visual Cumulative 50K of the required examinations (1) See Note (g).

Components (see note (c)).

Note a : t east part o t e overa tota of required examinations will be performed by the expiration of 1/3, 2/3, and the end of .the 10-year interval.

The total number of required examinations is dependent upon whether or not a system consists of multiple streams which perform the same (or redundant) functions as stated in ISC-242(a); the distribution of individual components to be examined will be determined in accordance with ISC-242(b), (c), and (d).

(c): At least 25K of the required examinations will be completed by the expiration of 1/3 of the 10-year inspection interval, with not more than 66-2/3% of the required examinations completed by the expiration of 2/3 of the 10-year inspection interval.

d) Present ultrasonic techniques and procedures may not be amenable to examination of pump castings.

e) Where accessibility to welds prohibits their examination, each weld will be evaluated on a case by case basis.

f) The regenerative, letdown, and shutdown heat exchangers, and the steam generators (secondary side) are to be examined.

Components and systems in which the contained fluid will be verified by periodic sampling and tests.

Portions of the shutdown cooling systems are to be examined.

i) Portions of the main steam and feedwater systems are to be examined.

Low pressure safety irdection pumps are to be examined.

REACTOR COOLANT SYSTEM SAFETY CLASS 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.3 The structural integrity of components identified in Section 3.2.2 of the FSAR as Safety Class 3 components shall be maintained at a level consistent with the acceptance criteria in Specification 4.4.10.3.

APPLICABILITY: ALL MODES.

ACTION:

With the structural integrity of any of the above components not con-forming to the above requirements, restore the structural integrity of the component to within its limit or isolate the affected components from service. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.10.3 The following inspection program shall be performed:

a ~ Inservice Ins ections The structural integrity of the Safety Class 3 components shall be demonstrated at least once per 40 months during periods of normal reactor operation or during system performance testing by verifying via visual inspections, as outlined by the inspection program shown in Table 4.'4-8, that there is no evidence of unanticipated component leakage, structural distress, or corrosion.

An initial report of any abnormal degradation of the structural integrity of the Safety Class 3 components detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be submitted pursuant to Specification 6.9.1 within 90 days after completion of the surveillance requirements of this specification.

ST. LUCIE - UNIT 1 3/4 4-53

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued

b. S stem Pressure Tests The structural integrity of the Safety Class 3 components shall be demonstrated at least once per 10 years by performing system pressure tests at the following test pressures:

For closed systems, at least 110 percent of the design pressure,

2. For open storage tanks, at least the nominal hydrostatic pressure developed with the tanks filled to design capa-city, and

3. Open-ended portions of systems may be exempted from pres-sure testing.

c ~ Pi e Han er Ins ections The structural integrity of the Safety Class 3 components shall be demonstrated at least once per 40 months by verifying via visual inspections that the supports and hangers for piping and components over 4 inches in diameter show no evidence of inadequate support, unintended restraint, or structural distress.

ST. LUCIE - UNIT 1 3/4 4-54

TABLE 4.4-8 INSERVICE INSPECTION PROGRAM SAFETY CLASS 3 COMPONENTS Class 3 piping greater than 4 inches and components with pipe connections greater than 4 inches will be pressure tested and visually examined near or at the end of the 10-year inspection interval. In addition, components will be visually examined during periods of normal operations or during system performance testing once during each 1/3 of the year 10-year inspection interval.

Open-ended portions of systems will not be pressur e tested.

REACTOR COOLANT SYSTEM CORE BARREL MOVEMENT LIMITING CONDITION FOR OPERATION 3.4.11 Core barrel movement shall be limited to less than the Amplitude Probability Distribution (APD) and Spectral Analysis (SA) Alert Levels for the applicable THERMAL POWER level.

APPLICABILITY: MODE 1.

ACTION:

a ~ With the APD and/or SA exceeding their applicable Alert Levels, POWER OPERATION may proceed provided the following actions are taken:

l. APD shall be measured and processed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

2. SA shall be measured within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months therea'fter and SA shall be processed at least once per 7 days, and

3. A Special Report, identifying the cause(s) for exceeding the applicable Alert Level, shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 30 days of detection.

b. With the APD and/or SA exceeding their applicable Action Levels, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months reduce THERMAL POWER BY > 254 of RATED THERMAL POWER and demonstrate, through mon7toring of the excore neutron detectors, that APD and SA have been reduced to below their applicable Alert Level limits or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C. With the measured levels of APD and/or SA differing from their baseline levels by more than 105, a Special Report describing the measured levels shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days of data processing.

ST. LUCIE - UNIT 1 3/4 4-56

REACTOR COOLANT SYSTEM CORE BARREL MOVEMENT Continued)

SURVEILLANCE RE UIREMENTS 4.4.11.1 Baseline Monitorin Core barrel movement Alert Levels and Action Levels, as determ>ned by APD and SA monitoring of the excore neutron detectors, shall be determined at nominal THERMAL POWER levels of 20K, 50%, 805 and 100% of RATED THERMAL POWER during the reactor startup test program; these Alert Levels and Action Levels shall be reported in a Special Report pursuant to Specification 6.9.2 within 31 days after initially reaching 100Ã of RATED THERMAL POWER.

4.4.11.2 Routine Monitovin Core barrel movement shall be determined to be less than t e A D an SA Alert Levels by using the excore neutron detectors to measure APD and SA at the following frequencies:

a. APD data shall be measured and processed at least once per 7 days.

b. SA data shall be measured and processed at least once at nominal THERMAL POWER levels of 20K, 50Ã, 80Ã and 100%%d of RATED THERMAL POWER after each refueling and at least once per 4 months thereafter.

4.4.11.3 ~Re orts The results of all periodic APD and SA monitoring shall be included 1n the Annual Operating Report for the period in which the monitoring was performed.

ST. LUCIE - UNIT 1 3/4 4-57

0 3/4.5 EMERGENCY CORE COOLING SYSTEMS ECCS SAFETY INJECTION TANKS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with:

a ~ The isolation valve open,

b. Between 1090 and 1170 cubic feet of borated water, c ~ A minimum boron concentration of 1720 PPM, and

d. A nitrogen cover-pressure of between 200 and 250 psig.

APPLICABILITY: MODES 1, 2 and 3.*

ACTION:

a ~ With one safety injection tank inoperable, except as a result of a closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in HOT SHUTDOWN within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one safety injection tank inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE RE UIREMENTS .

4.5.1 Each safety injection tank shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

l. Verifying the water level and nitrogen cover-pressure in the tanks, and

2. Verifying that each safety injection tank isolation valve is open.

With pressurizer pressure > 1750 psia.

ST. LUG IE - UNIT 1 3/4 5-1

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued

b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of ) 1/ of tank volume by verifying the boron concentration of the safety injection tank solution.

c. At least once per 31 days when the RCS pressure is above 1750 psia, by verifying that power to the isolation valve operator is removed by maintaining the breaker open under administrative control.

d. At least once per 18 months by verifying that each safety injec-tion tank isolation valve opens automatically under each of the following conditions make it:

1. When the RCS pressure exceeds 350 psia, and

2. Upon receipt of a safety injection test signal.

ST. LUCIE - UNIT 1 3/4 5-2

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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position

1. V-3659 1. Mini-flow l. Open isolation

2. V-3660 2. Mini-flow 2. Open isolation

b. At least once per 31 days on a STAGGERED TEST BASIS by:

l. Verifying that each high-pressure safety, injection pump:

a) Starts (unless already. operating) from the control room.

b) Develops a discharge pressure of > 1138 psig on recirculation flow.

c) Operates for at least 15 minutes.

2. Verifying that each low-pressure safety injection pump:

a) Starts (unless already operating) from the control room.

b) Develops a discharge pressure of > 175 psig on recirculation flow.

c) Operates for at least 15 minutes.

3. Verifying that upon a recirculation actuation signal, the containment sump isolation valves open.

4. Cycling each testabl'e, power operated valve in the flow path through at least one complete cycle of the full travel.

ST. LUCIE - UNIT 1 3/4 5-4

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued

5. Verifying that each valve (manual, power operated or automa-tic) in the flow path that is not locked, sealed, or other-wise secured in position, is in its correct position.

6. Verifying that each ECCS subsystem is'ligned to receive electrical power from separate OPERABLE emergency busses.

c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2. Of the areas affected within containment at the completion of containment entry when CONTAINMENT INTEGRITY is established.

d. At least once per 18 months by:

l. Verifying automatic isolation of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure is above 300 psig.

2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distr ess or corrosion.

3. Verifying that a minimum total of 86 cubic feet of solid granular trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

4. Verifying that when a representative sample of 0.5 + 0.1 lbs of TSP from a TSP storage basket is submerged, without agitation, in 50 + 1.0 gallons of 200 + 10'F borated water from the RWT, the pH of the mixed solution is raised to > 6 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ST. LUCIE - UNIT 1 3/4 5-5

EMERGENCY. CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued

e. At least once per 18 months, during shutdown, by;

1. Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.

2. Verifying that each automatic valve in the flow path actu-ates to its correct position on a Safety Injection Actuation Signal.

3. Verifying that each of the following pumps start automatical-ly upon receipt of a Safety Injection Actuation Signal;

a. High-Pressure Safety Injection pump.

b. Low-Pressure Safety Injection pump.

0 ST. LUCIE - UNIT 1 3/4 5-6

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T < 300'F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE high-pressure safety injection pump, and

b. An OPERABLE flow path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a recirculation actuation signal.

APPLICABILITY: MODES 3* and 4.

ACTION:

a ~ With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

SURVEILLANCE RE UIREMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

With pressuri2er pressure < 1750 psia.

ST. LUCIE UNIT 1 3/4 5-7

EMERGENCY CORE COOLING SYSTEMS REFUELING WATER TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank shall be OPERABLE with:

a. A minimum contained volume 371,800 gallons of borated water,

b. A minimum boron concentration of 1720 ppm,

c. A maximum water temperature of 100'F,

d. A minimum water temperature of 55'F when in MODES 1 and 2, and

e. A minimum water temperature of 40"F when in MODES 3 and 4.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION the refueling water tank inoperable, restore the tank to OPERABLE

'ith status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within,6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .,

SURVEILLANCE RE UIREMENTS 4.5 ' The RWT shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the water level in the tank, and

2. Verifying the boron concentration of the .water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWT temperature.

ST. LUCIE - UNIT 1 3/4 5-8

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT VESSEL CONTAINMENT VESSEL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 CONTAINMENT VESSEL INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without CONTAINMENT VESSEL INTEGRITY, restore CONTAINMENT VESSEL INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.1.1 CONTAINMENT VESSEL INTEGRITY shall be demonstrated:

At least once per 31 days by verifying that:

l. All containment vessel penetrations not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-2 of Specification 3.6.3.1, and

2. All containment vessel equipment hatches are closed and sealed.

b. By verifying that each containment vessel air lock is OPERABLE per Specification 3.6.1.3.

ST. LUCIE - UNIT 1 3/4 6-1

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of:

l. < La, 0.50 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P , (39.6 psig), or

2. < Lt, 0.32 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduced pressure of Pt, (19.8 psig).

b. A combined leakage rate of < 0.60 La for all penetrations and valves subject to Type B an3 C tests as identified in Table 3.6-1 when pressurized to Pa.

c. A combined leakage rate of < 0.12 La for all penetrations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to Pa.

APPLICABILITY: MODES 1, 2, 3'nd 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 La or 0.75 Lt, as applicable, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 La, or (c) with the combined bypass leakage rate exceeding 0.12 La, restore the leakage rate(s) to within the limit(s) prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE RE UIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and pro-visons of ANSI N45.4-1972:

a. Three Type A tests (Overall Integrated Containment Leakage Rate) .

shall be conducted at 40 + 10 month intervals during shutdown at either Pa (39.6 psig) or at Pt (19.8 psig) during each 10-year ST. LUCI E - UNIT 1 3/4 6-2

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued service period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection.

b. If any periodic Type A test fails to meet either .75 L or .75 L ,

the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet either .75 L or .75 L , a Type A test shall be performed at least every 18 mont(is until two con-secutive Type A tests meet either .75 L or .75 Lt at which time the above test schedule may be resumed.

c. The accuracy of each Type A test shall be verified by a supplemental test which:

Confirms the accuracy of the Type A test by verifying that the difference between supplemental and Type A test data is within 0.25 L or 0.25 Lt, a

2. Has a duration sufficient to establish accurately the change in leakage between the Type A test and the supplemental test.

3. Requires the quantity of gas injected into the containment

,or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage rate at P (39.6 psig) or Pt (19.8 psig).

a

d. Type B and C tests shall be conducted with gas at P (39.6 psig) at intervals no greater than 24 months except for tests in-volving air locks.

e. The combined bypass leakage rate shall be determined to be

( 0.12 L by applicable Type B and C tests at least once per 24 months except for penetrations which are not individually testable; penetrations not individually testable shall be "

determined to have no detectable leakage when tested with soap bubbles while the containment is pressurized to P (39.6 psig) during each Type A test.

f. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.

ST. LUG I E - UNIT 1 3/4 6-3

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

g. All Type A test leakage rates shall be calculated using observed data converted to absolute values. Error analyses shall be performed to select a balanced integrated leakage measurement system.

ST. LUCIE - UNIT 1 3/4 6-4

TABLE 3.6-1 CONTAINMENT LEAKAGE PATHS Location Test Penetration ~S stem ~Y1 T N b Service ~Te*

Makeup Water Gate (I-MV-15-1) Outside Primary Makeup Mater Bypass Check ( I-V-15-1347) Inside Station Air Globe (I-V-18-947) Outside Station Air Supply Bypass Globe (I-V-18-947 Outside Instrument Air Gate ( I-MV-18-1) Outside Instrument Air Supply Bypass Check ( I-V-18-957) Inside 10 Contai'nment Butterfly ( I-FCV-25-4) Inside'utside Containment Purge Type C Purge Butterfly (I-FCV-25-5) Exhaust Containment Butterfly (I-FCV-25-3) Inside Containment Purge Type C Purge Butterfly (I-FCV-25-2) Outside Supply 14 Waste Globe (V-6741) Outside Nitrogen supply to Bypass Management Check Y-6779 Outside SI Tanks 23 Component Butterfly (I-HCV-14-7) Outside RC Pump CW supply Bypass Butterfly (I-HCV-14-1) Outside Cooling'omponent 24 Butterfly ( I-HCV-14-6) Outside RC Pump CW Return Bypass Cooling Butterfly (I-HCV-14-2) Outside 25 Fuel Transfer Blind Flange Inside Fuel Transfer Bypass Tube 26 CVCS Globe (V-2515) Inside Letdown Line Bypass Globe (V-2516) Inside 28 Sampling Globe (V-5200) Outside Reactor Coolant Bypass Globe (V-5203) Outside Sample

TABLE 3.6-1 Continued Location Test Penetration ~S stem Valve Ta Number to Containment Service ~Te*

29 Sampling Globe (V-5202) Outside Pressurizer Steam Bypass Globe (V-5205) Outside Space Sample 29 Sampling Globe (V-5201) Outside Pressurizer Surge Bypass Globe (V-5204) Outside Line Sample 31 Waste Gate (V-6554) Outside Containment Vent Bypass Management Gate V-6555) Outside Header 41 Safety Injection Gate (V-3463) Outside Safety Injection Tank Bypass Tank Test Lines Gate (I-V-03-1307) Outside Fill and Sampling 42 Waste Gate (I-LCV-07-llA) Outside Reactor Cavity Sump Bypass Management Gate (I-LCV-07-llB Outside Pump Discharge 43 Waste Gate (V-6301) Outside Reactor Drain Tank Bypass Management Gate (V-6302) Outside Pump Suction 44 CVCS Gate (V-2505) Outside RC Pump Controlled Bypass Gate (I-SE-01-1) Inside Bleedoff 46 Fuel Pool Gate (I-V-07-206) Outside Refueling Cavity Bypass Cleanup Gate (I-V-07-189 Inside Purification Flow Inlet 47 Fuel Pool Gate (I-V-07-170) Outside Refueling Cavity Bypass Cleanup Gate (I-V-07-188) Inside Purification Flow Outlet 48 . Sampling Globe ( I-FSE-27-01,02, Inside H2 Sampling Type C 03,04)

Globe (I-FSE-27-08) Outside

TABLE 3.6-1 Continued Location Test I

Penetration ~S stem Valve Ta Number to Containment Service ~Te*

n Globe I-FSE-27-11) Outside Sampling Type C m 48 Sampling ( H2 Check (I-FSE-27-1341) Inside 51 Sampling Globe (I-FSE-27-5,6,7) Inside H2 Sampling Type C Globe (I-FSE-27-9) Outside 51 Sampling Globe (I-FSE-27-10) Outside H2 Sampling Type C Check (I-FES-27-1342) Inside 52a Sampling Gate (I-FCV-26-1) Inside Radiation Bypass Gate (I-FCV-26-2) Outside Monitoring 52b Sampling Gate (I-FCV-26-3) Inside Radi ation Bypass Gate (I-FCV-26-4) Outside Monitoring 52c Sampling Gate (I-FCV-26-5) Inside Radi ation Bypass Gate (I-FCV-26-6) Outside Monitoring Return 52d I LRT Gate (I-V00140(1325)) Inside ILRT Test Tap Bypass Gate (I-V00143(1325)) Outside 52e ILRT Gate (I-V00139(1322)) Inside ILRT Test Tap Bypass Gate (I-V00144(1322) Outside 54 ILRT Blind Flange Inside ILRT Pressure Bypass Gate (I-V00101(612)) Outside Connection 56 Containment Gate (V-25-11) Outside Hydrogen Purge Outside Bypass H2 Purge Gate (V-25-12) Outside Air Makeup 57 Containment Gate (V-25-13) Outside Hydrogen Purge Exhaust Bypass H2 Purge Gate (V-25-14) Outside

TABLE 3.6-1 Continued Location Test Penetration ~S stem " Valve Ta Number to Containment Service ~Te*

I Containment Gate (V-25-15) Outside Hydrogen Purge Exhaust Bypass Pl Purge Gate (Y-25-16) Outside H2 67 Vacuum Check (I-V-25-20) Inside Containment Type C Rel i ef Butterfly (I-FCV-25-7) Outside Vacuum Relief Vacuum Check (I-V-25-21) Inside Containment Type C Relief Butterfly (I-FCY-25-8) Outside Vacuum Relief Personnel N.A. None N.A. Ingress 8 Egress Type B**

Lock to Containment 4'scape N.A. None N.A. Emergency Ingress 8 Type B**

Lock Egress'to Containment Ch I

Maintenance N.A. None N.A. Vessel Maintenance Type B Hatch (Gasket Interspace)

Electrical N.A. All primary canisters N.A. Electrical Type B Penetrations except welded spares connections in PCY Main Steam Steel Tap 1- Outside Expansion Bellows Type B Containment Tap 2 Outside Nozzles Main Steam Steel Tap 1 Outside Expansion Bellows Type B Containment Tap 2 Outside Nozzles Feedwater Steel Tap 1 Outside Expansion Bellows Type B Containment Nozzles Tap 2 Outside

TABLE 3.6-1 Continued Location Test Penetration ~S stem Valve Ta Number to Containment Service Tyye*

Feedwater Steel Tap 1 Outside Expansion Bellows Type B Containment Tap 2 Outside Nozzles 25 Fuel Tube Tap 1 Inside Expansion Bellows Type 8 Steel Containment Tap 2 Outside Nozzles

Type C and bypass tests are conducted in the same manner, the only difference is in the acceptance criteria that is applicable.

- In accordance with Specification 4.6.1.3.b.

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of < 0.05 La at Pa, (39.6 psig).

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a 0 With an air lock inoperable, except as a result of an inoperable door gasket, restore the air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With an air lock inoperable due to an inoperable door gasket:

1. Maintain the remaining door of the affected air lock

'closed and sealed, and

2. Restore the air lock to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a ~ After each opening by verifying the seal leakage is < 0.01 L as determined. by precision flow measurement when measured for at least 30 seconds with:

1. The volume between the personnel air lock seals at a constant-pressure of 39.6 psig, and

2. The volume between the emergency air lock seals at a constant pressure of 10 psig.

ST. LUCIE - UNIT 1 3/4 6-10

I ~

~

~ I I ~ ~ ~ ~ ~ ~ ~ I

~ ~ ~ I I

~ ~ ~ ~

~ I ~ ~ I ~ I ~ I ~ ~

~ ~ I ~

CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between -0.7 and 2.4 PSIG.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST. LUCIE - UNIT 1 3/4 6-12

CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120'F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment average air temperature >. 120'F, reduce the average air temperatur e to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at three of the following locations and shall be determined at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

=Location

a. Containment fan cooler No. 1A air intake, elevation 45 feet.

b. Containment fan cooler No. 1B air intake, elevation 45 feet.

c. Containment fan cooler No. 1C air intake, elevation 62 feet.

d. Containment fan cooler No. 1D air intake, elevation 45 feet.

ST. LUCI E - UNIT 1 3/4 6-13

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Speci-fication 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE RE UIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the accessible interior and exterior surfaces of the vessel and veri'fying no apparent changes in appearance of the surfaces or other abnormal degra-dation. An initial report of any abnormal degradation of the containment vessel detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be*submitted to the Commission pursuant to Specification 6.9.1 within 90 days after completion of the surveillance requirements of this specification.

ST. LUCIE - UNIT,l 3/4 6-14

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWT on a Containment Spray Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal.

Each spray system flow path from the containment sump shall be via an OPERABLE shutdown cooling heat exchanger.

APPLICABILITY:, MODES 1, 2 and 3*.

ACTION:

a ~ With one containment spray system inoperable and all four containment fan coolers OPERABLE, restore the inoperable spray system to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one containment spray system inoperable and one containment fan cooler inoperable, restore either the inoperable spray system or the inoperable fan cooler to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a ~ At least once per 31 days on a STAGGERED TEST BASIS by:

1. Starting each spray pump from the control room,

2. Verifying, that on recirculation flow, each spray pump develops a discharge pressure of ~ 200 psig,

3. Verifying that each spray pump operates for at least 15 minutes, Applicable when pressurizer pressure is ~ 1750 psia.

ST. LUCIE - UNIT 1 3/4 6-15

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

4. Cycling each testable, power operated or automatic valve in the flow path through at least one complete cycle of full travel,

5. Verifying that 'upon a recirculation actuation signal, the containment sump isolation valves open and that a recirculation mode flow path via an OPERABLE shutdown cooling heat exchanger is established, and

6. Verifying that each valve (manual, power operated or automatic) in the flow path is positioned to take suction from the RWT on a Containment Pressure--High-High signal.

b. At least once per 18 months, during shutdown, by:

1. Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one, complete cycle of full travel.

2. Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure High-High signal.

~

3. Verifying that each spray pump starts automatically on a Containment Pressure--High-High signal.

c~ At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spr'ay nozzle is unobstructed.

ST. LUCIE - UNIT 1 3/4 6-16 0

CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 Four containment fan coolers shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a ~ With one containment fan cooler inoperable and both containment spray systems OPERABLE, restore the inoperable fan cooler to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one containment fan cooler inoperable and one containment spray system inoperable, restore either the inoperable fan cooler or the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.2.3 Each containment fan cooler shall be demonstrated OPERABLE at least once per 31 days on a STAGGERED TEST BASIS by:

a. Starting each unit from the control room,

b. Verifying that each unit operates for at least 15 minutes, and

c. Verifying a cooling water flow rate of > 1200 gpm to each cooling unit.

ST. LUCIE - UNIT 1 3/4 6-17

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION, FOR OPERATION 3.6.3.1 The containment isolation valves specified in Table 3.6-2 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more of the isolation valve(s) specified in Table 3.6-2 inoperable, either:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or c ~ Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange; or

d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.3.1.1 The isolation valves specified in Table 3.6-2 shall be demonstrated OPERABLE:

a. At least once per 92 days'y cycling each power operated or automatic valve testable during plant operation through at least one complete cycle of full travel.

ST. LUCIE - UNIT 1 3/4 6-18

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

b. Immediately prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of the cycling test, above, and verification of isolation time.

4.6.3.1.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLD SHUTDONN or REFUELING MODE at least once per 18 months by:

a. Verifying that on a containment isolation test signal, each isolation valve actuates to its isolation position,

b. Cycling each power operated or automatic valve through at least one complete cycle of full travel and measuring its isolation time, and

c. Cycling each manual valve not locked, sealed or otherwise secured in the closed position through at least one complete cycle of full travel.

ST. LUCI E - UNIT 1 3/4 6-19

TABLE 3.6-2 CONTAINMENT ISOLATION VALVES Penetration Testable During Isolation W1T VII b Number Function ~P1 0 i Ti 8 A. CONTAINMENT ISOLATION

1. I-FCV-25-4,5 10 Containment purge air exhaust, CIS No 5

2. I-FCV-25-2,3 ll Containment purge supply, CIS No 5

3. I-MV-15-1 7 Primary makeup water, CIS Yes 19

4. I-MV-18-1 9 Instrument air supply, CIS No 28

5. V-6741 14 Nitrogen supply to safety injection Yes 5 tanks, CIS

6. I-HCV-14-1 EE 7 23 Reactor coolant pump cooling water No supply, SIAS

7. I-HCV-14-6 5 2 24 Reactor coolant pump cooling water No return, SIAS

8. V-2515,2516 26 Letdown line, CIS, SIAS No 5

9. V-5200,5203 28 Reactor coolant sample, CIS Yes 5

10. V-5201,5204 29 Pressurizer surge line sample, CIS Yes 5

11. V-5202,5205 29 Pressurizer steam space sample, CIS Yes 5

12. V-6554,6555 31 Containment vent header, CIS Yes 5

13. I-LCV-07-11A,11B 42 Reactor cavity sump pump discharge, Yes 10 CIS

14. V-6301,6302 43 Reactor drain tank pump suction, CIS Yes 5

15. V-2505 44 Reactor coolant pump controlled No 5 bleedoff, CIS

16. I-SE-01-1 44 Reactor coolant pump controlled No bleedoff, CIS

TABLE 3.6-2 (Continued Penetration Testable During Isolation Valve Ta Number Number Function Plant 0 eration Time Sec B. MANUAL OR REMOTE MANUAL

1. I-V-18-947 Station air supply, Manual Yes NA

2. I-V-25-11,12 56 Hydrogen purge outside air make- Yes NA up, Manual (NC)

3. I-V-25-13,14, 57 5 58 Hydrogen purge exhaust, Manual Yes 15,16 (NC)

4. V-3463 41 Safety injection tank test line, Yes Manual (NC)

5. I-V-03-1307 41 Safety injection tank test line, Yes Manual (NC)

6. V-07206, V-07189 46 Refueling cavity purification flow Yes inlet, Manual (NC)

7. V-07170, V-07188 47 Refueling cavity purification flow Yes NA outlet, Manual (NC)

8. I-FSE-27-1,2,3, 48 Hydrogen sampling line, Remote Yes 4,8,10 manual

9. I-FSE-27-5,6,7, 51 Hydrogen sampling line, Remote Yes 9,11 manual

TABLE 3.6-2 Continued Penetration Testable During Isolation Valve Ta Number Number Function Plant 0 eration Time Sec

10. I-FCV-26-1 8( 2 52a Radiation monitoring Yes NA ll. I-FCV-26-3 5 4 52b Radiation monitoring Yes

12. I-FCV-26-5 5 6 52c Radiation monitoring, return Yes 13- I-V00140(1325) 52d ILRT test tap Yes I-V00143(1325)

14. I-V00139(1322) 52e ILRT test tap Yes I-V00144(1322)

15. I-V00101(612) 54 ILRT pressure connection Yes NA NA Manual Valve-Isolation time not

May be opened on an intermittent applicable.

under administrative control.

- Normally closed valves - Isolationbasis time not applicable.

CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1 The containment hydrogen analyzer and grab sample system shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION:

With either the hydrogen analyzer or grab sample system inoperable, restore the inoperable analyzer or grab sample system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.4.1.1 The hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days by:

a. Performing a CHANNEL CALIBRATION using sample gases containing:

l. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

b. Verifying that the analyzer is aligned to receive electrical power from an OPERABLE emergency bus.

4.6.4.1.2 The grab sample system shall be demonstrated OPERABLE at least once per 92 days by using the hydrogen sample pumps to draw a sample of the containment atmosphere into the grab sample canister. The hydrogen sample pumps shall be used on an alternating basis.

3/4 6-23

CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS - W LIMITING CONDITION FOR OPERATION 3.6.4.2 Two independent containment hydrogen recombiner systems shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION'ith one hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.

SURVEILLANCE RE UIREMENTS 4;6.4.2 Each hydrogen recombiner system shall be demonstrated OPERABLE:

'a ~ At least once per 6 months by verifying during a recombiner system functional test that the minimum heater sheath temper-ature increases to > 700 F within 90 minutes and is maintained for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

b. At least once per 18 months by:

1. Performing a CHANNEL CALIBRATION of all recombiner instru-mentation and control circuits.

n 4~ Verifying through a visual examination, that there is no evidence of abnormal conditions within the recombiners (i.e., loose wiring or structural connections, deposits of foreign materials, etc.).

ST. LUCI E - UNIT 1 3/4 6-24, 0

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

3. Verifying during a recombiner system functional test that the heater sheath temperature increases to > 1200'F within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and is maintained for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. Verifying the integrity of the heater electrical circuits by performing a continuity and resistance to ground test immediately following the above required functional test.

The resistance to ground for any heater phase shall be

> 10,000 ohms.

ST. LUCIE - UNIT 1 3/4 6-25

CONTAINMENT SYSTEMS 3/4.6.5 VACUUM RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.6.5.1 The containment vessel to annulus vacuum relief valves shall be OPERABLE with an actuation setpoint of 2.25 + 0.25 inches Water Gauge differential.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one containment vessel to annulus vacuum relief valve inoperable, restore the valve to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.5.1 The containment vessel to annulus vacuum relief valves shall be demonstrated OPERABLE:

a. At least once per 6 months by verifying valve partial opening

(> 5 percent of valve ful'1 travel) and that valve operation is not restricted by corrosion, dirt, wear or debris.

b. At least once per 3 years by verifying that the valves open fully within 8 seconds at 2.25 + 0.25 inches Water Gauge differential.

ST. LUCIE UNIT 1 3/4 6-26 0

CONTAINMENT SYSTEMS 3/4.6.6 SECONDARY CONTAINMENT SHIELD BUILDING VENTILATION SYSTEM LIMITING.CONDITION FOR OPERATION 3.6.6.1. Two independent shield building ventilation systems shall be OPERABLE.

APPLICABLILITY: 'ODES 1, 2, 3 and 4.

ACTION:

With one shield building ventilation system inoperable, restore the inoperable system to OPERABLE status wi thin 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.6.1 Each shield building ventilation system shall be demonstrated OPERABLE:

a ~ At least once per 31 days on'a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.

b. At least once per 18 months or (1) after any structural main-tence on the HEPA filter or charcoal adsorber housings, or (2) following painting,, fire or chemical release in any ven-tilation zone communicating with the system by:

1. Verifying that the charcoal adsorbers remove > 99/ of a halogenated hydrocarbon r efrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm + lOX.

2. Verifying that the HEPA filter banks remove > 994 of the DOP when they-are tested in-place in accordance with ANSI N510-1975 while operating the'entilation system at a flow rate of 6000 cfm + 104.

ST. LUCIE - UNIT 1 3/4 6-27

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS,Continued

3. Verifying that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of > 90Ã for radioactive methyl iodide when the sample is tested accordance with ANSI N510-]975 (130'C, 9M R.G.). The carbon samples not obtained from test canisters shall be prepared by either,:

a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal. sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4. Verifying a system flow rate of 6000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975.

C. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation by either:

l. Verifying that a laboratory analysis of a carbon sample obtained from a test canister demonstrates a removal efficiency of > 90K for radioactive methyl iodide when the sample is tested in accordance with ANSI N510-1975 (130'C, 95K R.H.); or

2. Verifying that a laboratory analysis of at least two carbon samples demonstrate a removal efficiency of > 90Ã for radio-active methyl iodide when the samples are tested in accordance with ANSI N510-1975 (130'C, 95K R. H. ) and the samples are pre-pared by either:

a) Emptying one entire bed from a removed adsorber tray, mixing the'dsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

ST. LUCIE - UNIT 1 3/4 6-28

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demon-strated OPERABLE by also:

a) Verifying that the charcoal adsorbers remove > 99$ of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI, N510-1975 while operating the ventilation system at a flow rate of 6000 cfm + lOX, and b) Verifying that the HEPA filter banks remove > 99$ of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation .,

system at a flow rate of 6000 cfm + 10Ã.

d, At least once per 18 months by:

l. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is ( 6.15 inches Water Gauge while operating the ventilation system at a flow rate of 6000 cfm + 10%,

2. Verifying that the air flow distribution is uniform within 20K across HEPA filters and charcoal adsorbers when tested in accordance with ANSI N510-1975.

3. Verifying that the filtration system starts automatically on a Containment Isolation Signal (CIS).

4. Verifying that the filter cooling makeup air and cross con-nection valves can be manually opened.

5. Verifying that each system produces a negative pressure of

> 2.0 inches W.G. in the annulus within 2 minutes after a Containment Isolation Signal (CIS).

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 991 of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the filtration system at a flow rate of 6000 cfm + 10Ã.

f, After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove > 99K of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the filtration system at a-flow rate of 6000 cfm + 10X.

ST. LUCIE - UNIT 1 3/4 6-29

CONTAINMENT SYSTEMS SHIELD BUILDING INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.6.2 SHIELD BUILDING INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without SHIELD BUILDING INTEGRITY, restore SHIELD BUILDING INTEGRITY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY- within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.6.2 SHIELD BUILDING INTEGRITY shall be demonstrated at least once per 31 days by verifying that the door in each access opening is closed except when the access opening is being used for normal transit entry and exit.

ST. LUCIE - UNIT 1 3/4 6-30

CONTAINMENT SYSTEMS SHIELD BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.6.3 The structural integrity of the shield building shall be main-tained at a level consistent with the acceptance criteria in Specification 4.6.6.3.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION With the structural integrity of the shield building not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE RE UIREMENTS 4.6.6.3 The structural integrity of the shield building shall be deter-mined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the accessible interior and exterior surfaces of the shield building and verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation. An initial report of any abnormal degradation of the shield building detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be submitted to the Commission pursuant to Specification 6.9.1 within 90 days after completion of the surveillance requirements of this specification.

ST. LUCIE - UNIT 1 3/4 6-31

3/4.7 PLANT SYSTEMS 3.4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE.

APPLICABILITY: MODES 1,=2 and 3.

ACTION'ith both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.7.1.1 Each main steam line code safety valve. shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 4.7-1, in accordance with Section XI of the ASME Boiler 'and Pre'ssure Vessel Code, 1974 Edition.

ST. LUCIE - UNIT 1 3/4 7-1

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS Maximum Allowable Power Maximum Number of Inoperable Safety Level-High Trip Setpoint Valves on An 0 eratin Steam Generator Percent of RATED THERMAL POWER 93.2 79.8 66.5

TABLE 4.7-1 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING 2 1/ ORIFICE SIZE Header A Header B a ~ 8201 8205 1000 psia 16 in.

b. 8202 8206 1000 psia 16 in.

C ~ 8203 8207 1000 psia 16 in. 2

d. 8204 8208 1000 psia 16 in.

e. 8209 8213 1040 psia 16 in.

8210 8214 1040 psia 16 in.

9. 8211 8215 1040 psia 16 in.

h. 8212 8216 1040 psia 16 in.

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor driven feedwater pumps, and

b. One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one auxiliary feedwater pump inoperable, restore at least three auxiliary feedwater pumps (two motor driven pumps and one capable of being powered by an OPERABLE steam supply system) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.7.1.2 Each auxiliary fe'edwater pump shall be demonstrated OPERABLE:

at At least once per 31 days by:

1. Starting each pump from the control room,

2. Verifying'hat:

a) Each motor driven pump develops a discharge pressure of > 1342 psig on-recirculation flow, and b) The steam turbine driven pump develops a discharge pressure of > 1342 psig on recirculation flow.

ST. LUCIE - UNIT 1 3/4 7-4

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

3. Verifying that each pump operates for at least 15 minutes.

4. Cycling each testable power operated or automatic valve in the flow path through at least one complete cycle of full travel.

5. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 18 months during shutdown by cycling each power operated valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.

ST. LUCIE UNIT 1 3/4 7-5

PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank shall be OPERABLE with a minimum contained volume of 116,000 gallons.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the condensate storage tank inoperable, restore the condensate storage tank to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.7.1.3 The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the water level.

ST. LUCIE - UNIT 1 3/4 7-6

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system shall be

< 0.10 pCi/gram DOSE EQUIVALENT I-131.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the specific activity of the secondary coolant system > 0.10 qCi/

gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-2.

ST. LUG IE - UNIT 1 3/4 7-7

TABLE 4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT MINIMUM AND ANALYSIS FRE(RUENCY

1. Gross Activity Determination 3 times per 7 days with a maximum

.time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples

2. Isotopic Analysis for DOSE a) 1 per 31 days, when-EQUIVALENT I-131 Concentration ever the gross activity determination idicates iodine concentrations greater than 10% of the allowable limit.

b) 1 per 6 months, whenever the gross activity deter-mination indicates iodine concentrations below lOX of the allowable limit.

ST. LUCIE - UNIT 1 3/4 7-8

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODE 1 With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed .within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

MODES 2 With one main steam line isolation valve inoper'able, sub-and 3 sequent operation in MODES 1, 2 or 3 may proceed after the inoperable valve is restored to OPERABLE status or the isolation valve is maintained closed; otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.7.1.5 Each main steam line isolation valve that is open shall be demonstrated OPERABLE by:

a. Part-stroke exercising the valve at least once per 92 days, and

b. Verifying full closure within 6 seconds on any closure actua-tion signal while in HOT STANDBY with T>v > 515'F during each reactor shutdown except that verific3tion of full closure within 6 seconds need not be determined more often than once per 92 days.

ST. LUCIE - UNIT 1 3/4 7-9

PLANT SYSTEMS SECONDARY WATER CHEMISTRY LIMITING CONDITION FOR OPERATION 3.7.1.6 The secondary water chemistry shall be maintained within the limits of Table 3.7-3 by use of All Volatile Treatment (AVT).

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

(To be determined in a manner set forth in the bases in approximately six months and to be imposed by a change to this Specification.)

SURVEILLANCE RE UIREMENTS 4.7.1.6 The secondary water chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.7-3.

ST. LUCIE - UNIT 1 3/4 7-10

TABLE 3.7-3 SECONDARY WATER CHEMISTRY LIMITS Water Sample Location Parameters*

Sample locations, parameters and limits to be established in approximately 6 months based upon test program described in bases.

TABLE 4.7-3 SECONDARY WATER CHEMISTRY SURYEILLANCE RE UIREHENTS Water Sample Location Parameters*

Sample locations, parameters and frequencies to be established in approximately 6 months based upon test program descr ibed in bases.

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be > 70'F when the pressure of either cool-ant in the steam generator is > 200 psig.

APPLICABILITY: ALL MODES.

ACTION:

With the requirements of the above specification not satisfied:

a. Reduce the steam generator pressure of the applicable side to

< 200 psig within 30 minutes, and

b. Perform an analysis to determine the effect of the overpres-surization on the structural integrity of the steam generator.

Determine that the steam generator remains acceptable for continued operation prior to increasing its temperature es above 200'F.

SURVEILLANCE RE UIREMENTS 4.7.2.1 The pressure in each side of the steam generators shall be determined to be < 200 psig at least once per hour when the temperature of either the primary or secondary coolant in the steam generators is

< 700F.

ST. LUCIE - UNIT 1 3/4 7-13

PLANT SYSTEMS 3 4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1 At least two independent component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.7.3.1 At least. two component cooling water loops shall be demonstrated OPERABLE:

At least once per 31 days on a STAGGERED TEST BASIS by:

l. Starting (unless already operating) each pump from the control room.

2. Verifying that each pump develops at least 93/ of the discharge pressure for the applicable flow rate as deter-mined from the manufacturer 's Pump Performance Curve.

3. Verifying that each pump operates for at least 15 minutes.

4. Verifying that each loop is aligned to receive electrical power from separate OPERABLE emergency busses.

5. Cycling each testable power operated or automatic valve servicing safety r elated equipment through at least one complete cycle of full travel.

6. Verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed or otherwise secured in position, is in its correct position.

ST. LUCIE - UNIT 1 3/4 7-14

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

b. At least once pel 18 months during shutdown, by:

1. Cycling each power operated (excluding automatic) valve servicing safety related equipment that is not testable during plant operation, through at least one complete cycle of full travel.

2. Verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection Actuation Signal.

ST. LUCIE - UNIT 1 3/4 7-15

PLANT SYSTEMS 3/4.7.4 INTAKE COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 At least two independent intake cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.,

ACTION:

With only one intake cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.7.4.1 At least two intake cooling water loops shall be demonstrated OPERABLE:

a ~ At least once per 31 days on a STAGGERED TEST BASIS by:

1. Starting (unless already operating) each pump from the control room.

2. Verifying that each pump develops at least 935 of the discharge pressure for the applicable flow rate as deter-mined from the manufacturer's Pump Performance Curve.

3. Verifying that each pump operates for at least 15 minutes.

4. Verifying that each loop is aligned to receive electrical power from separate OPERABLE emergency busses.

5. Cycling each testable power operated or automatic valve servicing safety related equipment through at least one complete cycle of full travel.

6. Verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.

ST. LUCIE - UNIT 1 3/4 7-16

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

b. At least once per 18 months during shutdown, by:

1. Cycling each power operated (excluding automatic) valve servicing safety related equipment that is not testable during plant operation, through at least one complete cycle of full travel.

2. Verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection Actuation Signal.

ST. LUCI E - UNIT 1 3/4 7-17

PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5.1 The ultimate heat sink shall be OPERABLE with:

a. Cooling water from the Atlantic Ocean providing a water level above -10.5 feet elevation, Mean Low Water, at the plant intake structure, and

b. An average water temperature of < 96'F.

APPLICABILITY: At al 1 times.

ACTION:

With the requirements of the above specification not satisfied, be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and remove flow barriers and provide cooling water from Big Mud Creek within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.7.5.1.1 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the average water temperature and water level to be within their limits.

4.7.5.1.2 The onsite equipment capability for removing the flow barrier between the intake structure and Big Mud Creek shall be verified at least once per 7 days.

ST. LUCIE - UNIT 1 3/4 7-18

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PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7.1 The control room emergency ventilation system shall be OPERABLE with:

a. Two booster fans,

b. Two isolation valves in each outside air intake duct,

c. Two isolation valves in the toilet area air exhaust duct,

d. One filter train, and

e. At least two air conditioning units.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a ~ With one booster fan inoperable, restore the inoperable fan to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one isolation valve per air duct inoperable, operation may continue provided the other isolation valve in the same duct is maintained closed; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With the filter train inoperable, restore the filter train to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With only one air conditioning unit OPERABLE, restore at least two air conditioning units to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ST. LUCIE - UNIT 1 3/4 7-20

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS 4.7.7.1 The control room emergency ventilation system shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air temperature is < 120'F.

b. At least once per. 31 days by:

l. Initiating flow through the HEPA filter and charcoal adsorber train and verifying that each booster fan operates for at least 15 minutes.

2. Starting (unless already operating) each air conditioning unit and verifying that it operates for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

At least once per 18 months or (1) after any structural main-tenance on the HEPA filter or charcoal adsorber housing, or (2) following painting, fire or chemical release in any venti-lation zone communicating with the system by:

1. Verifying that the charcoal adsorbers remove > 995 of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + 104.

2. Verifying that the HEPA filter banks remove > 99/ of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + 10Ã.

3. Verifying that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of > 90Ã for radioactive methyl iodide when the sample is tested in accordance with ANSI N510-1975 (130 C, 95% R.H.). The carbon'amples not obtained from test canisters'hall be prepared by either:

a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or ST. LUCIE - UNIT 1 3/4 7-21

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4. Verifying a system flow rate of 2000 cfm + 10/ during system operation when tested in accordance with ANSI N510-1975.

d. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation by either:

1. Verifying that a laboratory analysis of a carbon sample obtained from a test canister demonstrates a removal efficiency of > 90Ã for radioactive methyl iodide when the sample is tested in accordance with ANSI N510-1975 (130'C, 95Ã R. H. ); or

2. Verifying that a laboratory analysis of at least two car-bon samples demonstrate a removal efficiency of > 90K for radioactive methyl iodide when the samples are tested in accordance with ANSI N510-1975 (130'C, 95%%d R.H. ) and the samples are prepared by either:

a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of'he bed.

Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demonstrat-ed OPERABLE by also:

a) Verifying that the charcoal ad'sorbers remove > 99K of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + 10K, and b) Verifying that the HEPA filter banks remove > 99K of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + lOX.

ST. LUCIE - UNIT 1 3/4 7-22

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued e.. At least once per 18 months by:

l. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 4.15 inches Water Gauge while operating the ventilation system at a flow rate of 2000 cfm + lOX.

2. Verifying that on a containment isolation signal or chlorine accident detection signal, the system automatically isolates the control room within 35 seconds and switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks.

3. Verifying that the system maintains the control room at a positive pressure of > 1/8 inch W.G. relative to the outside atmosphere during system operation with < 100 cfm outside air intake.

f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99/ of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + 10Ã.

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove

> 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + 10/.

ST. LUCIE UNIT 1 3/4 7-23

PLANT SYSTEMS 3/4.7.8 ECCS AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8.1 Two independent ECCS area exhaust air filter trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one ECCS area exhaust air filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in, COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE RE UIREMENTS 4.7.8.1 Each ECCS area exhaust air filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiat-ing, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

l. Verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%.

2. Verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%.

ST. LUCIE - UNIT 1 3/4 7-24

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

3. Verifying that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of > 90% for radio-active methyl iodide when the sample Vs tested in accordance with ANSI N510-1975 (130'C, 95% R.H.). The carbon samples not obtained from test canisters shall be prepared by either:

a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4. Verifying a system flow rate of 30,000 cfm + 10Ã during system operation when tested in accordance with ANSI N510-1975.

C. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation by either:

l. Verifying that a laboratory analysis of a carbon sample

.obtained from a test canister demonstrates a removal efficiency of > 90K for radioactive methyl iodide when the sample is tested in accordance with ANSI N510-1975 (130'C, 95% R. H. ); or

2. Verifying that a laboratory analysis of at least two carbon samples demonstrate a removal efficiency of > 90% for radioactive methyl iodide when the samples are tested in, accordance with ANSI N510-1975 (130'C, 95K R. H. ) and the samples are prepared by either:

a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

ST. LUCIE - UNIT 1 3/4 7-25

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demonstrat-ed OPERABLE by also:

a) Verifying that the charcoal adsorbers remove > 99%%d of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%%d, and b) Verifying that the HEPA filter banks remove > 99%%d of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10/.

d. At least once per 18 months:

l. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 4.15 inches Water Gauge while operating the ventilation system at a flow rate of 30,000 cfm + 10/.

2. Verifying that the air flow distribution is uniform within 20/ across HEPA filters and charcoal adsorber s when tested in accordance with ANSI N510-1975.

3. Verifying that the filter train starts on a Safety Injec-tion Actuation Signal.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99%%d of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%%d.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove

> 99%%d of a halogenated hydrocarbon refrigerant test gas when they. are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10/.

ST. LUG IE - UNIT 1 3/4 7-26

PLANT SYSTEMS 3/4.7.9 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.9.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of > 0.005 microcu-ries of removable contamination.

APPLICABILITY: At al 1 times.

ACTION:

a 0 Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:

1 . Ei ther decontaminated and repaired, or

2. Disposed of in accordance with Commission Regulations.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 41 411 ~fd 1 leakage and/or contamination by:

-1 4 fd 4111 <<df

a. The licensee, or

b. Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

4.7.9.1.2 Test Fre uencies - Each category of sealed sources shall be tested at the frequencies described below.

a. Sources in use excludin startu sources reviousl subjected to core flux - At least once per six mont s for a scale sources containing radioactive material:

ST. LUCIE - UNIT 1 3/4 7- 27

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

1. With a half-life greater than 30 days (excluding Hydrogen 3), and

2. In any form other than gas.

b. Stored sources not in use - Each sealed source shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources transferred without a certificate indicating the last, test date shall be tested prior to being placed into use.

c. Startu sources - Each sealed startup source shall be tested wit sn 3 ays prior to being subjected to core flux and following repair or maintenance to the source.

4.7.9.1.3 ~Re orts

- A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days if source leakage tests reveal the presence of > 0.005 microcuries of removable contamination.

ST. LUG IE - UNIT 1 3/4 7-28

PLANT SYSTEMS 3/4.7.10 HYDRAULIC SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.10.1 All hydraulic snubbers listed in Table 3.7-2 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more hydraulic snubbers inoperable, restore the inoperable snubber(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.7.10.1.1 Each hydraulic snubber with seal material fabricated from ethylene propylene or other materials demonstrated compatible with the operating environment and approved as such by the NRC, shall be deter-mined OPERABLE at least once after not less than 4 months but within 6 months of initial criticality and in accordance with the inspection schedule of Table 4.7-3 thereafter, by a Visual inspection of the snubber.

Visual inspections of the snubber s shall include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connec-tions, and linkage connections to the piping and anchors. Initiation of the Table 4.7-3 inspection schedule shall be made assuming the unit was previously at the 6 month inspection interval.

4.7.10.1.2 Each hydraulic snubber with seal material not fabricated from ethylene propylene or other materials demonstrated compatible with the operating environment shall be determined OPERABLE at least once per 31 days by a visual inspection of the snubber. Visual inspections of the snubbers shall include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connections, and linkage connections to the piping and anchors.

ST. LUCIE - UNIT 1 3/4 7-2g

PLANT SYSTEMS HYDRAULIC SNUBBERS Continued SURVEILLANCE RE UIREMENTS Continued 4.7.10.1.3 During shutdown, 18 months after initial criticality and at least once per 18 months thereafter, a representative sample of at least 10 snubbers or at least 105 of all snubbers listed in Table 3.7-2, which-ever is less, shall be selected and functionally tested to verify correct piston movement, lock up and bleed. Snubbers selected for functional testing shall be selected on a rotating basis except snubbers identified in Table 3.7-2 as either "Especially Difficult to Remove" or in "High Radiation Zones" may be exempted from functional testing provided these snubbers were demonstrated OPERABLE during previous functional tests.

Snubbers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each snubber found inoperable during these functional tests, an additional minimum of 10K of all snubbers or 10 snubbers, whichever is less, shall also be functionally tested until no more failures are found or all snubbers have been functionally tested.

0 ST. LUCIE - UNIT 1 3/4 7-30

TABLE 3;7-2 SAFETY RELATED HYDRAULIC SNUBBERS*

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE TO REMOVE Yes or No Yes or No SS-1 1A MS, Steam Gen. 1A, Elev. I No Yes 62'en.

SS-2 1A MS, Steam 1A, Elev. I No Yes 62'en.

SS-3 1A MS, Steam 1A, Elev. I No Yes 62'en.

SS-4 1A MS, Steam 1A, Elev. I No Yes 62'en.

SS-5 1A MS, Steam 1A, flev. I No Yes 62'en.

SS-6 1A MS, Steam 1A, Elev. I No Yes 62'orI 62'en.

SS-7 1A MS, Steam 1A, Elev. I No Yes 62'en.

SS-8 1A MS, Steam 1A, Elev. I No Yes 62'en.

SS-1 1B MS, Steam 1B, Elev. I No Yes 62'en.

SS-2 lB MS, Steam 1B, Elev. ! No Yes 62'en.

SS-3 1B MS, Steam 1B, Elev. I No Yes 62'en.

SS-4 1B MS, Steam 1B, Elev. I No Yes 62'en.

SS-5 1B MS, Steam 1B, Elev. I No Yes 62'en.

SS-6 1B MS, Steam 1B, Elev. I No Yes 62'en.

SS-7 1B MS, Steam 1B, Elev. I No Yes 62'en.

SS-8 1B MS, Steam 1B, Elev. I No Yes

TABLE 3;7-2 SAFETY RELATED HYDRAULIC SNUBBERS*

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON LOCATION AND ELEVATION INACCESSIBLE ZONE TO REMOVE or I Yes or No Yes or No 1 Al RC, RCP Motor 1A1, Elev.

57'CP I No No lA2 RC, Motor lA2, Elev.

57'CP I No No 1Bl RC, Motor 181, Elev.

57'CP I No No 1B2 RC$ Motor 1B2, Elev. I No No 005-34A Bldg, Elev.

57'eactor RC RC$ A No No 005-34B Bldg, Elev.

68'eactor RC RC, A No No 005-36 Bldg, Elev.

68'eactor RC RC, A No No 005-12B Bldg, Elev.

80'eactor RC RC$ A No No 005-12A Bldg, Elev.

80'eactor RC RC, A No No 005-55C Bldg, Elev.

80'eactor RC RC, A No No 005-55B Bldg, Elev.

80'eactor RC RC$ A No No 005-62A Bldg, Elev.

80'eactor RC RC, A No No 005-89 Bldg, Elev.

80'eactor RC RC, A No No 005-90 Bldg, Elev.

80'eactor RC RC, .A No No 005-98 Bldg, Elev. 80' 80'eactor RC RC, A No No

TABLE 3.7-2 SAFETY RELATED HYDRAULIC SNUBBERS*

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE TO 'REMOVE or I Yes or No Yes or No MS 649-319 MS, Reactor Bldg,.Elev. A No No MS 548-5 MS, Bldg, Elev.

82'eactor A No No 82'.S.

MS 1076-3164 MS, Trestle, Elev. A No No 649-314 Bldg, Elev.

62'eactor MS MS, No No 649-314 Bldg, Elev.

55'eactor MS MS, No No 649-310 Bldg, Elev.

55'eactor MS MS, No No 548-16A Bldg, Elev.

50'eactor MS MS, No Yes MS 548-9 MS, Bldg, Elev.

30'eactor No Yes 5,48-9 Bldg, Elev.

50'eactor MS MS, No Yes 549-7 Bldg, Elev.

50'eactor BF BF, No No 549-7 Bldg, Elev.

40'eactor BF BF, No No 549-8 Bldg, Elev.

40'eactor BF BF, No Yes BF 549-11 Bldg, Elev.

40'eactor BF, No No BF 549-11 Bldg, Elev.

50'eactor BF, No No 549-17 Bldg, Elev.

50'eactor BF BF, A No Yes 661-407 Bldg, Elev. I 36'eactor BF BF, No No 661-407 Bldg, Elev. 40' I 40'eactor BF BF, No No

TABLE 3.7-2 SAFETY RELATED HYDRAULIC SNUBBERS*

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE TO REMOVE or I Yes or No Yes or No BF 661-416 BF., Reactor Bldg, Elev, I No No Elev, 50'ldg, BF 661-416 BF, Reactor I No No 50'ldg, BF 661-4020 BF, Reactor Elev. A No No SI, Reactor Elev. I 36'ldg, SI 968-210 No No SI, Elev.

16'ldg, SI 968-565 Reactor A No No SI, 25'ldg, SI 968-1205 Reactor Elev. A No No SI, Elev.

30'ldg, SI 968-1207 .Reactor A No No SI, Elev. I 18'ldg, SI 969-1 190 Reactor No No SI, Elev.

20'ldg, SI 969-1216 Reactor A No No SI, 18'ldg, SI 969-6193 Reactor Elev. A No No SI, Elev.

18'ldg, SI 969-6195 Reactor A No No SI, flev.

18'ldg, SI,969-6198 Reactor A No No SI, Elev.

18'ldg, SI 969-6201 Reactor A No No SI, Elev.

18'ldg, SI 969-6217 Reactor A No No SI, Elev.

18'ldg, SI 969-6217 Reactor A No No SI, I 18'ldg, SI 970-1210 Reactor Elev. No No SI, Elev. 20' I 33'ldg, SI 970-1248 Reactor No No

TABLE 3.7-2 SAFETY RELATED HYDRAULIC SNUBBERS*

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON; LOCATION AND ELEVATION INACCESSIBLE ZONE TO REMOVE or I Yes or No Yes or No SI 970-1251 SI, Reactor Bldg, A No No 20'lev.

SI 971-6 SI, Reactor Bldg, I No No 20'lev.

SI 971-1229 SI5 Reactor Bldg, I No No 20'lev.

SI 971-6229 SI$ Reactor Bldg, I No No 20'lev.

SI 971-6236 SI, Reactor Bldg, 20'1 I No No SI 972-1243 SI, Reactor Bldg, 4'lev.

ev. A No No I

25'lev.

SI 972-6240 SI, Reactor Bldg, No No 16'lev.

SI 973-240 SI, Reactor Bldg, A No No I

18'lev.

SI 973-6219 SI, Reactor Bldg, No No 18' 36'lev.

SI 973-6224 SI, Reactor Bldg, A No No SI 868-64 SI, RAB, Elev. A No No Elev.

4'AB, SI 868-111 SIO A No No SI, Elev.

4'AB, SI 868-163 A No No 4-'AB, SI 868-410 SI, Elev. A No No SI, 4'AB, SI 676-67 Elev. A No No SI, Elev.

4'AB, SI 676-67 A No No Elev.

4'AB, SI 676-105 SI> A No No

TABLE 3.7-2 SAFETY RELATED HYDRAULIC SNUBBERS*

SNUBBER SYSTEM SNUBBER INSTALLED OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE TO REMOVE A or I Yes or No Yes or No SI 676-105 SI, RAB, Elev. A No No 4'I, SI 676-127 RAB, Elev. A No No 4'I, SI 676-129 RAB, Elev. A No No 4'I, SI 676-250 RAB, Elev. 125'CCESSIBLE A No No 24'I, SI 676-2475 RAB, Elev. A No No 30'I, SI 676-2475A RAB, Elev. A No No 30'I, SI 676-4505 RAB, Elev. A No No 7'I, SI-V-1 RAB, Elev. A No No 4'I, SI-V-1 RAB, Elev. A No No SPS-41 7 Spray, Reactor 4'ressurizer Bldg, Elev, No No SPS-27 Spray, Reactor 50'ressurizer Bldg, Elev,. No No SPS 467 Spray, Reactor 50'ressurizer Bldg, Elev. No No SPS-777 Spray, Reactor 80'ressurizer Bldg, Elev. 80'S, A No No CS-832-118 Reactor Bldg, flev. A No Yes

TABLE 3.7-2 SAFETY RELATED HYDRAULIC SNUBBERS*

SNUBBER SYSTEM SNUBBER INSTALLED OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE TO REMOVE Aor I Yes or No Yes or No CS-878-115 CS, Reactor Bldg, Elev. A No No 18'C, CC-1865-9 Reactor Bldg, Elev. A No No CC-1899-48 Reactor 23'CCESSIBLE Bldg, Elev.

25'C, 25'C, A No No CC-1899-2208 Reactor Bldg, Elev. A No No 59'C, CC-1852-6241 Reactor Bldg, Elev.. 25'C, A No No CC-1865-2207 Reactor Bldg, Elev. A No No 59'C, CC-17-1 RAB, Elev. A No No CC-14-2 Elev. 20'C,-RAB, A No No

'C-21-.1 Elev. 20" 26'C;.RAB;-

A No No CC-21-5 CC, RAB, Elev. A' No No CC-23'-;2. 26'C;'RAB,".Elev.

No- No 26'H, CH-3-40 RAB, Elev. No No 34'H, CH-3-54 RAB, Elev. No No.-

21'H, CH-3-54A RAB, Elev. 21'H,

,A No No CH-3-75 RAB, Elev. No No

TABLE 3.7-2 SAFETY RELATED HYDRAULIC SNUBBERS*

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO.. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE TO REMOVE or I Yes or No Yes or No MS-649-313 MS, Reactor Bldg, Elev. I No No MS, Reactor'Bldg, Elev. 80' 80'S-649-313 No No

Snubbers may be added to safety related systems without prior License Amendment to Table 3.7-.2 provided that safety evaluations, documentation and reporting are provided in accordance with 10 CFR 50.59 and that a revision to Table 3.7-2 is included with a subsequent License Amendment request.

TABLE 4.7-3 HYDRAULIC SNUBBER INSPECTION SCHEDULE NUMBER OF SNUBBERS FOUND INOPERABLE NEXT REQUIRED DURING INSPECTION OR DURING INSPECTION INTERVAL* INSPECTION INTERVAL**

0 18 months + 25K 1 12 months + 25/

2 6 months + 25'A 124 days + 25'4 3 or 4 5,6,or7 62 days 31 days

+

25%

25K

)8

Snubbers may be categorized into two groups, "accessible" and "inaccessible". This categorization shall be based upon the snubber's accessibility for inspection during reactor operation. These two groups may be inspected independently according to the above schedule.

- The required inspection interval shall not be lengthened more than one step at a time.

/

0

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE:

1 - 125-volt D.C. bus, and 1 - 125-volt battery bank and charger supplying the above D. C. bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the above complement of D.C. equipment and bus OPERABLE, establish CONTAINMENT INTEGRITY within 8 hours.

SURVEILLANCE RE UIREMENTS 4.8.2.4.1 The above required 125-volt D.C. bus shall be determined OPERABLE and energized at least once per 7 days by verifying indicated power availability.

4.8.2.4.2 The above required 125-volt battery bank and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.

ST. LUCIE

~

- UNIT 1 3/4 8-13.

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS (Continued

2. The pilot cell specific gravity, corrected to 77'F, is

> 1.20.

3. The pilot cell voltage is > 2.084 volts.

4. The overall battery voltage is > 125 volts..

b. At least once per 92 days by verifying that:

1. The voltage of each connected cell is > 2.084 volts under float charge and has not decreased more than 0.14 volts from the value observed during the original acceptance test.

2. The specific gravity, corrected to 77'F, of. each con-nected cell is > 1.20 and has not decreased more than 0.01 from the average of the connected cells at the time of inspection.

3. The electrolyte level of each connected cell is between the minimum and maximum level indication marks.

C. At least once per 18 months by verifying that:

1. The cells, cell plates and battery racks show no visual indication of physical damage or deterioration.

2. The cell-to-cell and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

ST LUCIE - UNIT 1 3/4 8-11

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIRENENTS Continued 0

d. At least once per 18 months. during shutdown, by verifying that the battery capacity. with the charger disconnected, is adequate to either:

1. Supply and maintain in OPERABLE status all of the actual emergency loads for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when the battery is subjected to a battery service test, or

2. Supply a dummy load of the following profile for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months while maintaining the battery terminal voltage

> 100 volts:

a) > 773 amperes during the initial 40 seconds of the test, b) > 570 amperes during the remainder of the first hour of the test, and c) > 145 amperes during the remainder of the 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

At the completion of this battery test, the battery charger shall be demonstrated capable of recharging its battery at a rate of < 155 amperes while supplying normal D.C. loads. The battery shalT be charged to at least 95% capacity in < 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 805 of the manufacturer's rating when subjected to a performance discharge test. This performance discharge test shall be performed subsequent to the satisfactory completion of the required battery service test.

ST LUCI E - UNIT 1 3/4 8-12

ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energized from sources of power other than a diesel gene-rator set but aligned to an OPERABLE diesel generator set:

1 - 4160 volt Emergency Bus 1 - 480 volt Emergency Bus 3 - 480 volt Emergency MCC Busses 2 - 120 volt A.C. Instrument B'usses APPLICABILITY: MODES 5 and 6 ACTION:

With less than the above complement of A.C. busses OPERABLE and energized, establish CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE RE UIREMENTS 4.8.2.2 The specified A.C. busses shall be determined OPERABLE and energized from A. C. sources other than the diesel generators at least once per 7 days by verifying indicated power availability.

ST. LUCIE - UNIT 1 3/4 8-9

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.3 The following D.C. bus trains shall be energized and OPERABLE:

TRAIN "A" consisting of 125-volt D. C. bus No. 1A, 125-volt D. C.

battery bank No. 1A and a full capacity charger.

TRAIN "B" consisting of 125-volt D. C, bus No. 1B, 125-volt D.C.

battery bank No. 1B, and a full capacity charger.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one 125-volt D.C. bus inoperable, restore th'e inoperable bus to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With a 125-volt D.C. battery and/or an associated charger inoperable, restore the inoperable battery and/or charger to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.8.2.3.1, Each D.C. bus train shall be determined OPERABLE and energized at least once per 7 days by verifying indicated power availability.

4.8.2.3.2 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1. The electrolyte level of each pilot cell is between the minimum and maximum level indication marks.

ST. LUCIE - UNIT 1 3/4 8-10

ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class lE distribution system, and

b. One diesel generator set with:

l. Engine-mounted fuel tanks containing a minimum of 152 gallons of fuel,

2. A fuel storage system containing a minimum of 16,450 gallons of fuel, and

3. A fuel transfer pump.

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the minimum required A. C. electrical power sources are restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2a.5.

ST. LUCIE - UNIT 1 3/4 8-7

ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generator sets:

4160 volt Emergency Bus lA3 4160 volt Emergency Bus 1B3 480 volt Emergency Bus lA2 480 volt Emergency Bus 1B2 480 volt Emerg'ency MCC Busses lA5, 1A6, lA7 480 volt Emergency MCC Busses 1B5, 1B6, 1B7 120 volt A.C. Instrument Bus 1MA 120 volt A.C. Instrument Bus 1MB 120 volt A.C. Instrument Bus 1MC 120 volt A.C. Instrument Bus IMD APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With less than the above complement of A. C. busses OPERABLE, restore the inoperable bus to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying indicated power availability.

ST. LUCIE - UNIT 1 3/4 8-8

ELECTRICAL POWER SYSTEHS SURVEILLANCE RE UIREHENTS Continued voltage of 2500 volts. The megger test voltage shall be applied for at least one minute or until the reading remains steady for at least 15 seconds. The conductor to ground isolation resistance shall be verified to be at least 100 megohms.

2. -D.C. proof testing at > 25,000 volts each of the three installed spare 5000 volt cables (one each in the ducts between the switchgear and 1) the diesel generators,

2) the component cooling water pump motors, and 3) the intake cooling water pump motors);and verifying that for each of the cables, the measured leakage current, when monitored at nominal 30 second intervals for at least 10 minutes, does not increase after an initial current decrease and stabilization. If any one of the installed spare 5000 volt cables fail the D.C. proof test, the following actions shall be performed prior to increas-ing the RCS T above 200'F:

a) An inservice cable in the same specific category shall be disconnected, designated as the replacement spare cable, and subjected to the same D.C. proof test. If the results of testing this cable are acceptable, this cable shall be retained as the installed spare while the original spare cable shall be removed and replaced.

The newly installed cable shall be connected and placed inservice as the replacement for the newly designated spare. However, if the results of testing this cable are unsatisfactory, all Class lE 5000 volt cables in the same specific duct run category shall be D.C. proof tested at > 25,000 volts.

b) The cause of the failure(s) in the cables(s) shall be determined and reported to the Commission for evaluation and acceptability. If the failure was either caused by the cable's operating environment or if it was a generic type failure, all Class 1E 5000 volt underground cable installations shall be improved to a level that will significantly reduce the factor(s) identified to be the cause of failure, or all Class lE 5000 volt underground cables shall be replace by a type demonstrated to be acceptable.

ST. LUCIE - UNIT 1 3/4 8-5

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS Continued C. At least once per 18 months by megger testing a sample of the 600 volt and lower voltage Class lE underground cables at a minimum test voltage of 1000 volts and verifying a minimum conductor to ground isolation re-sistance of 25 megohms with all conductors in the control cable assemblies, except the one under test, grounded. The megger test voltage shall be applied for at least one minute or until the reading remains steady for at least 15 seconds.. The cables selected for megger testing shall include at least one of each type of cable (cables shall be categorized ac-cording to construction and materials used in fabrica-tion) in each of the ducts between the switchgear and 1) the diesel generators, 2) the component cooling water pump motors,. and 3) the intake. cooling water pump motors.

The cables selected for megger testing shall be selected on a rotating basis.

ST LUCIE - UNIT 1 3/4 8-6 0

ELECTRICAL POWER SYSTEM SURVEILLANCE RE UIREMENTS Continued 4.8.1.1.2 Each diesel generator set shall be demonstrated OPERABLE:

a. At least once per 31 days, on a STAGGERED TEST BASIS by:

l. Verifying the fuel level in the engine-mounted fuel tank.

2. Verifying the fuel level in the fuel storage tanks.

3. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the engine-mounted tank.

4. Verifying the diesels start from ambient condition.

5. Verifying the generator is synchronized, loaded to > 1300 kw, and operates for > 60 minutes.

6. Verifying the diesel generator set is aligned to provide standby power to the associated emergency busses.

b. At least once per 31 days by verifying that a sample of diesel fuel from the fuel storage tank is within the acceptable limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water and sediment.

C. At least once per 18 months during shutdown by:

l. Subjecting the diesels to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2. Verifying the generator capability to reject a load of >

600 hp without tripping.

3. Simulating a los's of offsite power in conjunction with a safety injection actuation signal, and:

lt a) Verifying de-energization of the emergency busses and load. shedding from the emergency busses.

b) Verifying the diesels start from ambient condition on the auto-start signal, energize the emergency busses with permanently connected loads, energize ST. LUCIE - UNIT 1 3/4 8-3

ELECTRICAL POWER SYSTEHS SURVEILLANCE RE UIREMENTS Continued the auto-connected emergency loads through the load sequencing system and operate for > 5 minutes while the generator is loaded with the emergency loads.

c) Verifying that on the safety injection actuation signal, all diesel generator trips, except engine overspeed and generator differential, are automa-tically bypassed.

4. Verifying the diesel generator set operates for > 60 minutes while loaded to > 3500 kw.

5. Verifying that the auto-connected loads to each diesel generator set do not exceed the 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months rating of 3730 kw.

6. Verifying that the automatic sequence timers are OPERABLE with each load sequence time within' 10% of its required value.

d. At least once per 18 months by verifying that each fuel transfer pump transfers fuel from each fuel storage tank to the engine mounted fuel tanks on each diesel via the installed cross con-nection lines.

4.8.1.1.3 The Class lE underground cable system shall be demonstrated OPERABLE:

a ~ Within 30 days after the movement of any loads in excess of 80/

of the ground surface design basis load over the cable ducts by pulling a mandrel with a diameter of at least 80Ã of the duct's inside diameter through a duct exposed to the maximum loading (duct nearest the ground's surface) and verifying that the duct has not been damaged.

b. At least once per 18 months, during shutdown, by:

1. Selecting on a rotating basis at least 3 (one each in the ducts between the diesel generators and the switchgear, between the switchgear and"the component cooling water pump motors, and between the switchgear and the intake cooling water pump motors) Class 1E -5000 volt underground cables and megger testing the selected cables at a minimum test ST. LUCIE - UNIT 1 3/4 8-4

3/4.8 ELECTRICAL POWER SYSTEMS 0 3/4.8.1 OPERATING A.C. SOURCES LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C.. electrical power sources shall be OPERABLE:

a. Two physically independent circuits between the offsite trans-mission network and the onsite Class 1E distribution system, and

b. Two separate and independent diesel generator sets each with:

1. Engine-mounted fuel tanks containing a minimum of 152 gallons of fuel,

2. A separate fuel storage system containing a minimum of 16,450 gallons of fuel, and

3. A separate fuel transfer pump.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a ~ With either an offsite circuit or diesel generator set of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.l.l.l.a and 4.8.1.1.2.a.4 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least two offsite circuits and two diesel generator sets to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one offsite circuit and one diesel generator set of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8. l. l. l.a and 4.8.1.1.2.a.4 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 ST. LUCIE - UNIT 1 3/4 8-1

ELECTRICAL POWER SYSTEMS ACTION Continued hour s. Restore at least two offsite circuits and two diesel generator sets to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours'.

With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generator sets by performing Surveillance Requirement 4.8. 1. 1.2. a. 4 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the diesel generator sets are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With two of the above required diesel generator sets in-operable, demonstrate the OPERABILITY of two offsite A.C.

circuits by performing Surveillance Requirement 4. 8. l. 1. l.a within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable diesel generator sets to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore at least two diesel generator sets to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.8. 1. 1. 1 Two physically independent .circuits between the offsite transmission network and the onsite Class 1E distribution system shall be determined OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying correct breaker alignments and indicated power availability.

ST. LUCI E - UNIT 1 3/4 8-2 0

3/4. 9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling cavity shall be maintained uniform and sufficient to ensure that the more restrictive of following reactivity conditions is met:

a. Either a Keff of 0.95 or less, which includes a 1/ Ak/k conservative allowance for uncertainties, or

b. A boron concentration of > 1720 ppm, which includes a 50 ppm conservative allowance for uncertainties.

APPLICABILITY: MODE 6*.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at > 40 gpm of 1720 ppm boron or its equivalent until K qq is reduced to < 0.95 or the boron concentratio'n is restored to > )720 ppm, whichever is the more restrictive.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.1.1. The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full length CEA in excess of 3 feet from its fully inserted position.

4.9.1.2 The boron concentration of the refueling cavity shall be determined by chemical analysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

"The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

ST. LUCIE - UNIT 1 3/4 9-1

REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two wide range logarithmic neutron flux monitors shall be operating, each with continuous visual indication in the con-trol room and one with audible indication in the containment; APPLICABILITY: MODE 6.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.2 Each wide range logarithmic neutron flux monitor shall be demonstrated OPERABLE by per formance of:

a. A CHANNEL FUNCTIONAL TEST at least once per 7 days.

b. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the start of CORE ALTERATIONS, and

c. A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.

ST. LUCIE - UNIT 1 3/4 9-2

REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for a minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

APPLICABILITY: During movement of irradicated fuel in the reactor pressure vessel.

ACTION:

With the reactor subcritical for less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , suspend all operations involving movement of irradiated fuel in the reactor pres-sure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

ST. LUCI E - UNIT 1 3/4 9-3

REFUELING OPERATIONS CONTAINMENT PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment penetrations shall be in the following status:

a. The equipment door closed and held in place by a minimum of four bol ts,

b. A minimum of one door in each airlock is closed,'nd

c. Each penetration, except as provided in Table 3.6-2 of Specification 3.6.3.1, providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. Closed by an isolation valve, blind flange, or manual valve, or

2. Be capable of being closed by an OPERABLE automatic containment isolation valve.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immedi-ately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.4 Each of the above required containment penetrations shall be determined to be either in its closed/isolated condition or capable of being closed by an OPERABLE automatic containment isolation valve within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the star t of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment by:

a. Verifying the penetrations are in their closed/isolated condition, or

b. Testing the containment isolation valves per the applicable portions of Specifications 4.6.3.1.1 and 4.6.3.1.2.

ST. LUCIE - UNIT 1 3/4 9-4

REFUELING OPERATIONS COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.

APPLICABILITY: During CORE ALTERATIONS.

ACTION:

When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS.

The provisions fo Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the star t of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .during CORE ALTERATIONS.

3/4 9-5

REFUELING OPERATIONS MANIPULATOR CRANE OPERABILITY LIMITING CONDITION FOR OPERATION 3.9.6 The manipulator crane shall be used for movement of CEAs or fuel assemblies and shall be OPERABLE wi-th:

a. A minimum capacity of 2000 pounds, and

b. An overload cut off limit of < 3000 pounds.

APPLICABILITY: During movement of CEAs or fuel assemblies within the reactor pressure vessel.

ACTION:

With the requirements for crane OPERABILITY not satisfied, suspend use of any inoperable manipulator crane from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.6 The manipulator crane used for movement of CEAs or fuel assem-blies within the reactor pressure vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pr'ior to the start of such operations by performing a load test of at least 2500 pounds and demonstrating an automatic load cut off when the crane load exceeds 3000 pounds.

ST. LUCIE - UNIT 1 3/4 9.-6

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REFUELING OPERATIONS COOLANT CIRCULATION LIMITING CONDITION FOR OPERATION 3.9.8 At least one shutdown cooling loop shall be in operation.

APPLICABILITY: MODE 6.

a. With less than one shutdown cooling loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all contain-ment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The shutdown cooling loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

C. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.8 A shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at a flow rate of ) 3000 gpm at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ST. LUCIE - UNIT 1 3/4 9-8 0

REFUELING OPERATIONS CONTAINMENT ISOLATION SYSTEM LIMITING CONOITION FOR OPERATION 3.9.9 The containment isolation system shall be OPERABLE.

APPLICABILITY: MODE 6.

ACTION:

With the containment isolation system inoperable, close each of the penetrations providing direct access from the containment atmosphere to the outside atmosphere. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.9 The containment isolation system shall be demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment isolation occurs on manual initiation and on a high radiation signal from two of the containment radiation monitoring instrumentation channels.

ST. LUCIE - UNIT 1 3/4 9-9

REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least'23 feet of water shall be maintained over the top of :

irradiated fuel assemblies seated within the reactor pressure vessel.

APPLICABILITY: During movement of fuel assemblies or CEAs within the reactor pressure vessel'while in MODE 6.

ACTION:

r With the requirements of the" above specification not satisfied, suspend all operations involving movement of fuel assemblies or CEAs wi'thin the pressure vessel.

SURVEILLANCE'E UIREMENTS 4.9.10 The water level shall be determined to 'be, at least'ts minimum required depth within'2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months pr,ior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movement of fuel. assemblies or CEAs..

ST. LUCIE - UNIT 1 3/4 9-10

REFUELING OPERATIONS STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 As a minimum, 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool .

ACTION:

With the'requirement of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to.within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.

ST. LUCIE - UNIT 1 3/4 9-11

REFUELING OPERATIONS FUEL POOL VENTILATION SYSTEM FUEL STORAGE LIMITING CONDITION FOR OPERATION 3.9.12 At least one fuel pool ventilation system shall be OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in the spent fuel pool.

ACTION:

a ~ With no fuel pool ventilation system OPERABLE, suspend all operations involving movement of fuel within the spent fuel pool or crane operation with loads over the spent fuel pool until at least one fuel pool ventilation system is restored to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.12 The above required fuel pool ventilation system shall be demonstrated OPERABLE:

a. At least once per 31 days. by ini,tiating flow through the HEPA filter and charcoal adsorber train,and verifying that the train operates for at least 15 minutes.

b. At least once per 18 months or (1) after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ven-tilation zone communicating with the system by:

ST. LUCIE - UNIT 1 3/4 9-12

REFUELING OPERATIONS SURVEILLANCE RE UIREMENTS Continued

l. Verifying that the charcoal adsorbers remove > 99/ of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 10,350 cfm + lOX.

2. Verifying that the HEPA filter banks remove > 99% of the DOP, when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 10,350 cfm + 10Ã.

3. Verifying that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of > 70/ for radioactive elemental iodide when the sample is tested in accordance with ANSI N510-1975 (130'C, 95/ R. H. ). The carbon samples not obtained from test canisters shall be prepared by either:

a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of'he bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4. Verifying a system flow rate of 10,350 cfm + 10Ã during system operation when tested in accordance with ANSI N510-1975.

ST. LUCIE - UNIT 1 -3/4 9-13

REFUELING OPERATIONS SURVEILLANCE RE UIREMENTS Continued

c. At least once per 18 months by:

l. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 4.15 inches Water Gauge while operating the ventilation system at a flow rate of 10,350 cfm + 10Ã.

2. Verifying that the air flow distribution is uniform within 20% across HEPA filters and charcoal adsorbers when tested in accordance with ANSI N510-1975.

3. Verifying that the ventilation system maintains the spent fuel storage pool area at a negative pressure of > 1/8 inches Water Gauge relative to the outside atmosphere during system operation.

d. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 10,350 cfm + 105.

e. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove

> 995 of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-]975 while operating the ventilation system at a flow rate of 10,350 cfm + 10%.

ST. LUG IE - UNIT 1 3/4 9-14

REFUELING OPERATIONS LIMITING CONDITION FOR OPERATION 3.9.13 The maximum load which may be handled by the spent fuel cask crane shall not exceed 25 tons.

APPLICABILITY: Whenever irradiated fuel assemblies, are in the storage

~p ACTION:

With the requirements of the above specification not satisfied, place load in a safe condition. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIRENENTS 4.9.13 The loaded weight of a spent fuel assembly cask shall be verified to not exceed 25 tons prior to attaching it to the spent fuel cask crane.

ST. LUG IE - UNIT 1 3/4 9-15

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided:

a. Reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s), and

b. All part length CEAs are at least 90K withdrawn and OPERABLE.

APPLICABILITY: MODE 2.

ACTION:

a ~ With the reactor critical (K f > 1.0) and with less than the above reactivity equivalent (vfilable for trip insertion or the part length CEAs not within their withdrawal limits, immediately initiate and continue boration at > 40 gpm of 1720 ppm boron or equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With the reactor subcritical (K < 1.0) by less than the above reactivity equivalent, im5Qiately initiate and continue boration at > 40 gpm of 1720 ppm boron or equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE RE UIREMENTS 4.10.1.1 The position of each full length and part length CEA either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 Each CEA not fully inserted shall be demonstrated OPERABLE by verifying its CEA drop time to be < 3.3 seconds within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

4.10.1.3 The part length CEAs shall be demonstrated OPERABLE by moving each part length CEA > 7.5 inches within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

ST.~ LUG IE - UNIT 1 3/4 10-1

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SPECIAL TEST EXCEPTIONS PRESSURE/TEMPERATURE LIMITATION - REACTOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.3 The minimum temperature and pressure conditions for reactor criticality of Specifications 3.1.1.5 and 3.4.9.1 may be suspended during low temperature PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5 percent of RATED THERMAL POWER,

b. The reactor trip setpoints on the OPERABLE power range neutron flux monitoring channels are set at < 20/ of RATED THERMAL POWER, and

c. The Reactor Coolant System temperature and pressure relationship is maintained within the acceptable region of operation shown on Figure 3.4-2.

APPLICABILITY: MODE 2.

ACTION:

a ~ With the THERMAL POWER > 5 percent of RATED THERMAL POWER, irenediately open the reactor trip breakers.

b. With the Reactor Coolant System temperature and pressure relationship within the region of unacceptable operation on Figure 3.4-2, immediately open the reactor trip breakers and restore the temperature-pressure relationship to within its limit within 30 minutes; perform the analysis required by Specification 3.4.9.1 prior to the next reactor criticality.

SURVEILLANCE RE UIREMENTS 4.10.3.1 The Reactor Coolant System shall be verified to be within the acceptable region for operation of Figure 3.4-2 at least once per hour.

4.10.3.2 The THERMAL POWER shall be determined to be < 5/ of RATED THERMAL POWER at least once per hour.

4.10.3.3 Each wide range logarithmic and power level channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating low temperature PHYSICS TESTS.

ST. LUCIE - UNIT 1 3/4 10-3

SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5/ of RATED THERMAL POWER, and

b. The reactor trip setpoints of the OPERABLE power level channels are set at < 20/ of RATED THERMAL POWER.

APPLICABILITY: During PHYSICS TESTS and Thermal-Hydraulic Tests.

ACTION:

With the THERMAL POWER ) 5% of RATED THERMAL POWER, immediately trip the reactor.

SURVEILLANCE RE UIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be < 5/ of RATED THERMAL POWER at least once per hour during PHYSICS TESTS:

4.10.4.2 Each wide range logarithmic and power level neutron flux monitoring channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

S . LUCIE - UNIT 1 3/4 10-4

SPECIAL TEST EXCEPTIONS CENTER CEA MISALIGNMENT LIMITING CONDITION FOR OPERATION 3.10.5 The requi'rements of Specifications 3.1.3.1 and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS to determine the isothermal temperature coefficient and power coefficient provided:

a. Only the center CEA (CEA Pl) is misaligned, and

b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.5.2 below.

APPLICABILITY: MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.3.1 and 3.1.3.6 are suspended, either:

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2. 1, or

b. Be in HOT STANDBY within 6 hours.

SURVEILLANCE RE UIREMENTS 4.10.5.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended and shall be verified to be within the test power plateau.

4.10.5.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5/ of RATED THERMAL POWER in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended.

ST. LUCIE - UNIT 1 3/4 10-5

BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

1 3/4. 0 APPLICABILITY BASES The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveil-lance Requirements within Section 3/4.

3.0.1 This specification states the applicability of each specifi-cation in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable.

3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement, 3.0.3 This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of the specification. For example, Specification 3.5.1 calls for each Reactor Coolant System safety injection tank to be OPERABLE and provides explicit ACTION requirements when one safety injection tank is inoperable. Under the terms of Specification 3.0.3', if more than one safety injection tank is inoperable, the facility is required to be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION statements.

The intent of this provisio'n is to insure that, facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded.

Exceptions to this provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications.

ST. LUG IE - UNIT 1 B 3/4 0-1

APPLICABILITY BASES 4.0.1 This specification provides that surveillance activities necessary to insure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL NODES or other conditions for which the Limiting Conditions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL NODES or other conditions are provided in the individual Surveillance Requirements.

4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations.

The tolerance values, taken either individually or consecutively over 3 test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not signifi-cantly degraded beyond that obtained from the nominal specified interval.

4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Conditions for Operation. Under this criteria, equipment, if

'imiting systems or components are assumed to be OPERABLE the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still meeting the Surveil-lance Requirements.

4.0.4 This specification ensures that the surveillance activities associated with a .Limiting Condition for Operation have been'performed within the specified time interval prior to entry into an OPERATIONAL NODE or other applicable condition. The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirement of the Limiting Condition for Operation.

Under the terms of this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance activities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into status. 'PERABLE ST. LUCIE - UNIT 1 B 3/4 0-2

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout cor'e life as a function of fuel depletion, RCS boron concentration, and RCS T . The most restrictive condition occurs at EOL, with T at no )Iced operating temperature, and is associated with a postu'flied steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 2.45K a,k/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN required by Specification 3.1.1.1 is based upon this limiting condition and is con-sistent with FSAR accident analysis assumptions. For earlier periods during the fuel cycle, this value is conservative. With T < 200'F, the reactivity transients resulting from any postulated ac83ent are minimal and a 1/ a,k/k shutdown margin provides adequate protection.

3/4.1.1.3 BORON DILUTION AND ADDITION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents strati fication and ensures that reactivity changes will be gradual during boron concentration changes in the Reactor Coolant System.

A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 11,400 cubic feet in approximately 26 minutes.

The reactivity change rate associated with boron concentration changes will be within the capability for operator recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT MTC The limiting values assumed for the MTC used in the accident and transient analyses were + 0.5 x 10 " ak/k/'F for THERMAL POWER levels

< 70% of RATED THERMAL POWER, + 0.2 x 10 4 ak k/'F for THERMAL POWER Tevels > 705 of RATED THERMAL and 2.5 x 10 ak/k/'F at RATED THERMAL POWER. Therefore, these limiting values are included in this specification.

Determination of MTC at the specified conditions ensures that the maximum positive and/or negative values of the MTC will not exceed the limiting values.

ST. LUG IE - UNIT 1 B 3/4 1-1

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1. 5 MINIMUM TEMPERATURE FOR CRITICALITY The MTC is expected to be slightly negative at operating conditions.

However, at the beginning of the fuel cycle, the MTC may be slightly positive at operating conditions and since it will become more positive at lower temperatures, this specification is provided to restrict reactor operation when T avg is significantly below the normal operating temperature.

3 4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility oper ation, The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200'F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The bor ation capability of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of. l.OX ak/k after xenon decay and cooldown to 200'F. The maximum boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 3,478 gallons of 8.0Ã boric acid solution from the boric acid tanks or 59,000 gallons of 1720 ppm borated water from the refueling water tank.

The requirements for a minimum=contained volume of 371,800 gallons of borated water in the refueling water tank ensures the capability for borating the RCS to the desired level. The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4. Therefore, the larger volume of borated water is specified here too.

With the RCS temperature below 200'F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restric-tions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

ST. LUCI E - UNIT 1 B 3/4 1-2

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORATION SYSTEMS Continued The boron capability required below 200'F is based upon providing a 1Ã hk/k SHUTDOWN MARGIN at 140'F during refueling with all full and part length control rods withdrawn. This condition requires either 5,650 gallons of 8.0/ boric acid solution from the boric acid tanks or 100,000 gallons of 1720 ppm borated water from the refueling water tank.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met.

The ACTION statements applicable to an immovable or untrippable CEA and to a large misalignment (> 15 inches) of two or more CEAs, require a prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrippable CEA, the loss of SHUTDOWN MARGIN.

For small misalignments (( 15 inches) of the CEAs, there is 1) a small degradation in the peaking factors relative to those assumed in generating LCOs and LSSS setpoints for DNBR and, linear heat rate, 2) a small effect on the time dependent long term power distributions relative to those used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a small effect on the available SHUTDOWN MARGIN, and 4) a small effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated, with the small misalignment of a CEA permits a one hour time interval during which attempts may be made to restore the CEA to within its alignment requirements prior to initiating a reduction in THERMAL POWER. The one hour time limit is sufficient to (1) identify causes of a misaligned CEA, (2) take appropriate corrective action to realign the CEAs and (3) minimize the effects of xenon redistribution.

Overpower margin is provided to protect the core in the event of a large misalignment (> 15 inches) of a CEA. However, this misalignment would cause distortion of the core power distribution. The reactor ST. LUCIE - UNIT 1 B 3/4 1-3

REACTIVITIY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Continued I

protective system would not detect the degradation in radial peaking factors and since variations in other system parameters (e.g., pressure and coolant temperature) may not be sufficient to cause trips, it is possible that the reactor could be operating with process variables less conservative than those assumed in generating LCO and LSSS setpoints.

Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt and significant reductiog in THERMAL POWER prior to attempting realignment of the misaligned CEA.

The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA. Conformance with these alignment requirements bring the core, within a short period of time, to a configuration consistent with that assumed in generating LCO and LSSS setpoints. However, extended operation with CEAs significantly inserted in the core may lead to perturbations in 1) local burnup, 2) peaking factors and 3) available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

Operability of the CEA position indicators (Specification 3.1.3.3) is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit. The CEA "Full In" and "Full Out" limits provide an additional independent means for determining the CEA positions. when the CEAs are at either their fully inserted or fully withdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be, verified by the "Full In" or "Full Out" limits.

CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

The maximum CEA drop time permitted by Specification 3.1.3.4 is the assumed CEA drop time of 3.3 seconds used in the accident analyses.

Measurement with Tav > 515'F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

ST. LUCIE - UNIT 1 B 3/41-4

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Continued The LSSS setpoints and the power distribution LCOs were generated based upon a core burnup which would be achieved with the core operating in an essentially unrodded configuration. Therefore, the CEA insertion limit specifications require that during MODES 1 and 2, the full length CEAs and par t length CEAs be nearly fully withdrawn. The amount of CEA .

insertion permitted by the Long Term Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the un-rodded burnup assumption but will still provide sufficient reactivity control. The Power Dependent Insertion Limits of Specification 3.1.3.6 are provided to ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels; however, long term operation at, these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configuration.

ST. LUCIE - UNIT 1 B 3/4 1-5

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POWER DISTRIBUTION LIMITS BASES TOTAL RADIAL PEAKING FACTOR - FAND AZIMUTHAL POWER TILT - T (continued)

ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the DNB Margin, Linear Heat Rate, Thermal Margin/Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT ) 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The value of T that must be used in the equat'ion F r

= F r (1 + T )

is the measured till. q T

The surveillance requirements for verifying that Fr and T , are within their limits provide assurance that the ac)ual values ot FTr and Tq do not exceed the assumed values. Verifying Fr after each fuel I

loading prior to exceeding 75K of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

3/4.2.4 FUEL RESIDENCE TIME The limitation on fuel burnup during the initial fuel cycle-ensures that fuel cladding, collapse will not occur. Performance data from sirplar fuel rods and analyses of the installed fuel rods show that cladding g~

collapse will not occur until well beyond the proposed first cycle of operation which is about 11,200-12,000 Effective Full Power Hours. How-ever, operation beyond the first cycle will require further analyses.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.

ST. LUCIE - UNIT 1 B 3/4 2-2

POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic survei llance of these parameters through instrument readout is sufficient to ensure that hte parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

ST. LUCIE - UNIT 1 B 3/4 2-3

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INSTRUMENTATION BASES RADIATION MONITORING INSTRUMENTATION Continued by the individual channels and 2) an alarm is initiated when the radiation level alarm setpoint is exceeded.

3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3/4.3.3. 3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that suffi-cient capbility is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility.

3/4.3.3.4. METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs",

February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

ST. LUG IE - UNIT 1 B 3/4 3-2

INSTRUMENTATION 3/4.3.3.6 CHLORINE DETECTION SYSTEMS The operabili.ty of the chlorine detection systems ensures that an accidental chlorine release will be detected promptly and the necessary protective actions will be automatically initiated to provide protection for control room personnel. Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically isolate the control coom and initiate its operation in the recirculation mode of operation to provide the required protection. The chlorine detection systems required by this specification are consistent with the recommendations of Regulatory Guide 1.95, "Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release",

February 1975.

e ST. LUCIE - UNIT 1 B 3/4 3-3

3 4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. STARTUP and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for the Power Level-High, Reactor Coolant Flow-Low, and Thermal Margin/Low Pressure trips have been reduced to their specified values. Reducing these trip setpoints ensures that the DNBR will be maintained above 1.30 during three pump operation and that during two pump operation the core void fraction will be limited to ensure parallel channel flow stability within the core and thereby prevent premature DNB.

A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure consi-derations require plant cooldown if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

3 4.4.2 and 3 4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 2 x 10 lbs per hour of saturated steam at. the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capa-bility and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia. The combined relief capacity of these valves is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMA POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.

ST. LUCIE UNIT 1 B 3/4 4-1

REACTOR COOLANT SYSTEM BASES SAFETY VALVES Continued Demonstration of the safety"valves'ift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure 'Vessel Code, 1974 Edition.

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressuri zer code safety valves and power operated relief valve against water relief.

The power'perated relief valve and steam bubble function to relieve RCS pressure during all 'design transients. Operation of the power operated relief valve in conjunction with a reactor"trip on a Pressurizer--

Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

3/4.4.5 STEAM GENERATORS One OPERABLE steam generator provides sufficient heat removal capa-bility to remove decay heat 'after a reactor shutdown. The requirement ;

for two steam generator s capable of removing decay heat, combined with the requirements of Specifications 3.7.1.1, .3.7.1.2 and 3.7.1.3 ensures adequate decay heat removal capabilities for RCS temperatures greater than 300'F if one steam generator becomes inoperable due to single failure considerations. Below 300'F, decay heat is removed, by

~

the shutdown cooling system.'The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity 'of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of .,Regulatory Guide 1.83, Revision l. Inservice inspection of steam generator tubing is essential in order to maintain survei llance of the conditions of the tubes in the event that there is evidence of mechanical damage'r progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. 'Inservice in'spection of steam general'or tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

ST. LUCIE - UNIT 1 B 3/4 4-2

REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAN GENERATORS Continued The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =

1 gallon per minute, total). Cracks having a primary-to-secondary leakage less than this limit, during operation wi 11 have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 1 gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40Ã of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20K of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9. 1 prior to resumption of plant operation. Such cases will-be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

ST. LUCIE - UNIT 1 B'3/4 4-3

REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.

3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of. an impending gross .failure of the pressure boundary.

Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associat-ed effects of exceeding the oxygen, chloride and fluoride limits are ST. LUCIE - UNIT 1 B 3/4 4-4

REACTOR COOLANT SYSTEM BASES time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that con-centrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite electrical power. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the St. Lucie site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/gram DOSE E(UIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL'OWER. Operation with specific activity levels exceeding 1.0 pCi/gram DOSE EQUIVALENT no I131 but within the limits shown on Figure 3.4-1 must be restricted to more than 10 percent of the unit's yearly operating time since the activity levels allowed by Figure 3.4-1 increase the 2 hour thyroid dose at the site boundary by a factor*of up to 20 following a postulated steam generator tube rupture.

ST. LUCIE - UNIT 1 B 3/4 4-5

REACTOR COOLANT SYSTEM BASES Reducing Tavg to < 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.'he surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible justified by the data obtained.

if 3/4.4. 9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are, designed to with-stand the effects of cyclic loads due to system temperature and pressure These cyclic loads are introduced by normal load transients, .'hanges.

reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 5.2.1 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown'ates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall ~

produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive, stresses to alleviate the tensile stresses induced by the internal pressure. 'end Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the becomes the controlling location. The thermal gradients, estab-

'essel lished during heatup produce tensile stresses at the outer. wall of the vessel. These stresses are additive to the pressure induced tensile stresses. which are already present. The thermal induced stresses at the

'outer wall of the vessel are tensile and are dependent on 'both the rate of heatup,and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

ST. LUG I E - UNIT 1 B 3/4 4-6

REACTOR COOLANT SYSTEM BASES The heatup and cooldown limit curves (Figure 3.4-2) are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup or cooldown rates of up to 100'F per hour. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figure 2.4-2.

The reactor vessel materials have been tested to determine their initial RTNpT; the results of these test are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E)1 Mev) irradiation will cause an increase in the RTNgy. Therefore, an adjusted reference temperature, based upon the fluence, can be predicted using Figure B 3/4.4-1. The heatup and cooldown limit curves shown on Figure 3.4-2 include predicted adjustments for this shift in RTNDT at the end of the applicable service period, as well as adjustments for possible errors in the pressure and temperature sensing instruments.

The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance .

with ASTM E185-73, reactor vessel material irradiation surveillance 0 specimens installed near the inside wall of the reactor vessel 'in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition

=

shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalcu-lated when the aRTNDT determined from the surveillarice capsule is dif-ferent from the calculated LRTNOT for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and-for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix 6 to 10 CFR 50.

The maximum RTNDT for all reactor coolant, system pressure-retaining materials, with. the exception of the reactor pressure vessel, has been determined to be 50'F. The Lowest Service Temperature limit line shown on Figure 3.4-2 is based upon this RTNDT since Article NB-2332 (Summer Addenda of 1972) of Section III:of the ASME Boiler and Pressure Vessel Code requires the Lowest Service .Temperature to.be RTNDT + 100'F ST. LUCIE - UNIT 1 B 3/4 4-7

TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS 50 FT-LB/35* MIN. UPPER SHELF COMP MATERIAL CU P NDTT MIL TEMP F RTNDT FT LB **

COMPONENT CODE TYPE 5 5 F LONG TRANS F LONG TRANS Vessel Flange C-1-1 A508C1.2 - .008 +20 +30 128 Forging Bottom Head Plate C-10-1 A533BC1.1 .010 -40 +18 118 Bottom Head Plate C-9-2 A533BC1.1 .011 -40 -26 145 Bottom Head Plate C-9-3 A533BC1.1 .013 -70 -34 148 Bottom Head Plate C-9-1 A533BCl.l .011 -30 +2 135 Inlet Nozzle C-4-3 A508C1.2 .005 0 -23 109 Inlet Nozzle C-4-2 A508C1.2 .004 0 -5 140 Inlet Nozzle C-4-1 A508C1.2 .005 +10 -30 141 Inlet Nozzle C-4-4 A508C1.2 .004 0 -50 132 Inlet Nozzle Ext. C-16-3 A508C1.2 .001 +10 0 135 Inlet Nozzle Ext. C-16-2 A508C1.2 .011 +10 0 135 Inlet Nozzle Ext. C-16-1 A508C1.2 .011 +10 0 135 Inlet Nozzle Ext. C-16-4 A508C1.2 .Oll +10 0 135

Average Value from Curve

- Minimum Value at 100K Shear

TABLE B 3/4.4-1 Cont'd REACTOR VESSEL TOUGHNESS 50 FT-LB/35* MIN. UPPER SHELF COMP MATERIAL CU P NDTT MIL TEMP F RTNDT FT- LB **

COMPONENT CODE TYPE F LO~NG N. F LONG TRANS Outlet Nozzle C-3-1 A508C1.2 - .009 +10 +50 118 Outlet Nozzle C-3-2 A508C1.2 .010 -20 +60 108 Outlet Nozzle Ext. C-17-1 A508C1.2 .013 +20 +27 126 Outlet Nozzle Ext. C-17-2 A508C1.2 - .013 +20 +27 126 Upper Shell Plate C-6-3 A533BC1.1 - .011 -10 24 117 Upper Shell Plate C-6-2 A533BC1.1 - .010 -30 113 Upper Shell Plate C-6-1 A533BC1.1 - .012 +10 34 104 Inter. Shell Plate C-7-1 A533BC1.1 0.11 0. 004 0 +10 126 Inter. Shell Plate C-7-2 A533BC1.1 0.11 0.004 -30 +10 126 Inter. Shell Plate C-7-3 A533BC1. 1 .0. 11 0. 004 -30 +30 124 Lower Shell Plate C-8-3 A533BC1.1 0.12 0.004 0 +13 135 Lower Shell Plate C-8-1 A533BC1 .1 0.15 0. 006 -10 +16 126 Lower Shell Plate C-8-2 A533BC1.1 0.15 0.006 0 +16 122 Closure Head Flange C-2 A508C1 . 2 - . 008 +10 140 Closure Head Peels C-21-2 A533BC1.2 - .012 -30 +26 132 Closure Head Peels C-21-2 A533BC1.1 - .012 -30 +26 132

-=-TABLE B 3/4.4-1 (Cont'd REACTOR VESSEL TOUGHNESS 50 FT-LB/35* MIN. UPPER SHELF COMP MATERIAL CU P NDTT MIL TEMP F RTNDT FT-LB **

COMPONENT CODE TYPE F LONG TRANS F LONG TRANS Cl.osure He'ad Peels C-21-1 A533BC1.1 .013 -10 -10 125 Closure Head Peels C-21-1 A533BCl.l .013 -10 -10 125 Closure Heal Peels C-21-2 A533BC1.1 .012 -30 +26 132

, Closure Head Peels C-21'-3 A533BC1.1 .013 -40 +16 121 Closure Head Dome C-20-1 A533BC1.1 .014 -10 +37 101

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O ct z 100 Ijjt I'I 1018 10te FLUENCE,n/cm >1Mev FIGURE B 3/4.4.1 Nil-DuctilityTemperature increase as a Function of Fast Neutron Fluence (E> 1mev)

REACTOR COOLANT'SYSTEM BASES for piping, pumps and valves. Below this temperature, the system pressure must be limited to a maximum of 201. of the system's hydrostatic test pressure of 3125 psia.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removi,ng and testing these specimens are provided in Table 4. 4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fati-gue analysis performed in accordance with the ASME C'ode requirements.

3/4.4.10 STRUCTURAL INTEGRITY The required inspection programs for the Reactor Coolant System components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

To the extent applicable, the inspection program for the Reactor Coolant System components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code "Inservice Inspection of Nuclear Reactor Coolant Systems", 1971 Edition, and Addenda through Winter 1972.

All areas scheduled for volumetric examination have been pre-service mapped using equipment, techniques and procedures anticipated for use during post-operation examinations. To assure that consideration is given to the use of new or improved inspection equipment, techniques and procedures, the Inservice Inspection Program will be periodically reviewed on a 5 year basis.

The use of conventional nondestructive, direct visual and remote visual test techniques can be applied to the inspection of most reactor coolant loop components except the reactor vessel. The reactor vessel requires special consideration because of the radiation levels and the requirement for remote underwater accessibility.

The techniques anticipated for inservice inspection include visual inspections, ultrasonic, radiographic, magnetic particle and dye pene-trant testing of selected parts.

ST. LUCIE - UNIT 1 B 3/4 4-12

REACTOR COOLANT SYSTEM BASES The nondestructive testing for repairs on components greater than 2 inches diameter gives a high degree of confidence in the integrity of the system, and will detect any significant defects in and near the new welds. Repairs on components 2 inches in diameter or smaller receive a surface examination which assures a similar standard of integrity. In each case, the leak test will'ensure leak tightness during normal operation.

For normal opening and reclosing, the structural integrity of the Reactor Coolant System is unchanged. Therefore, satisfactory performance of a system leak test at 2235 psia following each opening and subsequent reclosing is acceptable demonstration of the system's structural inte-rity. These leak tests will be conducted within the, pressure-temperature limitations for Inservice Leak and Hydrostatic Testing and Figure 3.4-2.

The Safety Class 2 and 3 components will be pressure tested at least once toward the end of each inspection interval (10 years). The Safety Class 2 components having a design temperature above 400'.F will be pressure tested at not less than 125 percent of the system design pressure while those components having a design temperature of 400'F and below will be pressure tested at 110 percent of design pressure. The Safety Class 3 components will be pressure tested at the levels indicated in Specification 4.4.10.3b.

3/4.4.11 CORE BARREL MOVEMENT This specification 'is provided to ensure early detection of excessive core barrel movement if it should occur. Core barrel movement will be detected by using four excore neutron detectors to obtain Amplitude Probability Distribution (APD) and Special Analysis (SA). Baseline core barrel movement Alert Levels and Action Levels at nominal THERMAL POWER levels of 20%, 50%, 80% and 100% of RATED THERMAL PO>lER will be determined during the reactor startup test program.

A modification to the required monitoring program may be justified by an analysis of the .data obtained and by an examination of the affected parts during the plant shutdown at the end of the first fuel cycle.

T. LUCIE - UNIT 1 B 3/4 4-13

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EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.4 REFUELING WATER TANK RWT The OPERABILITY of the RWT as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWT minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of the RWT and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assump-tions are consistent with the LOCA analyses.

ST. LUCI E - UNIT 1 B 3/4 5-2

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CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structural is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.70 psi and 2) the containment peak pressure does not exceed the design 'pressure of 44 psig during steam line break accident conditions.

'he maximum peak pressure obtained from a steam line break accident is 41.6 psig. The limit of 2.4 psig for initial positive containment will limit the total pressure to 44.0 psig which is the design 'ressure pressure and is consistent with the accident analyses.

3/4.6.1.5 AIR TEMPERATURE

'he limitation on containment air temperature ensures that the containment vessel temperature does not exceed the design temperature of 264'F during LOCA conditions. The containment temperature limit is consistent with the accident analyses.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the con-tainment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pr essure of 41.6 psig in the event of a steam -line break accident. A visual inspection in conjunction with Type A leakage test is sufficient to demonstrate this capability.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that con-tainment depressurization and cooling capbility will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.

ST. LUG IE - UNIT 1 B 3/4 6-2

CONTAINMENT SYSTEMS BASES 3/4.6 2. 2

~ CONTAINMENT COOLING SYSTEM The OPERABILITY of the containment c'ooling system ensures that 1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when operated in conjunction with the containment spray systems during post-LOCA condi,tions.

3/4.6.3 CONTAINMENT ISOL'ATION 'VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmos-phere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit, during post-LOCA conditions. Either recombiner unit is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions, 2) radiolytic decomposition of water and 3) corrosion of metals within containment.

The containment fan coolers are used in a secondary function to ensure adequate mixing of the containment atmosphere following a LOCA.

This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

3/4.6.5 VACUUM REL'IEF VALVES The OPERABILITY of the containment vessel to annulus vacuum relief valves ensures that they will open at a pressure differential of 2.25 + 0.25 inches Hater Gauge. This condition is ne'cessary to prevent exceeding the containment design limit for internal pressure differential of 0.70 psi.

ST. LUCIE - UNIT 1 8 3/4 6-3

CONTAINMENT SYSTEMS BASES 3/4.6.6 SECONDARY CONTAINMENT 3/4.6.6.1 SHIELD BUILDING VENTILATION SYSTEM The OPERABILITY of the shield building ventilation systems ensures that containment vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions used in the accident analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions.

3/4.6.6.2 SHIELD BUILDING INTEGRITY SHIELD BUILDING INTEGRITY ensures that the release of radioactive materials from the primary containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with operation of the shield building ventilation system, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

3/4.6.6.3 SHIELD BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the con-tainment shield building will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to provide 1) protection for the steel vessel from external missiles, 2) radiation shielding in the event of a LOCA, and 3) an annulus surrounding the steel vessel that can be maintained at a negative pressure within two minutes after a LOCA.

ST. LUCIE - UNIT 1 B 3/4 6-4

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PLANT SYSTEMS BASES 106.5 Power Level-High Trip Setpoint for,two loop operation Total, relieving capacity of all safety valves per steam line in lbs/hour (5.95 x 10'bs/hr.)

Maximum relieving capacity of any one safety valve in lbs/hour (7.44 x 10s lbs/hr.)

3/4.7.1.2 AUXILIARY FEEDWATER PUMPS The OPERABILITY of the auxiliary feedwater pumps ensures that the Reactor Coolant System can be cooled down to less than 300'F from normal operating conditions in the event of a total loss of off-site power.

Any two of the three auxiliary feedwater pumps have the required capacity to provide sufficient feedwater flow to remove reactor decay heat and r educe the RCS temperature to 300'F where the shutdown cooling system may be placed into operation for continued cooldown.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 300'F in the event of a total loss of off-site power. The minimum water volume is sufficient to maintain the RCS at HOT STANDBY conditions for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with steam discharge to atmosphere.

3/4.7. 1 .4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction "

of 10 CFR Part 100 limits in the event of a steam line rupture. The dose calculations for an assumed steam line rupture include the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the accident analyses.

ST. LUCI E - UNIT 1 8 3/4 7-2

PLANT SYSTEMS BASES 3/4.7.1.5 HAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and') limit the pressure rise within containment in the event the steam line rupture occurs within contain-ment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 SECONDARY WATER CHEMISTRY A test program will be conducted during approximately the fir st 6 months of operation after initial criticality to establish the appropriate limits on the secondary water chemistry parameters and to determine the appropriate frequencies for monitoring these parameters. The results of this test program will be submitted to the Commission for review. The Commission will then issue a revision to this specification specifying the limits on the chemistry parameters and the frequencies for monitoring these parameters.

The test program will include an analysis of the chemical con-stitutent of the makeup water for the St. Lucie Plant. The analysis shall identify the various traces of ions which upon concentration may have the potential for inducement, for stress corrosion in the steam generator tubing. The test program shall also evaluate the efficiency of the water treatment systems in the St. Lucie facility for removal of such ions. and the potential for addition of other ions resulting from the treatment method. The test program shall analyze concentration phenomena and'he concentration rates in the steam generator -and the secondary water system and shall consider concentration in the recirculat-ing cooling water system.

3 4.7.2 STEAM GENERATOR PRESSURE TEMPERATURE LIMITATION 4

The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not .exceed the maximum allowable fracture toughness stress limits. The limitations of 70'F and 200-psig are based on a steam generator RTNDT of 50'F and are sufficient to prevent brittle fracture.

ST. LUCIE - UNIT 1 B 3/4 7-3

PLANT SYSTEMS BASES 3/4.7.3 COMPONENT COOL'ING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

'3/4.7.4 INTAKE'COOLING WATER SYSTEM The OPERABILITY of the intake cooling water system ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4;7.5 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level and temperature ensure that sufficient cooling capacity is available to either 1) pro-vide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits.

The limitations on minimum water level and maximum temperature are based on providing an adequate cooling water supply to safety related equipment until cooling water can be supplied from Big Mud Creek.

3/4.7.6 FLOOD PROTECTION The limitation on flood protection ensures that facility will be adequately protected from flooding.

3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumenta-tion cooled by this system and 2) the control room will remain habitable ST. LUCIE - UNIT 1 B 3/4 7-4

PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM Continued for operations personnel during and following all credible accident con-ditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting,the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent, and maintaining the chlorine concentration within acceptable limits during and following a chlorine accident. This limitation is-consistent with the requirements of General Design Criteria 10 of Appendix "A", 10 CFR 50.

3/4.'7.8 ECCS AREA VENTILATION SYSTEM The OPERABILITY of the ECCS area ventilation system ensures that radioactive materials leaking from the 'ECCS equipment following a LOCA are filtered prior to reaching the environment. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the accident analyses.

'3/4.7. 9 SEALED SOURCE 'CONTAMINATION The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the probable leakage from the source material. The limitations on removable contamina-tion for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. guantities of interest to this specification which are exempt from the leakage testing are consistent with the criteria of 10 CFR Parts 30.11-20 and 70.19. Leakage from sources excluded from the requirements of this specification is not likely to represent more than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.

3/4.7.10 HYDRAULIC 'NUBBERS The hydraulic snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. The only snubbers excluded from this inspection'rogram are those installed on nonsafety related systems and then only if their fai lure or failure of the system on which they are installed, would have no adverse effect on any safety related system.

ST. LUG I E - UNIT 1 B 3/4 7-5

PLANT SYSTENS BASES The inspection frequency applicable to snubbers containing seals fabricated from materials which have been demonstrated compatible with their operating environment (only ethylene propylene compounds to date) is based upon maintaining a constant level of snubber protection.

Therefore, the required inspection interval varies inversely with the observed snubber failures. The number of inoperable snubbers found during an inspection of these snubbers determines the time interval for the next required inspection of these snubbers. Inspections performed before that interval has elapsed may be used as a 'new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 251) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

To provide further assurance of snubber reliability, a representa-tive sample of the installed snubbers will be functionally tested during plant shutdowns at 18 month intervals. These tests wi,ll include stroking of the snubbers to verify proper piston movement, lock-up and bleed. Observed failures of these sample snubbers will require functional testing of additional units. To minimize personnel exposures, snubbers installed in high radiation zones or in especially difficult to remove locations (as identified in Table 3.7-2) may be exempted from these functional testing requirements provided the OPERABILITY of these snubbers was demonstrated during functional testing at either the completion of their fabrication or at a subsequent date.

ST. LUCIE - UNIT 1 B 3/4 7-6

3/4.8 ELECTRICAL POWER SYSTEMS BASES The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility. The minimum specified indepen-dent and redundant A.C. and D.C; power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A" to 10 CFR 50.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the accident analyses and are based upon maintaining at least one of each of the onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source.

The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that 1) the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control capability is available for monitoring and maintaining the facility status.-

ST. LUCIE - UNIT 1 B 3/4 8-1

3 4.9 REFUEL'ING OPERATIONS.

BASES 3/4.9:1 BORON CONCENTRATION The limitations on minimum boron concentration (1720 ppm) ensure that: 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volumes having direct access to the reactor vessel. The limitation on K fq of no greater than 0.95 is sufficient to prevent reactor criticality with all full length rods (shutdown and regulating) fully withdrawn.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the, wide range logarithmic range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to move-ment of irradiated fuel assemblies in 'the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3/4.9. 4 CONTAINMENT'ENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressur-ization potential while in the REFUELING MODE.

3/4.9. 5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

Sf. LUCIE UNIT 1 B 3/4 9-1

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REFUEL'I NG OPERATIONS BASES 3/4.9.12 FUEL POOL VENTILATION SYSTEM-FUEL STORAGE The limitations on the fuel handling building ventilation system ensures that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

3/4.9.13 SPENT FUEL CASK CRANE The maximum load which may be handled by the spent fuel cask crane is limited to a loaded single element cask which is equivalent to approximately 25 tons. This restriction is provided to ensure .the structural integrity of the spent fuel pool in the event of a dropped cask accident. Structural damage caused by dropping a load in excess of a loaded single element cask could cause leakage from the spent fuel pool in excess of the maximum makeup capability.

ST. LUG I E - UNIT 1 B 3/4 9-3

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SECTlON 5.0 PESrGN FEATURES

5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area is shown on Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone is shown on Figure 5.1-2.

FLOOD CONTROL 5.1.3 The flood control provisions (dunes and slope protection) shall be designed and maintained in accordance with the original design provisions contained in Section 2.4.2.2 of the FSAR.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The containment structure is comprised of a steel containment vessel, having the shape of a right circular cylinder with a hemispheri-cal dome and ellipisoidal bottom, surrounded by a reinforced concrete shield building. The radius of the shield building is at least 4 feet greater than the radius of circular cylinder portion of the containment vessel at any point.

5.2.1.1 CONTAINMENT VESSEL a ~ Nominal inside diameter = 140 feet.

b. Nominal inside height = 232 feet.

c~ Net free volume = 2.5 x 10 cubic feet.

d. Nominal thickness of vessel walls = 2 inches.

e. Nominal thickness of vessel dome = 1 inch.

f. Nominal thickness of vessel bottom = 2 inches.

ST. LUCIE - UNIT 1 5-1

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DESIGN FEATURES 5.2.1.2 SHIELD BUILDING a ~ Minimum annular space = 4 feet.

b. Annulus nominal volume = 543,000 cubic feet.

c ~ Nominal outside height (measured from top of foundation base to the top of the dome) = 230.5 feet.

d. Nominal inside diameter = 148 feet.

e. Cylinder wall minimum thickness = 3 feet.

f. Dome minimum thickness = 2.5 feet.

g. Dome inside radius = 112 feet.

DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment vessel is designed and shall be maintained for a maximum internal pressure of 44 psig and a temperature of 264'F.

PENETRATIONS 5.2.3 Penetrations through the containment structure are designed and shall be maintained in accordance with the original design pro-visions contained in Sections 3.8.2.1.10 and 6.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 176 fuel rods clad with Zircoloy-4. Each fuel rod shall have a nominal active fuel length of 136.7 inches and contain a maximum total weight of 2250 grams uranium. The initial core loading shall have a maximum enrichment of 2.83 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have maximum enrichment of 3.1 weight percent U-235.

ST. LUG I E UNIT 1 5-4

DESIGN FEATURES CONTROL ELEMENT ASSEMBL'IES 5.3.2 The reactor core shall contain 73 full length and 8 part length control element assemblie's. The control element assemblies shall be designed and maintained in accordance with the original design provisions contained in Section 4.2.3.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant sys'em is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650'F, except for the pressurizer which is 700'F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 11,100 + 180 cubic feet at a nominal T of 567'F.

5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pur-suant to the applicable Surveillance Requirements.

5.6 FUEL STORAGE CRITICALITY 5.6.1 The new and spent fuel storage racks are designed and shall be maintained with a nominal center-to-center distance of 21 inches for new fuel assemblies and 18 inches for spent fuel assemblies placed in the ST. LUCIE - UNIT 1 5-5

DESIGN FEATURES CRITICALITY Continued storage racks to ensure a K equivalent to < 0.95 with the storage pool filled with unborated Ner. The K of < 0.95 includes the conservative assumptions as described in Qction 9.1 of the FSAR.

DRAINAGE 5.6.2 The fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 56 feet.

CAPACITY 5.6.3 The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 304 fuel assemblies of which the 45 fuel assemblies in the 5 x 5 array and 5 x 4 array nearest the fuel cask compartment shall have decayed for at least 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months .

5. 7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as seismic Class I in Section 3.2.1 of the FSAR shall be designed and maintained to the original design provisions contained in Section 3.7 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.8 METEOROLOGICAL TOWER LOCATION 5.8.1 The meteorological tower location shall be as shown on Figure 5.1-1.

5.9 COMPONENT. CYCLE OR TRANSIENT LIMITS 5.9.1 The components identified in Table 5.9-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.9-1.

ST. LUCIE - UNIT 1 5-6

TABLE 5.9-1 COMPONENT CYCLIC OR TRANSIENT LIMITS COMPONENT CYCLIC OR TRANSIENT LIMITS DESIGN CYCLE OR TRANSIENT Reactor Coolant System 40 Cycles of loss of load without 100'X to 0/. RATED THERMAL immediate reactor trip POWER 40 cycles of loss of offsite 100K to OX RATED THERMAL A.C. electrical power POWER 400 reactor trips lOOX to OX RATED THERMAL POWER 16 inadvertent auxiliary Spray line 650'F to 120'F spray cycles in 1.5 seconds 200 leak tests Pressure > 2235 psig 10 hydrostatic pressure tests Pressure > 3110 psig Secondary System 5 steam line breaks Complete loss of secondary pressure 200 leak tests Pressure > 985 psig 10 hydrostatic pressure tests Pressure > 1235 psig

SECTION 6.0 ADMINISTRATIVE CONTROLS

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this respon-sibility during his absence.

6. 2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Operator shall be in the control room when fuel is in the reactor.

c. At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.

d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

e. All CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

ST. LUCIE - UNIT 1 6-1

EXECUTIVE VICE PRESIDENT GROUP VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT COMPANY NUCLEAR NUCLEAR AFFAIRS ENVIRONMENTAL PLANNING 5 TIESEARCH REVIEW BOARD VICE PRESIDENT POWER RESOURCES MANAGER MANAGER OF MANAGER OF MANAGER OF MANAGER OF NRC LICENSING DUALITYASSURANCE NUCLEAR ANALYSIS POWER RESOURCES POWER RESOURCES NUCLEAR SERVICES POWER RESOURCES PLANT'ANAGER SUPERVISOR

'INCLUOES HEALTH PHYQCS, FUEL NUCLEAR MANACEMENT,OPERATIONS FOLLOW, OEQGN REVIEW. REACTOR ENGIN.

SEE FIGURE S.E.E EERING, RADIOCHEMISTRY, ETC.

'POWER RESOURCES STAFF. NUCLEAR I

I CONTINUEO ON FIGURE CE.E Figure 6.2-1 Offsite Organization for Facility Management and Technical Support

PLANT MANAGER POWER RESOURCES STAFF HEALTH PHYSICS SPECIALIST I

I MAINTENANCE TECHNICAL ADMINISTRATIVE UALITY CONTRO OPERATIONS SUPERINTENDENT STAFF ASSISTAN'T STAFF SUPERINTENDENT INSTRUMENT ASSISTANT ASSISTANT SUPERINTENDENT SUPERINTENDENT SECURITY CHEMISTRY ,HEALTH SRO TRAINING AND REACTOR SUPERVISOR SUPERVISOR PH YSICS OPERAT/ONS CONTROL ELECTRICAL MECHNIGAL SUPERVISOR SUPERVISOR SUPERVISOR SUPERVISOR SUPERVISOR MAINTENANCE MAINTENANCE SRO ENG/NEERS ENGINEERS PLANT ENGINEERS ENGINEERS ENGINEERS ENGINEERS'ECHNICIAN" GUARDS CLERKS AND ANO ANO TECHNICIANS TECHNICIANS SUPERVISOR TECHNICANS TECHNICIANS TECHN/CIANS I/$ 5 CHIEF MAINTENANCE WATCH ELECTRICIAN FOREMAN ENGINEER I/$ 5 INSTRUMENT RO ANO ELECTRICIANS MACHINIST CONTROL CENTER CONTROL APPRENTICES MECHANICS OPERATOR spEc/AUGTs HELPERS APPRENTICES APPRENTICES HELPER I/S 5 RO POSITION TITLE I /$ 5 UNLICENSED OPERATORS SRO AEC SENIOR OPERATOR LICENSE RO AKC REACTOR OPERATOR LICENSE IO NUMSKR RKOUIREO PKR SHIFT 5 TOTAL NI/MSER IN CLASSIFICATION Figure 6.2-2 Facility Organization St. Lucie Plant, Unit 1

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION LICENSE APPLICABLE MODES CATEGORY 1, 2, 3, 5 4 5 5 6 SOL OL Non-Licensed

Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS after the initial fuel loading.

Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accomodate un-expected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

ST. LUCI E - UNIT 6-4

ADMINI STRATI VE CONTROLS 6.3 FACILITY STAFF UAL'IFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum, qualifications of ANSI N18.1-1971,for comparable positions, except for the Radiation Protection Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction, of the Training Supervisor and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.5 REVIEW AND AUDIT 6.5.1 FACILITY REVIEW GROUP FRG FUNCTION 6.5.1.1 The Facility Review Group shall function to advise the Plant Manager on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The Facility Review Group shall be composed of the:

Member: Plant Manager Member: Operations Superintendent Member: Operations Supervisor Member: Maintenance Superintendent Member: Instrument & Control Supervisor Member: Reactor Supervisor Member: Health Physics Supervisor Member: Technical Supervisor Member: Chemistry Supervisor Member: equality Control Supervisor Member: Assistant Plant Supt. Mechanical Member: Assistant Plant Supt. Electrical The FRG Chairman shall be designated in writing.

6-5

ADMINISTRATIVE CONTROLS ALTERNATES 6.5.1.3 All alter nate members shall be appointed in writing by the FRG Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in FRG activities at any one time.

MEETING FRE UENCY 6.5.1.4 The FRG shall meet at least once per calendar month and as convened by the FRG Chairman.

QUORUM 6.5.1.5 A quorum of the FRG shall consist of the Chairman and four members including alternates.

RESPONSIBILITIES 6.5.1.6 The Facility Review Group shall be responsible for:

a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

C. Review of all proposed changes to Appendix "A" Technical Specifications.

d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

ST. LUCIE - UNIT 1 6-6

ADMINISTRATIVE CONTROL'S

e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Manager of Power Resources Nuclear, the Vice President of Power Resources and to the Chairman of the Company Nuclear Review Board.

Review of those REPORTABLE OCCURRENCES requires 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notifica-tion to the Commission.

9, Review of facility operations to detect potential safety hazards.

Performance of special reviews and investigations and reports thereon as requested by the Chairman of the Company Nuclear Review Board.

Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the Company Nuclear Review Board.

Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the Company Nuclear Review Board.

0 AUTHORITY 6.5.1.7 The Facility Review Group shall:

a. Recommend to the Plant Manager written approval or disapproval of items considered under 6.5.1.6(a) through (d) above.

b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

c~ Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President of Power Resources and the Company Nuclear Review Board of disagreement between the FRG and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

ST. LUCIE - UNIT 1 6-7

ADMINISTRATIVE CONTROLS RECORDS 6.5.1.8 The Facility Review Group shall maintain written minutes of each meeting and copies shall- be provided to the Vice President of Power Resources and Chairman of the Company Nuclear Review Board.

6.5.2 COMPANY NUCLEAR REVIEW BOARD CNRB FUNCTION 6.5.2. 1 The Company Nuclear Review Board shall function to provide independent review and audit of designated activities in the areas of:

a. nuclear power plant operations

b. nuclear engineering

c. chemistry and radiochemistry

d. metallurgy I

e. instrumentation and control f.

g.

radiological safety mechanical and electrical engineering 0

h. quality assurance practices I~

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I'T.

LUCIE - UNIT 1 ,6-8 '

ADMINI STRATI VE; CONTROLS COMPOSITION 6.5.2.2 The CNRB shall be composed of the:

Member: Vice President of Nuclear Affairs Member: Chief Engineer Power Plants Member: .Vice President of Power Resources Member: Power Plant Engineering Supervisor Member: Manager of Power Resources - Nuclear Member: Manager of QA Member: Power Plant Engineering Supervisor The CNRB Chairman shall be designated in writing.

ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the CNRB Chairman to serve on a temporary basis; however, no more than two alter-nates shall participate as voting members in CNRB activities at any one time.

CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the CNRB Chairman to provide expert advice to the CNRB.

MEETING FRE UENCY 6.5.2.5 The CNRB shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.

QUORUM 6.5.2.6 A quorum of CNRB shall consist of the Chairman or his designated alternate and four members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the facility.

ST. LUCIE - UNIT 1 6-9

ADMINISTRATIVE CONTROLS REVIEW 6.5.2.7 The CNRB shall review:

a ~ The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.

b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

C. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50. 59, 10 CFR.

d. Proposed changes in Technical Specifications or licenses.

e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

Significant operating abnormaliti'es or deviations from normal and expected performance of plant equipment that affect nuclear safety.

g, REPORTABLE OCCURRENCES requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Coranission.

h. Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

Reports and meetings minutes of the Facility Review Group.

ST. LUCIE - UNIT 1 6-10

ADMINISTRATIVE CONTROLS AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the CNRB. These audits shall encompass:

a. The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b. The performance, training and qualifications of the entire facility staff at least once per 12 months.

C. The results of all actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.

d. The performance of all activities required by the guality As'surance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.

e. The Facility Emergency Plan and implementing procedures at least 'once per 24 months.

f. The Facility Security Plan, and implementing procedures at least once per 24 months.

g. Any other area 'of facility operation considered appropriate by the CNRB or the Executive Vice President.

AUTHORITY 6.5.2.9 The CNRB shall report to and advise the Executive Vice President on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.

ST. LUCIE - UNIT 1 6-11

ADMINISTRATIVE CONTROLS RECORDS 0 6.5.2.10 Records of CNRB activities shall be prepared, approved and distributed as indicated below:

Minutes of each CNRB meeting shall be prepared, approved and forwarded to the Executive Vice President within 14 days following each meeting.

b. Reports of reviews encompassed by Section 6.5.2.7 above,. shall be prepared, approved and forwarded to the Executive Vice President within 14 days following completion of the review.

C. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Executive Vice President and to the manage-ment positions responsible for the areas audited within 30 days after completion of the audit.

6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notificati >> to the Commission shall be reviewed by the FRG and submitt~-d to the CNRB and the Vice President of Power Resources.

ST. LUCIE - UNIT 1 6-12

ADMINISTRATIVE CONTROLS 6.7 SAFETY L'IMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The facility shall be placed in at least HOT STANDBY within one hour.

b. The Safety Limit violation shall be reported to the Commission, the Vice President of Power Resources and to the CNRB within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the FRG. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems'r structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the CNRB and the Director of Power Resources within 10 days of the violation.

6.8 PROCEDURES I

6.F 1 Written procedures shall be established, implemented and main-tained covering the activities referenced below:

a ~ The applicable procedures recommended'in Appendix "A" of Regulatory 'Guide 1.33, November, 1972.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Secur ity Plan, implementation.,

a

e. Emergency Plan implementation.

6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, .shall be reviewed by the 'FR'G and approved by the Plant Manager prior to implementation and reviewed periodically as set forth in administrative procedures. P 4

ST. LUCIE - UNIT 1 6-13

ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed by the FRG and approved by the Plant Manager within 7 days of implementation.

6.9 REPORTING RE UIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 Information to be reported to the Commission, in addition to the reports required by Title 10, Code of Federal Regulations, shall be in accordance with the Regulatory Position in Revision 4 of Regulatory Guide 1.16, "Reporting of Operating Information - Appendix "A" Technical Specifications."

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a ~ Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.

b. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.

c~ Inservice Inspection Program Reviews, Specifications 4.4.10.1 and 4.4.10.2.

d. ECCS Actuation, Specifications 3.5.2 and 3;5.3.

e. Sealed Source leakage in excess of limits, Specification 4.7.9.1.3.

f. Seismic event analysis, Specification 4.3.3.3.2.

g. Beach survey results, Specification 4.7.6.1.1 ST. LUCIE - UNIT 1 6-14

ADMINISTRATIVE CONTROLS 0 6. 10

h. Core Barrel Movement, RECORD RETENTION Specifications 3.4. 11 and 4.4.11.

6.10.1 The following records shall be retained for at least five years:

a ~ Records and logs of facility operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections, repair and r eplacement of principal items of equipment related to nuclear safety.

C. All REPORTABLE OCCURRENCES submitted to the Commission.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of reactor tests and experiments.

f. Records of changes made to Operating Procedures.

g. Records of radioactive shipments.

h. Records of sealed source leak tests and results.

Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a. Records and drawing changes reflecting facility design modifi-cations made to systems and equipment described in the Final Safety Analysis Repor t.

b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c. Records of facility radiation and contamination surveys.

d. Records of radiation exposure for all individuals entering radiation control areas.

ST. LUCI E - UNIT 1 6-15

ADMINISTRATIVE CONTROLS

e. Records of gaseous and'liquid radioactive material released to the environs.

f. Records of transient or operational cycles for those facility components identified in Table 5.9-1

g. Records of training and qualification for current members of the plant staff.

h. Records of in-service inspections performed pursuant to these Technical'pecifications.

Records of guality Assurance activities required by the gA Manual.

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the FRG and the CNRB.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, 'maintained and adhered to for all operations involving personnel radiation exposure.

6.12 RESPIRATORY PROTECTION PROGRAM ALLOWANCE 6.12. 1 Pursuant to 10 CFR 20. 103(c)(1) and (3), allowance may be made for the use of respiratory protective equipment in conjunction with activities authorized by the operating license for this facility in determining whether individuals in restricted areas are exposed to concentrations in excess of the limits specified in Appendix B, Table I, Column 1, of 10 CFR 20, subject to the following conditions and limitati.ons:

a. The limits provided in Section 20. 103(a) and (b) shall not be exceeded.

ST. LUCIE - UNIT 1 6-16

ADMINISTRATIVE CONTROLS

b. If the radioactive material is of such form that intake through the skin or other additional route is likely, individual exposures to radioactive material shall be controlled so that the radioactive content of any critical organ from all routes of intake averaged over 7 consecutive days does not exceed that which would result from inhaling such radioactive material for 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months at the pertinent concentration values provided in Appendix B, Table I, Column I, of 10 CFR 20.

C. For radioactive materials designated "Sub" in the "Isotope" column of Appendix B, Table I, Column 1 of 10 CFR 20, the concentration value specified shall be based upon exposure to the material as an external radiation source. Individual exposures to these materials shall be accounted for as part of the limitation on individual dose in 520.101. These materials shall be subject to applicable process and other engineering controls. II PROTECTION PROGRAM 6.12.2 In all operations in which adequate limitation of the inhalation of radioactive material'y the use of process or other engineering controls is impracticable, the licensee may permit an individual in a restricted area to use respiratory protective equipment to limit the inhalation of airborne radioactive material, provided:

a. The limits specified in 6.12.1 above, are not exceeded.

b. Respiratory protective equipment is selected and used so that the peak concentrations of airborne radioactive material inhaled by an individual wearing the equipment do not exceed the pertinent concentration values specified in Appendix B, Table I, Column 1, of 10 CFR 20. For the purposes of this subparagraph, the concentration of radioactive material that is inhaled when respirators are worn may be determined by dividing the ambient airborne concentration by the protection factor specified in Table 6.'l2-1. for the respirator protective equipment worn. If the intake of radioactivity is later determined by other measurements to have been different than that initially estimated, the later quantity shall be used in evaluating the exposures.

ST. LUCIE - UNIT 1 6-17

ADMINISTRATIVE CONTROLS

c. The licensee advises each respirator user that he may leave the area at any time for relief from respirator use in case of equipment malfunction, physical or psychological discomfort, or any other condition that might cause reduction in the protection afforded the wearer.

d. The licensee maintains a respiratory protective program adequate to assure that the requirements above are met and incorporates practices for respiratory protection consistent with those recommended by the American National Standards Institute (ANSI-Z88.2-1969). Such a program shall include:

l. Air sampling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper selection of respiratory protective equipment.

2. Written procedures to assure proper selection, supervision, and training of personnel using such protective equipment.

3. Written procedures to assure the adequate fitting of respirators; and the testing of respiratory protective equipment for OPERABILITY immediately prior to use.

4. Written procedures for maintenance to assure full effec-tiveness of respiratory protective equipment, including issuance, cleaning and decontamination, inspection, repair, and storage.

5. Written operational and administrative procedures for proper use of respiratory protective equipment including provisions for planned limitations on working times as necessitated by operational conditions.

6. Bioassays and/or whole body counts of individuals (and other surveys, as appropriate) to evaluate individual exposures and to assess protection actually provided.

e. The licensee uses equipment approved by the'.S. Bureau of Mines under its appropriate Approval Schedules as set forth in Table 6.12-1. Equipment not approved under U.S. Bureau of Mines Approval Schedules shall be used only if the licensee has evaluated the equipment and can demonstrate by testing, or on the basis of reliable test information, that the material and performance characteristics of the equipment are at least equal to those afforded by U.S. Bureau of Mines approved equipment of the same type, as specified in Table 6.12-1.

ST. LUCIE UNIT 1 6-18

ADMINISTRATIVE CONTROLS

f. Unless otherwise authorized by the Commission, the licensee shall not assign protection factors in excess of those specified in Table 6.12-1 in se'lecting and using respiratory protective equipment.

REVOCATION 6.12.3 The specifications of Section 6.12 shall be revoked in their entirety upon adoption of the proposed change to 10 CFR 20, Section 20.103, which would make such provisions unnecessary.

6.13 HIGH RADIATION AREA 6.13.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20:

a ~ A High Radiation Area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation Work Permit and any individual or group of individ-uals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A High Radiation Area in which the intensity of radiation is greater than 1000 mrem/hr shall be subject to the provisions of 6.13.1.a above, and in addition locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty.

ST. LUCI E - UNIT 1 6-19

TABLE 6.12-1 PROTECTION FACTORS FOR RESPIRATORS PROTECTION FACTORS GUIDES TO SELECTION OF E UIPMENT PARTICULATES AND BUREAU OF HINES/NATIONAL INSTITUTE VAPORS AND GASES FOR OCCUPATIONAL SAFETY AND HEALTH EXCEPT TRITIUM APPROVALS DESCRIPTION MODEST OXIDEs I. AIR-PURIFYING RESPIRATORS Facepiece, half-mask ,

full7 NP 5 30 CFR Part ll Subpart K Facepiece, NP 100 30 CFR Part ll Subpart K II. ATMOSPHERE-SUPPLYING RESPIRATOR

l. Airline res irator Facepiece, Facepiece, half-mask full CF 100 30 CFR Part ll Subpart J CF 1,000 30 CFR Part 11 Subpart J Facepiece, full7 D 100 30 CFR Part ll Subpart J Facepiece, full PD 1,000 30 CFR Part ll Subpart J Hood Suit CF CF 5

5 30 CFR Part 6

ll Subpart J

2. Self-contained breathin a aratus SCBA Facepiece, full D 100 30 CFR Part ll Subpart H Facepiece, full PD 1,000 30 CFR Part 11 Subpart H Facepiece, full R 100 30 CFR Part ll Subpart H III. COMBINATION RESPIRATOR Any combination of air- Protection factor for 30 CFR Part ll 5 11.63(b) purifying and atmosphere- type and mode of opera-supplying respirator tion as listed above

TABLE 6.12-1 Continued TABLE NOTATION

~

See the following symbols:

CF: continuous flow

~

D demand NP: negative pressure (i.e., negative phase during inhalation)

PD: pressure demand (i.e., always positive pressure)

R: recirculating (closed circuit) z(a) For purposes of this specification the protection factor is a measure of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radio-active material outside the respiratory protective equipme~t to that inside the equipment (usually inside the facepiece) under conditions of use. It is applied to the ambient airborne concentration to estimate the concentration inhaled by the wearer according to the following formula:

Ambient Airborne Concentration Concentration Inhaled Protection Factor (b) The protection factors apply:

(i) only for trained individuals wearing properly fitted

'respirators used and maintained under supervision in a well-planned respiratory protective program.

(ii) for air-purifying respirators only when high efficiency

[above 99.9Ã removal efficiency by U.S. Bureau of Mines type dioctyl phthalate (DOP) testj particulate filters and/or sor bents appropriate to the hazard are used in atmospheres not deficient in oxygen.

(iii) for atmosphere-supplying respirators only when supplied with adequate respirable air.

Excluding radioactive contaminants that present an absorption or submersion hazard. For tritium oxide approximately half of the intake occurs by absorption through the skin so that an overall protection factor of not more than approximately 2 is appropriate when atmosphere-supplying respirators are used to protect against tritium oxide. Air-purifying respirators are not recommended for use against tritium oxide. See also footnote s, below, concerning supplied-air suits and hoods.

ST. LUCI E - UNIT 1 6-21

TABLE 6.12-1 (Continued TABLE NOTATION 4 Under chin type only. Not recommended for use where it might be possible for the ambient airborne concentration to reach instan-taneous values greater than 50 times the pertinent values in Appendix B, Table I, Column 1 of 10 CFR Part 20.

s Appropriate protection factors must be determined taking account of the design of the suit or hood and its permeability to the contaminant under conditions of use. No protection factor greater than 1,000 shall be used except as authorized by the Commission.

6 No approval schedules current available for this equipment. Equip-ment must be evaluated by testing or on basis of available test information.

" Only for shaven faces.

NOTE 1: Protection factors for respirators, as may be approved by the U.S. Bureau of Mines or the National Institute for Occupational Safety and Health according to approval schedules for respirators to protect against airborne radionuclides, may be used to the extent that they do not exceed the protection factors listed in this Table. The protection factors in this Table may not be appropriate to circumstances where chemical or other respiratory hazards exist in addition to radioactive hazards. The selection and use of respirators for such circumstances should take into account approvals of the U.S.

Bureau of Mines or the National Institute for Occupational Safety and Health in accordance with its applicable schedules.

NOTE 2: Radioactive contaminants for which the concentration values in Appendix B, Table I of 10 CFR Part 20 are based on internal dose due to inhalation may, in addition, present external exposure hazards at higher concentrations. Under such circumstances, limitations on occupancy may have to be governed by external dose limits.

ST. LUCIE - UNIT 1 6-22 0

APPENDIX B TO OPERATING LICENSE NO. DPR-67 ENVIRONMENTAL TECHNICAL SPECIFICATIONS FOR Florida Pover & Li ht Com an St. Lucie Unit No. l Docket No. 50-335 lHAR Ol 1976

TABLE OF CONTENTS

~Pa e 1.0 DEFINITIONS 1.1 National Power Emergency. 1-1 l-l

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.2 A Regional Emergency. ~ ~ ~ ~ ~ ~ ~

I'

~

1.3 Reactor Emergency . 1-1 1.4 Circulating Water System. l-l l-l

~ ~ ~ ~ ~ ~ ~

1.5 Frequency Definitions l-l

~ ~ ~ ~

1.6 Total Residual Chlorine . ~ ~ ~

1.7 Intake Temperature. ~ ~ ~ 1-2 1.8 Discharge Temperature . ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-2 1.9 Dissolved Oxygen. 1-2 1.10 Limiting Conditions 1-2 1.11 Salinity. ~ ~ ~ ~ ~ ~ ~ 1-2 1.12 Continuous Recording. 1-2 1.13 Channel Calibration . 1-2 1.14 Channel Functional Test 1-2 1.15 Batch Releases. ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-3 1.16 Continuous Release. 1-3

2. 0 LIMITING CONDITIONS 2-1 General . ~ ~ 1 t ~ ~ ~ 2-1 2.1 Thermal 2-1 2.1.1 Maximum Discharge Temperature. 2-1 2.1.2 Maximum Condenser Temperature Rise ; . ~ ~ ~ ~ ~ ~' 2-2 2.2 Chemical. 2-2 2.2.1 Biocides . ~ ~ ~ ~ ~ ~ ~ 2-3 2.2.2 pH . 2-3 2.3 Reserved. ~ ~ ~ ~ ~ ~ ~ 2-3 2.4 Radioactive Effluents . 2-3 2.4.1 Liquid Waste Effluents 2-4 2.4.2 Liquid Waste Sampling and Monitoring . 2-5 2.4.3 Gaseous Waste Effluents. ~ ~ 2-11 2.4.4 Gaseous Waste Sampling and Monitoring. 2-15 2.4.5 Solid Waste Handling and Disposal. 2-20 3.0 ENVIRONMENTAL SURVEILLANCE. 3-1 3.1 Non-Radiological Surveillance . 3-1 3.1.A ABIOTIC. 3-1

TABLE OP- CONTENTS (Cont'd)

~Pa e 3.1.A,1 Biocides 3-1 3.1.A.2 Heavy Metals . 3-1 3.1.A.3 P H ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3-1 3.'1.A.4 Dissolved Oxygen ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3-2 3.1.A.5 Salinity . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3-2 3.1.A.6 Temperature. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ 3-2 3.1.B BIOTIC. 3-3

a. Benthic Organisms ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ 3-4

b. Plankton. 3-4

c. Nektonic Organisms. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3-4

d. Macrophytes 3-4

e. Water Quality . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3-4

f. Migratory Sea Turtles ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ 3-4 3.2 Radiological Environmental Monitoring . 3-5 3.3 Onsite Meteorological Monitoring. 3-16 4.0 SPECIAL SURVEILLANCE 6 SPECIAL STUDY ACTIVITIES 4-1 4.1 Entrainment of Aquatic Organisms. ~ ~ 4-1 4.2 Impingement of Aquatic Organisms. 4-1 4.3 Minimum Effective Chlorine Usage, 4-2'.

0 ADMINISTRATIVE CONTROLS 5-1 5.1 Responsibili,ty. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-1 5.2 Organization. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-1 5.3 Review and Audit. ~ ~" ~ 5-1 if a Limiting Condition

~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~

5.4 Action to be Taken Is Exceeded. 5-4 5.5 Procedures. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-4 5.6 Reporting Requirements. 5-5 5.6.1 Routine Reports. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-5 5.6.2 Non Routine Reports. 5-7 5.6.3 Changes in Environmental" Technical Specifications. 5-,16 5.7 Records Retention . 5-16 6.0 SPECIAL CONDITIONS. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-1 6.1 Light Screen to Minimize Turtle Disorientation. 6-1

LIST OF TABLES No. ~Pa e No.

2.4.-1 Radioactive Liquid Sampling and Analysis . 2-6 2.4.-2 Liquid Waste System Location of Process and Effluent Monitors and Samplers >equired by Technical Specifications 2-9 2.4.-3 Gamma and Beta Dose Factors for St. Lucie Plant, Unit.l 2-12 2.4-4 Radioactive Gaseous Waste Sampling and Analysis 2-16 2.4-5 Gaseous Waste System Location of Process and Effluent Monitors and Samplers Required by Technical Specifications 3.2.-1 Operational Environmental Radiological Surveillance Program Sampling Locations and Vectors Sampled 3-6 3.2.-2 Operational Environmental Radiological Surveillance Program .

3.2.-3 Detection Capabilities for Environmental Sample Analysis 3-14 5.6.1-1 Effluent and Waste Disposal.

I 5.6.1-A Gaseous Effluents Summation of All Releases.

5.6.1-B Gaseous Effluents. ~ ~

5.6.1-C Liquid Effluents Summatioq of All Releases .

5.6.1-D Liquid Effluents 5-13 5.6.1-E Solid Waste and Irradiated Fuel Shipments.

5.6.1-F Format for Environmental Radiological Monitoring Program Summary. 5-6

LIST OF FIGURES No. ~Pa e 3.2-1 Radiological Environmental Surveillance Sampling Stations . ~ ~ ~ ~ ~ ~ ~ ~

3~2 2 Inset to Figure 3.2-1 . ~ ~ ~ ~ ~ ~ ~ ~ 3-10 5.2-1 FP&L Corporate Organization Environmental Affairs ~ ~ ~ ~ ~ . ~ ~ ~ 5

DEFINITI0NS I

The definitions for terms used in these environmental technical specifications are listed below.

National Power Emer enc Shall mean any event causing authorized Federal officials to require or request that Florida Power and Light supply electricity to pqints within or without the State of Florida.

A Re ional Emer enc Shall mean any of the following occurrences within the State of Florida:

(1) a catastrophic natural-disaster including hurricanes, floods, and tidal waves; or (2) other emergencies declared by State, county, municipal, or Federal authorities during which an uninterrupted supply of electric power is vital to public health and safety.

Reactor Emer enc Shall mean an unanticipated equipment malfunction necessitating prompt remedial action to avoid endangering the public health or safety.

Circulatin Water S stem Comprised of the following; velocity cap, intake pipe, intake canal, discharge canal, discharge pipe, "Y" port discharge and miscellaneous mechanical devices. The recirculation canal is included, if constructed.

Fre uenc Definitions Daily -'Not less than 360 times per annum.

Weekly Not less than 4& times per annum interval may vary by.3 days.

Monthly Not less than 12 times per annum interval may vary by 15 days.

Quarterly - Not less than 4 times per annum - interval may vary by 30 days.

Semi-annually Not. less than 2 times per annum interval may vary by 60 days.

Refueling at refueling intervals not to exceed 24 months.

Total Residual Chlorine The amount of free and combined available chlorine present in water.

1-2 Intake Tem erature The temperature of the cooling water as measured at the plant intake structure.

Dischar e Tem erature h

The temperature of the cooling water as measured near the terminus of the discharge canal.

Dissolved 0 en Oxygen dissolved in the condenser cooling water, and expressed in milligrams per liter.

Limitin Conditions Those conditions to be imposed on plant effluents and operating practices which may have an adverse impact on the environment.

~Sal knit The total amount of solid material in grams contained in one kilogram of sea water when all the carbonate has been converted to oxide, the bromine and iodine replaced by chlorine, and organic matter completely oxidized.

Continuous Recordin Recording of a measured parameter on a chart by a single pen or a multi-point recorder with less than a one-minute interval between successive printing of the same parameter.

Channel Calibration A Channel Calibration shall be the adjustment of the channel output such that it corresponds with specified range and accuracy to known values of the parameter which the channel monitors. The Channel Calibration shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the Channel Functional Test.

Channel Functional Test A Channel Functional Test shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify operability including alarm and/or trip functions.

1-3 Batch Releases Discontinuous release of gaseous or liquid effluent which takes place-over a finite period of time, usually hours or days.

Continuous Release'elease of gaseous or liquid effluent which is essentially uninterrupted for extended periods during normal operation of the facility.

2.0 LIMITING CONDITIONS General 2.0.1 The circulating water system shall be operated to result in an acceptable environmental impact. Flexibility of operation is permitted, consistent with consideration of health and safety, to ensure that the public is provided a dependable source of power even under unusual operating conditions which may set forth in this specification, as provided below in 2.0.2 and 2.0.3.

2.0.2 During a national power emergency, a regional emergency, reactor emergency,,.

or any time when the health or safety of the public may be endangered inability of Florida Power and Light to supply electricity from any by'he other sources available to"it, the operating limits provided in this specification shall be inapplicable. However, during s'uch emergencies, the operating limits shall not be exceeded except as is'necessitated by the emergency.

2.0.3 Whenever, in accordance with paragraphs 2.0.1 and 2.0.2 above, Florida Power and Light exceeds the operating limits otherwise imposed, notification shall be made to the Director of the Region II Regional Office of the Office of Inspection and Enforcement, in accordance with 5.6.2.a.

2.1 THERMAL, 2.1.1 Maximum Dischar e Tem erature

~Ob ective The purpose of this specification is to limit thermal stress to the aquatic.

ecosystem by limiting the temperature rise in the Atlantic Ocean, .in the area of the subaqueous discharge, during operation.

S ecification The thermal discharge of St. Lucie Unit No. 1 into the Atlantic Ocean shall be limited to a maximum release temperature of ill'F and shall not cause a temperature rise in excess of 1.5'F above ambient surface temperature outside a 400 acre zone of mixing during the months of June through September, nor a 4'F rise during the remaining months. In addition, the surface temperature conditions within the zone of mixing shall not exceed a rise of 5.5'F over ambient temperature nor a maximum temperature of 93'F as an instantaneous maximum at any point.

Thermal defouling of the intake pipeline is allowed subject to a maximum release temperature of 120'F and a maximum surface temperature rise of 2'F.

Under the following conditions, which may be expected to cause the discharge temperature to be higher than design, the maximum discharge temperature shall be limited to 115'F: 1) Condenser and/or circulating water pump maintenance; 2) Throttling circulating water pumps to minimize use of chlorine;

3) Fouling of circulating water system.

2-1

2-2 Temporary transients due to accidental loss of circulating water system components may cause temperature rises in excess of limitations stated above. Uariances due to these transients shall be limited to no more than 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months per month.

Monitorin Re uirement A continuous temperature measurement system shall be installed in the canal at approximately mid-depth. Temperatures shall be transmitted"'ischarge to the control room.

A continuous'temperature monitoring station located within 500 feet from the primary monitoring device shall be used as a backup system if the primary system fails. In this event this station shall be checked every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months until the primary system is restored. See Section 3.1.A.6 for complete details of the monitoring program.

2.1.2 Maximum Condenser Tem erature Rise Under normal full power operation, the temperature rise across the condenser shall not exceed 24'7. 'nder the following conditions, the condenser temperature rise shall not exceed 35'P for greater than a 72-hour period:

1) Condenser and/or circulating water pump maintenance; 2) Throttling circulating water pumps to minimize use of chlorine; 3) Fouling of circulating water system.

Monitorin Re uirements The AT across the condenser shall be determined once per hour while the unit is in operation. The system's accuracy and precision is as described in Section 3.1.A.6 of Appendix B of the technical specifications.

Bases The limitations provide reasonable assurance that the overall aquatic ecosystem in the area of the thermal plume will experience an acceptable environmental impact. The placement of the temperature monitoring instrument in the discharge canal will give the temperature of the discharge mixing with the receiving water.

water'efore 2.2 CHEMICAL Ob ective The purpose of these specifications is 1) to minimize impacts to the quality of the Atlantic Ocean, 2) to protect the local biota from lethal and sublethal effects of exposure to chemical discharges due to operation of the plant,

3) to assure that continued multiple use of the receiving waters by human populations is protected, and 4) to control the quality of the receiving medium.

2-3 Biocides S ecification Total Residual Chlorine" shall not exceed 0.1 mg/1 at any time at the terminus of the discharge canal (prior to entering the ocean outfall). If this level is exceeded, adjustments to the infection system shall be made to reduce the concentration, and each succeeding chlorination period shall be monitored until the concentration is within the specification. Chlorine shall not be added for more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per day.

Monitorin Re uirements A grab sample of condenser cooling water shall be taken weekly in the discharge canal and analyzed for total residual chlorine. The samples shall be taken during the period of chlorination. The time of beginning the chlorination and when the sample was taken shall be logged.

Bases'hen injected, chlorine is diluted by the cooling water and consumed in the process of controlling slime. To be sure that enough chlorine is injected to control the slime, the residual chlorine concentration will be approxi-mately 1 mg/1 at the condenser outlet.

2.2. 2 ~H

~S ecificaeicc The pH of the cooling water in the discharge canal shall not be less than 6.0 nor greater than 9.0 pH units.

Monitorin Re uirement pH shall be measured on a f

daily basis in the discharge canal,""and it shall be accomplished using either a grab sample or recorder.

Bases The pH limits set forth will provide reasonable assurance of an acceptable environmental impact when discharging waters to the Atlantic Ocean.

2.3 RESERVED 2e4 RADIOACTIVE EFFLUENTS Ob ective To define .the limits and conditions for the controlled release of radioactive materials in liquid and gaseous effluents to the environs to ensure that

2-4 these releases are as low as practicable. These releases should not result in radiation exposures in unrestricted areas greater than a few percent of natural background exposures. The concentration of radioactive materials in effluents shall be within the limits specified in 10 CFR Part 20.

To ensure that the releases of radioactive material above background to unrestricted areas be as low as practicable, the following design ob)ectives apply:

For liquid wastes:

a. The annual dose above background to the total body or any organ of an individual from all reactors at a site should not exceed 5 mrem in an unrestricted area.

b. The annual total quantity of radioactive materials in liquid waste, excluding tritium and dissolved gases, discharged from each reactor should not exceed 5 Ci.

For gaseous wastes:

C~ Tge annual total quantity of noble gases above background discharged from the site should result in an annual air dose due to gamma radia-tion of less than 10 mrad, and an annual air dose due to beta radiation of less than 20 mrad, at any location near ground level which could be occupied by individuals at or beyond the boundary of the site, and that no individual in an unrestricted area will receive an annual dose to the total body greater than 5 mrem or an annual skin dose greater than 15 mrem from this release quantity.

d. The annual total quantity of all radioiodines and radioactive material in particulate forms with half-'lives 'greater than eight days," above background, from all reactors at a site should not result in an annual dose to any organ of an individual in an unrestricted area from all pathways of exposure in excess of 15 mrem.

e. The annual total quantity of iodine-131 discharged from each reactor at a site should-not exceed 1 Ci.

Li uid Waste Effluents a~ The concentration of radioactive materials released in liquid waste effluents from all reactors at the site shall not exceed the value specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for unrestricted areas.

b. The cumulative release of radioactive materials in liquid waste effluents, excluding tritium and dissolved gases, shall not exceed 10 Ci/reactor/

calendar quarter.

co The cumulative release of radioactive materials in liquid waste effluents, excluding tritium and dissolved gases, shall not exceed 20 Ci/reactor in any 12 consecutive months.

2-5

d. During release of radioactivp wastes, the effluent control monitor shall be set to alarm and to initiate the automatic closure of each waste isolation valve prior to excee'ding the limits specified in 2.4.l.a above, except as provided in 2.4.2.d below.

e. The operability of each'automatic isolation .valve in the liquid radwaste discharge lines shall be demonstrated quarterly.

The equipment, installed in the liquid radioactive waste system shall be maintained and shall'e operated to process radioactive 'liquid wastes prior to their discharge when the pro)ected cumulative release could exceed 1.25 Ci/reactor/calendar quarter, excluding tritium and dissolved, gases.

go The maximum radioactivity to be contained in any liquid radwaste,tank that can be discharged directly to the environs shall not exceed 10 Ci, excluding tritium and dissolved gases.

I

h. If the cumulative release of radioactive materials in liquid effluents, excluding tritium and dissolved gases, exceeds 2.5 Ci/reactor/calendar quarter, the licensee shall make an investigation to identify the causes for such releases, define and initiate a program of action to reduce such releases to the design objective levels listed in Section 2.4, and report these actions to the NRC in accordance with Specification 5.6.2.b(1).

An unplanned or uncontrolled offsite release of radioactive materials in liquid effluents,in 'excess of 0.5,curies requires notification.

This notification', shall be in accordance. with Specification '5..6.2.b(3).

Li uid Waste Sam lin and Monitorin

a. Plant records shall be maintained of the radioactive concentration and volume before dilution of liquid waste intended for discharge and, the average dilution flow 'and length of time over which each discharge occurred. Sample analysis results and other reports shall be submitted as required by Section 5.6.1 of these Specifications. Estimates of the sampling and'nalytical errors -associated with each reported value shall be included.

b. Prior to release of each batch of liquid waste, a sample shall be taken from that batch and analyzed for 'the concentration of each significant gamma energy peak in accordance with Table 2.4-1 to demonstrate compliance with Specification 2.4.1 using the flow rate into which the waste is discharged during the period of discharge.

II C ~ Sampling and analysis,.of liquid radioactive waste shall be performed in accordance with Table 2.4-1. f

~ ~ ~ ~

~ IIIII '

~ - ~ ) = ~

~ II ~ ~

~ ~

~ II ~ ~

~ ~ ~ ~ ~

IIIII f II II ~

=: ~ . ~

~ I ~ ~

~ II ~ ~

l .. I

-: ~ -I

~ I ~ ~

~ ~

~ II ~ ~

2-7 TABLE 2.4-1 (Cont'd) 1 The detectability limits, for activity analysis are based on the technical feasibility and on the potential significance in the environment of the quantities released. For some nuclides, lower detection limits may be readily achievable, and when nuclides are measured below the stated limits, they should also be reported.

2 For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations; Under these circumstances, it will be more appropriate to calculate the concentrations of such radionuclides using measured ratios with those radionuclides which are routinely identified and measured.

3 A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged.

4 To be representative of the average quantities and concentrations of radioactive materials in liquid effluents, samples should be collected in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite should be thoroughly mixed in order for the composite sample to be representative of the average effluent release.

5 For dissolved noble gases in water, assume a MPC of 4 x 10 -5 pCi/ml of water.

6 When operational or other type of limitations preclude< specific gamma spectrum analysis of each tank, gross activity measurements shall be made to estimate the quantity and concentration of radioactive material released in the batch and a weekly sample composited from proportional aliquots from each batch released during the week shall be analyzed for the principal gamma emitting radionuclides.

7 No sampling required when cold and drained.

8 Should the reactor coolant system activity technical specification limits be exceeded, the power level and cleanup or purification flow rate at the sample time shall also be reported.

9 Required if the component cooling water (CCW) monitor is out of service or if the CCW monitor indicates activity in excess of 10 5 pCi/cc.

2-8

d. The radioactivity in liquid wastes shall be continuously monitored and recorded during release. Whenever>>these monitors are inoperable for a period not to exceed *72 hours, two independent samples of each tank to be discharged shall be analyzed and two plant personnel shall independ-ently check valving prior to the discharge. If these monitors are inoperable for a period exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , no release from a liquid waste tank shall be made and any release in progress shall be terminated.

e. The flow rate of liquid radioactive waste shall be continuously measured and recorded during release. If the flow monitors are inoperable, flow shall be determined and recorded each hour based upon an the'elease estimate of the flow rate of the system.

f. All liquid effluent radiation monitors shall be calibrated at least quarterly by means of a radioactive source which has been calibrated to a National Bureau of Standards source. Each monitor shall also have a functional test monthly and an instrument check prior to making a release.

g. The radioactivity in steam generator"blowdown shall be continuously monitored and recorded. With one steam generator blowdown monitor inoperable, the sampling system shall be realigned so that the operable monitor is receiving flow from both steam generators. Whenever both monitors are inoperable, the blowdown flow shall be diverted to the waste management system and the direct release to the environment terminated.

h. The points of release to the environment to be monitored in this section 2.4.2 include all the monitored release points as provided for in Table 2.4-2.

Bases The release of radioactive materials in liquid waste effluents to unrestricted areas shall not exceed the concentration limits specified in 10 CFR Part 20 and should be as low as practicable in accordance with the requirements of 10 CFR Part 50.36a. These specifications provide reasonable assurance that the resulting annual dose to the total body or any organ of an individual in an unrestricted area will not exceed 5 mrem. At the same time, these specifications permit the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20.

It is expected that by using this operational flexibility under unusual operating conditions, and exerting every effort to keep levels of radioactive material in liquid wastes as low as practicable, the annual releases will not exceed a small fraction of the concentration limits specified in 10 CFR Part 20. I

TABLE, 2. 4-2 ST. LUCIE PLANT LIQUID WASTE SYSTEM LOCATION OF PROCESS AND EFFLUENT MONITORS AND SAMPLERS REQUIRED BY TECHNICAL SPECIFICATIONS High Auto Control Grab Measurement Liquid Process Stream or Radiation 'to Continuous Sample Gross Dissolved Isotopic Level Release Point Alarm Isolation Valve Monitor Station Activity I Gases Alpha H-3 Analysis Alarm Miscellaneous Waste Sample (Test) Tank X . X X X (Waste and Boric Acid Condensate Tanks)

Detergent Waste Collector Tank X X (Laundry Drain Tanks)

Primary Coolant System Liquid Radwaste Discharge Pipe X Steam Generator Blow-down System X X X Outdoor Storage Tanks (potentially radio-active)

Component Cooling Systems X X Turbine Building Floor Drains (Storm Drains) x(') x( )

(a) Grab samples to be taken and analyzed each -5 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the gross activity in the secondary coolant system exceeds 10 uCilml.

2-10 The design ob)ectives have been developed based on operating experience taking into account a combination of variables including defective fuel, primary system leakage, primary to secondary system leakage, steam generator blowdown and the performance of the various waste treatment systems, and are consistent with 10 CFR Part 50.36a.

Specification 2.4.l.a requires the licensee to limit the concentration of radioactive materials in liquid waste effluents released from the site to levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for unrestricted areas. This specification provides assurance that no member of the general public will be exposed to liquid containing radioactive materials in excess of limits considered permissible under the Commission's Regulations.

Specifications 2.4.1.b and 2.4.1.c establish the upper limits for the release of radioactive materials in liquid effluents. The intent of these specifica-tions is to permit the licensee'the flexibility of operation to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the levels normally achievable when the plant and the liquid waste treatment systems are functioning as designed. Releases of up to these levels will result in concentrations of radioactive material in liquid waste effluents"at small percentages of the limits specified in 10 CFR Part 20.-

Consistent with the requirements of 10 CFR Part 50 Appendix A Design C riterion 64, Specifications 2.4.l.d and 2.4.l.e require operation of suitable equipment to control and monitor the releases of radioactive materials in liquid wastes during any period that these releases are taking place.

Specification 2.4.l.f .xequ'ires, that the licensee maintain .and operate the equipment installed in the liquid waste systems to reduce the release of radioactive materials in liquid effluents to as low as practicable with the requirements of 10 CFR Part 50.36a. Normal'se and main- con-'istent tenance of installed equipment in the liquid waste system provides reasonable assurance thag the quantity released will not exceed the design objective.

In order to keep releases of xadioactive materials as low as practicable, the specification requires operation of equipment whenever it appears that the projected cumulative discharge rate will exceed one-fourth of this design objective annual quantity during any calendar quarter.

Specification 2,4.l.g restricts the amount of radioactive material that could be inadvertently released to the environment to an amount that will not exceed the Technical Specification limit.

In addition to limiting conditions for operation listed under Specifications 2.4.1.b and 2.4.lc, the reporting requirements of Specification 2.4.l.h delineate that the licensee shall identify the cause whenever the cumulative release of radioactive materials in liquid waste effluents

2-11 exceeds one-half the design ob5ective annual quantity during any calendar quarter and describe the proposed program of action to reduce such releases to design ob)ective levels on a timely basis. This report must be filed within 30 days following the calendar quarter in which the release -occurred as required by'pecification..'5.6.2 of these Technical Specifications.

Specification 2.4.l.i provides for reporting spillage or release events which,'while below. the limits of 10 CFR Part 20, could result in releases higher than the design obgectives.

The sampling and monitoring requirements. given'nder Specification 2.4.2 provide assurance that'adioactive materials in liquid wastes are properly controlled and monitored in conformance wit:h"the-requirements of Design Criteria 60 and 64. These'requirements provide the data for the licensee and the Commission .te.'evaluate 'the plant's, performance relative to radio-active liquid wastes released: to the environment. Reports on the radioactive materials released in liquid'aste effluents are furnished to the Commission according to Section 5.6.1 'of these Technical Specifications. On the basis of such reports and'ny additional.information the Commission may obtain fram the 'licensee or others, 'the Commission.may'rom time to time require the licensee to 'take such. action as the Commission deems appropriate.-

I The points of release to the environment to"be menitored in Section 2.4.2 include all the monitored release points as provided for in Table 2.4-2.

Gaseous Waste Effluents The terms used in these .Specifications are as follows:

subscript v, refers to vent releases subscript. i, refers to individual noble gas nuclide (Refer to Table 2.4-3 for the noble gas nuclides considered)

QT

~ the total noble gas release rate (Ci/sec)

~ ZQi sum.of the individual noble. gas radionuclides determined to be present by isotopic analysis K ~ the average total body dose factor due to gamma emission (rem/yr per Ci/sec)

L ~ the average skin dose factor due to beta emissions (rem/yr per Ci/sec)

~ the average air dose factor due to beta emissions (rad/yr per Ci/sec)

N ~ the average air dose factor due to gamma emissions (rad/yr per Ci/sec)

TABLE 2 ~ 4-3 GAMA'ND BETA DOSE FACTORS FOR STa LUCIE PLANTd UNIT 1

-6 3 X/Q = 2,1 x -10 sec/m DOSE FACTORS FOR VENTS Ki L M N NOBLE GAS iV RADIONUCLIDE Total Body Skin Beta Air Gamma Air

~rem/ r ~ram/ r rad/ar ~rad/

Ci/sec Ci/sec Ci/sec Ci/sec Kr-83m 5.8,K 10 0 0.6 0. 028 Kr-85m 0. 88 3.1 4.1 0. 92 Kr-85 0. 014 2.8 4.1 0.015 Kr-87 1.9 20 22 2.0 Kr-88 6.0 5.0 6.2 6.3 Kr-89 0 5 21 22 0. 52 Xe-131m 0.4 1.0 2.3 0.5 Xe-133m 0.3 2.1 3.1 0.41 Xe-133 0. 36 0. 64 2.2 0. 45 Xe-135m 0.64 1.5 1.6 0. 68 Xe-135 1.5 3.9 5.2 1.6 Xe-137 0.072 26 27 0. 076

'.5 Xe-138 8.7 10 la6

0 2-13 The values of K, L, M and N are to be determined each time isotopic analysis is required as delineated.in Specification 2.4.4. Determine the following using the results of the noble gas radionuclide analysis:

K ~ (1/QT)~QiK i'

(1/QT)/QiLi

- (1/Q,)B,L, i

N - (1/Q,)B,N i

where the values of Ki, Li, M and Ni are provided in Table 2.4-3, and are site dependent gamma and beta dose factors Q ~ the measured release rate of the radioiodines and radioactive materials in particulate forms with half-lives greater than eight days.

a. (1) The release rate .limit of noble gases from the site shall be such that

, 2.0 Q K < 1 and Oi33 QTv(L + 1.1N )

ll (2) The release rate limit of all radioiodines and radioactive materials in particulate form with half-lives greater than eight days, released to. the environs as part of the gaseous wastes from the site shall be such that 5.5 x 10 Q (1) The average release rate of noble gases from the site during any calendar quarter shall be such that 13 Q N < 1 and 63 Q M <1 (2) The average release rate of gases from the site during any 12 consecutive months shall be

2-14 25 Q and 13 Q M (3) The a verage release rate per site of all'adioiodines and radio-active materials in particulate form with half-lives greater than eight day during a'n y calendar quarter shall be such that 13 5.5x10 3 Q < 1 (4) The average release rate per site of all radioiodines and radio-active materials in particulate form with half-lives greater than eight days during any period of 12 c'onsdcutive months'shall be s ch that 25 5.5x10 (5) The amount of iodine-.131 released during any calendar quarter shall not exceed 2 Ci/reactor.

(6) The amount of iodine-131 released during any period of the 12 consecutive months shall not exceed 4 Ci/reactor.

Should any of the conditions of 2.4.3.c(1), (2) or (3) listed below exist, the licensee shall make an investigation to, identify'he causes of the release rates, define and initiate a program of action to reduce the release rates to design objective levels listed in Section 2.4 and report these actions to the NRC within 30 days from the end of the quarter during which the releases occurred.

(1) If- the average release rate of noble gases from the site during any calendar quarter is such that 50 Q H or 25 Q M (2) If the average release rate per site of all radioiodines and radioactive materials in particulate form with half-lives greater than eight days during any calendar quarter is such that 50 5.5xl0 Q > 1 If the I

(3) amount of iodine-131 released during any calendar quarter is greater than 0.5 Ci/reactor.

2-15

d. During the release of gaseous wastes from the gas decay tanks, the gaseous discharge monitor shall be operating and set to alarm and'o initiate the automatic closure of the waste gas discharge valve prior to exceeding the limits specified in 2.4.3.a above. Whenever this monitor is inoperable for a period not to exceed seven days'two independent samples of each g'as"decay tank to be, discharged shall be analyzed and two plant personnel shall independently check valving prior to the discharge. If this monitor is inoperable for a period exceeding seven days, no release from a gas decay tank shall be made and any release in progress shall be terminated. The operability of each automatic isolation valve shall be demonstrated quarterly.

e. The maximum activity to be contained in one waste gas storage tank shall not exceed 110,000 curies (considered as Xe-133).

An unplanned or uncontrolled offsite release of radioactive materials in gaseous effluents in excess of 5 curies of noble gas or 0.02 curie of radioiodine in gaseous form requires notification. This notifica-tion shall be in accordance with Specification 5.6.2.b(3).

Gaseous Waste Sam lin and Monitorin

a. Plant records shall be maintained and reports of the sampling and analyses results shall be submitted in accordance with Section 5.6 of these Specifications. Estimates of 'the sampling and analytical error associated with each reported value should be included.

b. Gaseous releases to the environment, except from the turbine building ventilation exhaust and'as noted in Specification 2.4.4.c, shall be continuously monitored for gross radioactivity. Whenever these monitors are inoperable, grab samples shall be taken and analyzed daily for gross radioactivity. If these monitors are inoperable for more than seven days, these releases shall be terminated.

C ~ During the release of gaseous, wastes from the primary system waste gas holdup system, the iodine collection device, and the particulate collection device shall be operating, except as noted in 2.4.3.d., above.

d~ All waste gas effluent monitors shall be calibrated at least quarterly by means of a known radioactive source which has been calibrated to a National Bureau of Standards source. Each monitor shall have a functional test at least monthly and instrument check at least daily.

e. Sampling and analysis of radioactive, material in gaseous waste, including particulate forms and radioiodines shall be performed in accordance with Table 2.4-2. The points of release to the environment to be monitored in this Section 2.4.4 include all the monitored release points'as provided for in Table 2.4-4.

2-16 TABLE 2.4-4 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS Gaseous Sampling Type of Detectable Source Frequency Activity Analysis Concentrgtion (pCi/ml) 42 A. Waste Gas Decay Tank Each Tank Princi 1 Gamma Emitters 10 Releases H-3 10 43 B. Containment Purge Each Purge Princi al Gama Emitters 10 Releases H-3 10 5 ~- 42' C. Condenser Air Ejector Monthly-- Princi al Gamma Emitters 10 H-3

~42g3 D. Environmental Release Monthly ~Priaci al Gamma Emitt:ers 10 Points (Gas Samples)

'. H-3 10 Weeklv (Charcoal I-131 10 Sam le)

'I-133,I-135 Monthly (Charcoal Sam le I 10- 0 Weekly (Particulates) Principal Gamma Emitters 10

((Ba-La-140, I-131, and (Particulates) ! others Monthly Composite4 l

-11

.Gross e l

Quarterly Composite 4 ',Sr-90, Sr-89 . 10" (Particulates) 1 The above detectability limits for activity and analysis are based on technical feasi-and on the potential significance in the environment of the quantities released. 'ility For some nuclides, lower detection limits may be readily achievable, and when nuclides are measured below the stated limits, they should also be reported.

For 'certain mixtures of gamma emitters, it may not be possible to measure radionuclides at levels near their sensitivity limits when other nuclides are present in the sample at much higher levels. Under these circumstances, it will be more appropriate to cal-cu3.ate the levels of such radionuclides using observed ratios with those radionuclides which are measurable.

Analyses shall also be performed following each refueling, startup, or similar opera-4tional occurrence which could alter the mixture of radionuclides.

To be representative of the average quantities and concentrations of radioactive materials in particulate form released in gaseous effluents, samples should be collected in pro-5portion to the rate of flow oi the effluent stream.

Required when the gross activityi in tbedsecoadary coa1ant system, as required to be determined in Appendix A of these technice3. specifications, exceeds 10 uCi/ml.

i ~n Bt.d(~.V . il '" i '.'i

TABLE 2.4-S ST. LUCIE PLANT GASEOUS WASTE SYSTEM LOCATION OF PROCESS AND EFFLUENT MONITORS AND SAMPLERS REQUIRED BY TECHNICAL SPECIFICATIONS Grab Auto Control to Continuous Sample Measurement Process Stream or Release Point Alarm Isolation Valve Monitor Station Noble Gas I Particulate H-3 Alpha Waste Gas Storage Tanks X X Condenser Air Effector X X X X Building Ventilation Systems Reactor Containment Building (whenever there is flow to Plant Vent) X Auxiliary Building (to Plant Vent) X X X" Fuel Handling 6 Storage Buildings X X Radwaste Area (to Plant Vent)

Steam Generator Blowdown Tank Vent or Condenser Ventb x(') X(6) X Turbine Gland Seal Condenser X Waste Evaporator Condenser Vent c,e X a

If any or all of the process streams or building ventilation systems are routed to a single release point, the need for a continuous monitor at the individual discharge point to the main exhaust duct is eliminated. One continuous monitor at the final release point is sufficient.

b In some PWRs the steam generator blowdown tank vent is routed to the main turbine condenser, and the need for a continuous monitor at this release point is eliminated.

For PWRs in which the waste evaporator condenser is vented directly to the atmosphere.

Monitoring system will be installed by March 1, 1976, if this system is still operational.

Monitored via Condenser Air E)ector System.

2-18 Beses The release off radioactive materials in gaseous, waste effluents to unrestricted areas shall not exceed the concentration iimits specified in 10 CFR Part 20 and should be as low as practical in accordance with the requirements of 10 CFR Part 50.36a. These Specifications provide reasonable assurance that the resulting annual air dose from the site due to gamma radiation will not exceed 10 mrad, and an annual air dose from the site due to beta radiation will not exceed 20 mead from noble gases, that no individual in an unrestricted area will receive an annual dose to the total body greater than 5 mrem or an annual skin dose greater than 15 mrem f'rom fission product noble gases, and that the annual dose to any organ of an individual from radioiodines and radioactive material in particulate form with half-lives greater than eight days will not exceed 15 mrem per si'te.

At the same time these Specifications permit the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided with a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20. Even with this operational flexibility under unusual operating conditions, if the licensee exerts every effort to keep levels of radioactive material in gaseous waste effluents as low as practicable, the annual'eleases will not exceed a'mall fraction of the concentration limits specified in 10 CFR Part 20.

The design objectives have been developed based on operating experience taking into account a combination of system variables including defective fuel, primary system leakage, primary to secondary system leakage, steam generator blowdown, and the performance of the various waste treatment systems.

1 Specification 2.4.3.a(1) limits the release rate of noble gases from the site so that the corresponding annual gamma and beta dose rate above back-ground to an individual in an unrestricted area will not exceed,500 mrem to the total body or 3000 mrem to the skin in compliance with the limits of 10 CFR Part 20.

For Specification 2.4.3.a(1), gamma and beta dose factors for the individual noble gas radionuclides have been calculated for the plant gaseous release points and are provided in Table 2.4-3. The expressions used to calculate these dose factors are based on dose models derived in Section 7 of Meteorolo and Atomic Ener -1968 and model techniques provided, in Draft Regulatory Guide 1.AA.

Dose calculations have been made to determine the site boundary location with the highest anticipated dose rate from noble gases using onsite meteorological data and the dose expressions provided in Draft Regulatory Guide 1.AA. The dose expression considers the release point location, building wake effects, and the physical characteristics of the radionuclides.

2-19 The offsite .location with the highest anticipated annual'dose from released noble gases is 1600 meters in the North direction.

The release rate Specifications for a radioiodine and radioactive material in particulate form with half-lives greater than eight days are dependent on existing radionuclide pathways to man. The pathways which were examined for these Specifications are: 1) individual inhalation of airborne radio-nuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption -by man, and 3) deposition onto grassy areas where milch animals graze with consumption of the milk by man. Methods for doses to the thyroid via these pathways are described in Draft

'stimating Regulatory Guide 1.AA. The offsite location with the h'ighest 'anticipated thyroid dose rate from radioiodines and radioactive material in particulate form with half-lives greater than eight days was dete'rmined using onsite meteorological data and the expressions described in Draft Regulatory Guide 1.AA. Specification 2.4.3.a(2) limits the release rate of radioiodines and radioactive material in particulate form with half-lives greater than eight days, so that the-corresponding annual thyroid dose via the most restrictive pathway is less than 1500 mrem.

For radioiodines and radioactive material in particulate form with half-lives greater than eight 'days, the most restrictive location is a residence located 3,000 meters in the WSW.direction (vent X//~5,.5x10 sec/m3).

Specification 2.4.3.b establishes upper offsite levels for the releases of noble gases and radioiodines and radioactive material in particulate form with half-lives greater than eight days at twice the design objective .annual quantity during any calendar quarter, or four times the design objective annaul quantity 'during any period of 12 consecutive months. In addition to 'the limiting conditions for operation of Specifications 2.4.3.a and 2.4.3.b, the reporting requirements of 2.4.3.c provide that the cause shall be identified whenever the release of gaseous effluents exceeds one-half the design objective annual quantity during any calendar quarter and that the proposed program of action to reduce such release rates to the 'design objectives shall be described.

Specification 2.4.3.d requires that suitable equipment to monitor and control the radioactive gaseous releases are operating during any period these releases are taking place.

P Specification 2.4.3.e limits the maximum quantity of radioactive gas that can be contained in a waste gas storage tank. The calculation of this quantity should assume instantaneous ground release, a X/g based 5 percent meteorology, the average gross energy is 0.19 Mev per disintegration (considering Xe-133 to be the principal emitter) and exposure occurring at the minimum site boundary radius using a semiinfinite cloud model. The calculated quantity will limit the offsite dose above background to 0.5 rem or less, consistent with Commission guidelines.

~ 2-20 Specification 2.4.3.f provides for reporting release events which, while below the limits of 10 CFR Part 20, could result in releases higher than the design objectives.

The sampling and monitoring requirements given'nder Specification 2.4.4 provide assurance that ra'dioactive materials released in gaseous waste effluents are properly controlled and monitored in conformance with the requirements of Design Criteria 60 and 64. These requirements provide'he data for the licensee and'he Commission to evaluate the plant's performance relative to radioactive waste effluents released to the environment. Reports on the quantities of radioactive materials released in gaseous effluents are furnished to the Commission on the basis of Section 5.6.1 of these Technical Specifications. On the basis of such reports and any additional information the Commission may obtain from the licensee or others, the Commission may from time to time require the licensee to take such action as the Commission deems appropriate.

Specification 2.4.4.b excludes monitoring the turbine building ventilation exhaust since this release is expected to be a negligible release point.

Many PWR reactors do not have turbine building enclosures. To be'onsistent in this re'quirement for all PWR reactors, the monitoring of gaseous releases from turbine buildings is not required.

2.4.5 Solid Waste Handlin and Dis osal

~ as Measurements shall be made to determine or estimate the total curie quantity and 'principle radionuclide "composition of all radioactive solid waste shipped offsite.'eports

b. of the radioactive solid waste shipments, volumes, principle radionuclides, and total'urie quantity, shall be submitted,in

. accordance with Section 5.6.1; .

Bases h I The requirements for solid radioactive waste handling and disposal given under Specification 2.4.5 provide assurance that solid radioactive materials stored at the plant and shipped offsite are'packaged in conformance with 10 CFR,Part 20, 10 CFR Part 71, and 49 CFR Parts 170-178.

3.0 ENVIRONMENTAL SURVEILLANCE Non-Radiolo ical Surveillance 3.1.A ABIOTIC 3.1.A.1 Biccides

~Ob ective The purpose of this surveillance is to monitor Total Residual Chlorine in the discharge canal to insure that no adverse impact on the environment is occurring.

S ecification l

Total Residual Chlorine shall be monitored in the discharge canal on a weekly basis while'a condenser .section is being chlorinated. See Section 2.2.1 for limiting conditions.

Re ortin Re uirement Total Residual Chlorine concentration shall be reported in the Annual Environmental Monitoring Report.

~ *~

~Ob ective The purpose of this study is to monitor heavy metals concentrations in the intake and discharge canals to detect any measurable increase in heavy metals.

S ecification Grab samples shall be taken on a monthly basis. at the intake and'discharge canals and analyzed for Mercury, Arsenic, Chromium, Copper, Iron, Lead,,

Nickel, and Zinc.

Re ortin Re uirement Concentrations shall, be reported in the Annual Environmental Monitoring Report.

3.1.A.3 pH

~Ob ective The purpose of this surveillance is to monitor pH in the receiving waters to insure that pH is not being raised or lowered from the specified limits, in order to prevent an adverse environmental impact.

3-2 S ecification pH shall be monitored daily using grab samples or" a recorder in the discharge canal. See Section 2.2.2 for limiting conditions.

Re ortin Re uirements pH measurements shall be reported in the Annual Environmental Monitoring Report.

3.1.Ae4 Dissolved Ox en

~Ob ective The purpose of this surveillance is to monitor dissolved oxygen (DO).

S ecification DO shall be monitored weekly, using grab samples, in the intake and discharge canals.

Re ortin Re uirements Concentrations shall be reported in the Annual Environmental Monitoring Report.

3. 1.A. 5 ~tel icit

~Ob ective The purpose of this specification is to measure salinity concentrations in receiving waters.

S ecification Salinity shall be monitored by grab samples on a weekly basis in the discharge canal.

Re ortin Re uirements Salinity concentrations shall be reported in the Annual Environmental Monitoring Report.

Ob ective To provide temperature data to limit thermal stress to the aquatic ecosystem.

3-3 S ecification A continuous temperature measurement system shall monitor circulating water-temperature at the intake to Unit 1 and in'"the discharge canal. Both intake.

and discharge water temperature monitors shall have an accuracy of +2'P.

Signals shall be transmitted 'to the control room and displayed. The system shall have an alarm function to alert the control room operator of circulating water temperatures being. at the maximum allowable limit.

A back-up system shall also be operable to monitor temperatures whenever the primary system fails. -The back-up system does not have to transmit temperatures to the control room. Its overall accuracy shall also be +2'F.

The maximum discharge temperature limitations shall be as described in Section 2.1.1.

In order to demonstrate compliance with the temperature rise limitations outside the zone of mixing, infrared aerial photography shall be employed, along with field measurements for ground truth. Pour flights shall be scheduled during the first year of operation of Unit No. 1 after the unit is available for loading above 80K power level. Flights shall be spaced at approximately three month intervals, weather -permitting, when the unit is operating at a power level of 80% or greater.

To demonstrate compliance-with the temperature rise limitations within the zo'ne of mixing, two self-contained recording thermographs shall be used. =

One thermograph shall be located at the surface of the water, at the point of maximum surface temperature'of the Unit No. 1 discharge. This point has been determined by previous modeling studies.

second thermograph shall be located at the'surface near the intake velocity cap of the Unit No. 1 to determine ambient temperature. These thermographs shall have a sensitivity of 0.5'P in a,range-from 40'P to 100'F.

Re ortin Re uirement Results of this thermal monitoring program shall be summarized in the Annual Environmental Monitoring Report.

BIOTIC

~ob ective To determine the effects of plant operation on the planktonic, nektonic, and benthic populations of the Atlantic Ocean near- the discharge during plant operation.

3-4 S ecification The biological conditions shall be assessed, 1) in terms of abundance and compositions of the marine biotic community, and 2) the relationship between certain chemical and physical properties of the waters and the character of the biological community.

The five sampling locations established during a pre-operational baseline biology program will be utilized for plankton, trawl, and benthic collec-tions. The sampling schedule will be as follows:

a. Benthic Or anisms Benthic organisms will be co11ected quarterly and inventoried as to type and abundance of major t'axonomic groups present.

b. Plankton Plankton samples w&l be collected monthly. Both zooplankton and phytoplankton species will be identified as to kind and abundance.

Chlorophyll "a" analysis will be performed as 'a measure of primary productivity.

C ~ Nektonic Or anisms - Samples will be col'lected monthly by'rawling, seining, or other suitable method. Types and numbers of organisms present will be determined, including species of migratory fish of commercial and sports fisheries value such as blue fish and mackerel.

and identified as to species and abundance.

e.

mid-depth, and surface levels at the same time as the. biotic samples are collected. Parameters studied will be'temperature, salinity, dissolved oxygen content., turbidity, and selected nutrients.

Mi rator Sea Turtles The species, numbers, and nesting characteristics of sea turtles that migrate in from the sea and nest along the east coast"of Florida will be determined on the FPL shoreline property and selected adjacent control areas in 1975 and 1977. A study shall be conducted to determine the effects of the discharge thermal plume on turtle nesting patterns and turtle hatchling migration. In addition, control studies on temperature stress, hatching, and rearing factors will be conducted using turtle eggs from displaced nests.

Based on the data obtained, predictions will be made on the impact of the plant's operation on baseline biological conditions and current uses of the waters.

Florida Power and Light will review the data after two years of plant operation.

If effects attributable to the plant are found acceptable, the results shall be reviewed by NRC to determine if the biotic program, or any portion thereof, should be terminated.

3-5 Re ortin Re uirement Results of the biological program shall be reported in the 'Annual Environmental Monitoring Report.

3.2 RADIOLOGICAL ENVIRONMENTAL MONITORING

~Ob ective The Operational Radiological Environmental Surveillance Program is conducted to measure radiation levels and radioactivity in the environs, and to'assist in verifying any projected or anticipated radi'oactivity release resulting from plant operations which could bring about public exposure to radiation.

S ecifications 3.2.a Environmental samples shall be collected at the designated locations shown in Table 3.2-1 and Figures 3.2-1 and 3.2-2.

3.2.b The criteria for the type and the number of samples to be collected at a given sampling location, the frequency of collection, and the type and frequency of radioactivity analysis to be completed on the collected samples shall be as shown in Table 3.2-2.

Direct radiation shall be, measured by thermoluminescence dosimetry (TLD) at locations shown in Table 3.2-1 and Figures 3.2-1 and 3.2-2. The system shall be capable of measuring 26 mrem/year with a precision of +10% at the 95%

confidence level based on a quarterly collection frequency.

3~2~c The radiation detection capabilities of the radioanalytical methods used shall be as shown in Table 3;2-,3.

3.2.d A census of gardens producing fresh leafy vegetation for human consumption shall be conducted near the end of the growing season t'o determine their location with respect to the plant site. This census is limited to gardens having an area of 500 ft2 or more, and shall be conducted under the following conditions:

l. Nothin a 1 mile radius of the plant site, enumerated by door-to-door or equivalent counting technique.

2. If no milk-producing animals are located in the vicinity of the site, as determined by Specification 3.2.e below, the census described in item 1, above, shall be extended to a distance of 5 miles from the site.

3. If this ct census reveals the existence of a garden at a location yielding a calculated thyroid dose. greater than that from a previously sampled garden, the new location shall replace the garden previously having the maximum iodine concentration. Also, any location from which fresh leafy vegetables can no longer be obtained may be dropped from the

TABLE 3. 2-1 ST. LUCIE PLANT OPERATIONAL RADIOLOGICAL ENVIROiiiNTALSURVEILLANCE PROG1VA SAMPLING LOCATIONS AND VECTORS SAMPLED Station No. Descri tion B~earin

Distance* Vector Sam led H03 Meadowbrook Dairy, Glades Cut-off Road, 260 22.526 km Milk

~

St. Lucie County (14.00 mi)

H08 Florida Power & Light Company Substa- 293 9.170 km Soil, Air Particulates & Iodine, tion White City, Weatherby Road west (5.7 mi) Direct Radiation of U. S. 1

'09 Florida Power & L'ight Company Substa- 196 11.745 km Soil, Air Particulates & Iodine, tion west of U.S. 1, just south of (7.30 mi) Direct Radiation St. Lucie County Line H10 Indian River Field Laboratory, Univer- 300 19.308 km Food Crops (Citrus), Air Particulatesh sity of Florida, west of SR 713 (12.00 mi) Iodine, Direct Radiation, Soil Hll City of Ft. Pierce, Water System 323 14.480 km Potable Water (Well) City of Ft.

Collected at St. Lucie County Health (9 mi) Pier'ce Department, Ave., "C", Ft. Pierce H12 Florida Power & Light Company Substa- 180 19.308 km Potable Water (Well) City of Stuart, tion SR 76 west of U.S. 1, Stuart, (12.00 mi) Air Particulates & Iodine, Direct Martin County Radiation H13 On Site, Point north of Big Mud Creek 312 1021 m Surface Water, Bottom Sediment at Indian River (0.63 mi) (Estaurine)

TABLE 3.2 1 (Continued)

Station Descri tion ~Bearin

Distance* Vector Sam led No.

H14 Employees Parking Lot, southeast of 160 503 m Air Particulates S Iodine, Direct Containment (0.31 mi) Radiation H15 Site, Beach near Discharge Structure 0 On 89 808 m Ocean Water 6 Bottom Sediment, (0.50 mi) Aquatic Biota H16 Beach (ocean) opposite Blind Creek 31 1509 m Ocean Bottom Sediment, Beach Sand (0.94 mi)

H19 On Site, Beach south of. Intake Canal 161 1494 m Ocean Bottom Sediment, Beach Sand (0..90 mi)

H22 Lentz Groves, U.S. 1 210 8. 849 km Food Crop (Citrus)

(5.50 mi)

H23 Montauk Groves, U.S. 1, south of 270 7.562 km Food Crop (Citrus)

Easy Street (4.70 mi)

H24 Poster Groves, U.S. 1, north of 300 8.608 km Food Crop (Citrus)

Tumblin Kling Road (5.35 mi)

H25 Childs Groves, Bell Avenue, west of 297 11. 263 km Food Crop (Citrus)

Sunrise Blvd. (7.00 mi)

H26 Wouters Groves, west-of SR 713 on .314 21.720 km Food Crop (Citrus)

Immokola Road (13.50 mi)

H30 Residence, 7609 Indian River Drive 245 3. 218 km Ground Water (Well), Soil, Air (2.00 mi) Particulates 6 Iodine, Direct Radiation

TABLE 3,2-1 (Continued)

Station No. Descri tion Bea~ince* Distance* Vector Sam led H31 North Port St. Lucie Water System, 2500 10.619 km Potable Water (Well) Port Prima Vista Blvd. (6.60 mi) St. Lucie H32 Department of Health and Rehabilit- 3380 30.571 km Aquatic Biota, Ocean Water &

ative Services Entomology Laboratory, (19. 00 mi) Bottom Sediment, Air ParticQlates East of U.S. 1, Vero Beach 6 Iodine,.'.Soil, Direct Radiation, Beach Sand H33 On Site, between Canals, east of AIA 1380 945 m Air Particulates 6 Iodine, Direct (0.59 mi) Radiation H34 On Site, Meteorological Tower 270 762 m Air Particulates & Iodine, Direct (0.47 mi) 'adiation H36 On Site, Discharge Canal west of AIA 101'05(0.19 m mi)

Surface Water, Bottom Sediment H39 Vista Royal Condominium, 1 mile north 338 32.180 km Pood Crop (Citrus) of H32, east of U.S. 1, Vero Beach (20.00 mi)

H40 Davis Dairy, Military Trail, west of Boynton 172o 89.770 km Milk Beach, Palm Beach County (55.77 mi)

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'I I I SAMPLING LOCATIONS .

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ST. LUCIE .PLANT SAMPLING LOCATIONS Q.Agr, PRoPERrv .uHz..

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TABLE 3.2.2

)0 SHEET 1 OPERATIONAL ENVIRONMENTAL RADIOLOGICAL SURVEILLANCE PROGRAM ST. LUCIE PLANT Exposure Pathway Collection Type and Rxequency and/or Sam le Criteria and Sam lin Locations Fre uenc of Anal sis

1. AIR 1.1 Particulate and Comparison on-site versus off-site Weekly Gross Beta Iodine & reference locations: Gamma spectral analysis 3 'locations on-site, north, east, of monthly composite

& southeast of the plant- H 34, Radioactive Iodine H 14, H 33 Sr-89 & 90 (Quarterly Composite) 5 locations off-site within a radius of 10 miles of plant~H 08, H 09, H 10, H 12, H 30, and 1 control location: H32 1.2 Direct Radiation Comparison of on-site versus off- Quarterly Determine direct radiation site & reference locations: exposure by TLD readout 3 locations on-site, north, east, (mean of 2 TLDs)

& southeast of the plant:H 34, H14, H33' locations off-site within a radius of 10 miles of plant:H 08, H 09, H 10, H 12, H 30, and 1 control location: H32

2. MATER 2.1 Surface Water 2.1.1 Discharge Canal 1 location, west-of AIAg H36 Monthly Gamma spectral analysis Tritium (Quarterly Composite)

Sr-89 & 90 (Quarterly Composite) 2.1.2 Ocean 2 locations; H15 & H32 (Control) Monthly Gamma spectral analysis I Tritium (Quarterly Composite)

Sr-89 & 90 (Quarterly Composite) 2.1.3 Estuarine 1 location;- Big Mud Creek: H13 Quarterly Gamma spectral analysis -.

Tritium 2.2 Ground Water (well) 1 location;.-Re'sidence, 7609 Indian .Semi-annually Gamma Spectral Analysis River Drive: H30 Gross Beta Tritium

)0 TABLE 3.2.2

, ~

SHEET 2 OPERATIONAL ENVIRONMENTAL RADIOLOGICAL SURVEILLANCE PROGRAM ST. LUCIE PLATE Exposure Pathway Collection Type and Frequency Criteria and Sam lin Locations Fre enc of Analysis

2. WATER (coco'd)

2. 3 Potable Hater 1 location, =City of Ft. Pierce, Quarterly Gamma spectral analysis (wells). drinking water supply, H 11 Gross Beta 1 locations City of Stuart, Tritium drinking water supply, H 12 1 location, Port St. Lucie, drinking water supply, H 31 3~ BOTTOM SEDIMENT 3.1 Discharge Canal 1 location~ west of AIAs H36 Semi-annually Gamma spectral analysis Sr-90

3. 2 Ocean 1 location, beach'west, of discharge Semi-annually Gamma spectral analysis-structure: H15 Sr-90 1 location, offshore, 1 mile north of dischargess HI6 1 location, offshore, 1 mile south of discharge: H19 1 location, offshore, Vero Beach:

H32 (Control) 3.3 Beach (sand) 1 location, east; of Blind Creek, 1 Semi-annually Gamma spectral analysis mile north of discharges H16 Sr-90 1 location, near intake, 1 mile south of discharge: H19 1 location, Vero Beach: H32 (Control) 3.4 Estuarine 1 location,'ig=Mud Creeks 813 Semi-annually Gamma spectral analysis 4.1 Crustacea 1 location; vicinity of discharge Semi-annually Gamma spectral analysis (Lobster or crab structure: H15' or shrimp) location, Vero Beachs H32 (Control)

TABLE 3.2.2 SHEET 3 OPERATIONAL ENVIRONMENTAL RADIOLOGICAL SURVEILLANCE PROGRAM ST- LUCIE PLANT Exposure Pathway Collection Type and Frcguency and/or Sam le Criteria and Sam lin Locations Pre uenc of Anal sis (cont'd)

4. 2 Fish 4 2 1 Carnivores 1 location, vicinity of discharge Semi-annually Gamma spectral analysis structure: H 15 Sr<<89 S 90 1 location, Vero Beach: H32 (Control)

I 4.2.2 Herbivores 1 location,,vicinity of discharge Semi-annually Gamma spectral analysis

'strur tore: H15 Sr-89 6 90 1 location, Vero Beach< H32 (Control) 5~ TERRESTRIAL B.

1 location within 15 mile radius Semi~nthly of plant and in the prevailing Gamma spectral analysis Sr-89 6 90 wind direction from the planti H03 1-131 1 location, 55.77 mi see& of the Monthly Gamma spectral analyyis plant, Palm Beach County H40 '(Control) Sr-89 6-90 I-131 Dairy herd census Semi-annually 5.2 Biota 5.2.1 Food Crop 6 locations, H10, H22, H23, H24, Harvest Time Gamma spectral analysis (Citrus) H25i H26 Sr-89 6 90 1 1ocation,.Vero Beach: H39 (control) Harvest Time gamy gpygtral analysis 5.2.2 Food Crop 1 location as determined by garden Harvest Time Gamma spectral analysia (Edible Leafy census (Specification 3,2.d) I-131 vegetation)

5. 3 Soil 5 locations within a 15 mile radius Once per 3-year Gamma spectral analysis of plant: H03, H08, H09, H10, H30. period Sr-90 1 location, Vero Beach: H32 (Control)

TABLE 3.2-3 St. Lucie Plant: Detection Ca abilities for Environmental Sam le Anal sis Media Detection Capabilities*

Analysis Water Airborne Particulates Fash, Meat Milk Vegetation Soil (pCi/1) or Gas or Poultry [pCi/1) (pCi/kg,wet) (pCi/kg,dry)

( Ci/m3) ( i/k wet)

Gross beta 0.8 0. 002.

3H 199. 0 54 Mn 6. 0. 17.0 59FFe 5.0 58CCo 7.0 19. 0 6'0 Co 7.0 20.0 65 Zn 14. 0 39.0 89SSr 1,6 0.005 2.0 8.0 90SSr 0.8 0.002 4.0 1.0 10.0 95 Zr-Nb 7.0 7.6 0.008 0.5 16.0 134CCs 6.0 0.008 18. 0 6.0 26. 0 137CCs 7.0 0.008 18.0 7.0 26.0 140 Ba-La 8.0 0.008

Nominal: LLD's (lower limit of detection) calculated as defined in HASL-300, Bev. 8/73, pp 08-Olg 02, 03, at the 90t confidence level. The detection levels for radionuclides analyzed by gamma spectrometry. will vary according to the number of radionuclides encountered in environmental samples.

3-15 surveillance program after notifying NRC in writing that such vegetables are no longer grown at that location.

3~2~e A census of animals producing milk for human consumption shall be conducted semiannually to determine their location and number with respect to the plant site. The census shall be conducted qnder the following conditions:

1. Within a 1 mile radius from the plant site or within a 15 mrem/year-isodose line=(as calculated using dose models presented in Regulatory Guide 1.42), whichever is larger, enumeration by a door-to-door or equivalent counting technique'.

2. Within a 5 mile radius for cows and a 15 mile radius for goats, enumera-tion by using referenced information from county agricultural agents or other reliable sources.

If it is determined from the census that animals are present at a location which yields a calculated thyroid'ose greater than that from previously sampled animals, the new location shall be added to the surveillance program as soon as practicable. The sampling location having the lowest calculated dose may be dropped from the surveillance program 3 months after sampling begins at the new location. Also, any location from which milk can no longer be obtained may be dropped from the surveillance program after notifying NRC in writing that milk-producing animals are no longer present at that location.

3.2.f Deviations from the required sampling schedule are permitted if specimens are not obtainable due 'to', hazardous conditions, seasonal'navailability, or malfunction of autom'atic sampling equipment. In the latter case, every reasonable effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be described in the annual report.

A deviation of not greater than one'week from the required frequency of analysis for gross beta activity, as shown in Table 3.2-2, is permitted if equipment failure delays the analyses. Every reasonable effort shall, be made to effect expeditious equipment repair.

~Re ortkn All required reports from this Operational Radiological Environmental Surveillance Program shall be. prepared and presented in the manner described in Section 5.6.1.B of these Environmental Technical Specifications.

Bases The program is designed to determine existing radioactivity levels and to detect changes in radiation levels in the air, water and land environment which may be attributed to the oper'ation of the plant. The methods,

3-16 procedures and techniques used were developed during the preoperational phase and have provided background measurements that will be used as a base for distinguishing significant changes in radioactivity in the site environs.

3.3 ONSITE METEOROLOGICAL MONITORING

~Ob ective

'The objective of onsite meteorological monitoring is to adequately measure and document meteorological conditions at the site, specifically at heights above ground that are representative of atmospheric conditions that exist at all effluent release points.

S ecification The onsite meteorological monitoring program shall conform to the recommenda-tions and intent of Regulatory Guide 1.23, Gnsite Meteorolo ical Pro rams, and include instruments to sense wind speed and direction at 33 ft and 190 ft, vertical temperature gradient between 33 ft and '200 ft, and ambient dry bulb and dewpoint temperatures at 33 ft. The location of the meteorological tower shall be located approximately 2400 ft north of the reactor complex.

Re ortin Re uirements Meteorological data shall be summarized and reported in a format consistent with the recommendations of Regulatory Guides 1.21 and 1.23. Summaries of data- and observations shill be available to the U. S. Nuclear'Regulatory Commission'pon request. If the outage of any meteorological instrument(s) exceeds seven consecutive days, the'total outage time and dates of outage, the cause of the outage, and the instrument(s),involved shall be reported .

within 30 days of the initial time of the outage to the U. S. Nuclear Regulatory Commission, Office of Inspection and Enforcement, with a copy to the Office "of Nuclear Reactor Regulation, Division of Technical Review.

Any modifications to the meteorological monitoring program, as described above, or alterations of'he area near the meteorological tower that would, interfere with the measurement of meteorological conditions representative of the site, shall have the written approval of the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, prior to the initiation of the modification or alteration.

Bases The collection of meteorological data at the plant site will provide information which will be used to develop atmospheric diffusion parameters to estimate potential radiation doses to the public resulting from actual routine or accidental releases of, radioactive materials to the atmosphere.

4' SPECIAL SURVEILLANCE AND SPECIAL STUDY. ACTIVITIES Entrainment of A uatic Or anisms

~ob ective The purpose of this study is to assess the effects on planktonic organisms of passage through the plant condensers. Specialists in the biological sub-disciplines of zooplankton and ichthyology will perform appropriate portions of this study. Pigures obtained for the intake and discharge canals will be compared to data collected at a control station.

S ecification Samples shall be collected from the intake and discharge canals and a control station at monthly intervals when the unit is in operation to identify the organisms involved, and to attempt to quantify how many of each organism are potentially affected. Biomass measurements, numbers of eggs collected, and numbers and identification 'of larvae to the level of major taxonomic groups, if possible shall be performed. Present "state-of-the-art" informa-tion shall be used to attempt to quantify the mortality of the organisms due to entrainment. This program shall determine the seasonal abundance of fish eggs and larvae.

Re ortin Re uirements Results of this study shell be summarized in the Annual'Environmental Monitoring Report. If, at the end of two years, no significant problem is evident, an option to formally delete this portion of the Technical Specifications may be initiated.

4e2 Im in ement'of A uatic Or anisms

~0b ective The purpose of this study fs to assess the .impingement of aquatic organisms on intake screens and the environmental -impact of the impingement.

S ecification Intake screens washings shall be examined for a =consecutive twenty-four hour period, twice a week whenever the Unit 1 circulating water pumps are operating.

The collected washings shall be analyzed for the species present, number of each individual species caught, total biomass of each species, and the average size of the individuals caught.

Re ortin Re uirements Data collected shall be analyzed monthly for the first year of operation and a report sent to the NRC within 45 days of each monthly period. After'he first year of operation, the data shall be analyzed every six months, and the results summarized in the Annual Environmental Monitoring Report.

4-1

Minimum. Effective Chlorine Usa e Ob ective k (

The purpose of this study is to determine the minimum amount of chlorine necessary which will afford adequate protection 'to the condenser while avoiding unnecessary discharge of chlorine to the environment.

S ecification A program shall be initiated after Unit 1 has.initially reached 75% power level. The initial chlorine injection rate shall be determined based on preoperational data, previous experience, and laboratory chlorine demand tests. After reaching a power plateau above 75% power, a controlled incremental reduction of the chlorine injection rate shall be implemented.

Condenser fouling shall, be monitored. in coordination with chlorine reduction.

Re ortin Re uirements The results of this study shall be summarized. in the Annual Environmental Monitoring Report. When the minimum'level of chlorine usage, as determined by the study, has been reached, a proposal shall be submitted to the NRC to terminate the study.

5.0 ADMINISTRATIVE CONTROLS The purpose of this section is to describe the administrative and management controls necessary to provide continuing protection to the environment, and to implement the environmental technica1 specifications (ETS).

5.1 Res onsibilit The Vice President of the Environmental'Department has the ultimate responsi-bility for the implementation of the ETS. He may delegate to other departments and/or organizations the work of establishing and executing portions of the ETS, but shall retain responsibility thereof.

The Environmental Department is respensible for executing the non-radiological biotic and special studies sections of the ETS. The Vice President of Power Resources is responsible for executing the non-radiological abiotic, radio-active effluents, and the Radiological 'Environmental. Surveillance sections.

The Plant Quality"Control group shall be responsible for the day-to-day verification of compl'iance with the ETS. The Manager of Quality Assurance shall be responsible for periodic audits, conducted according to the corporate Quality 'Assurance program, to insure compliance with the ETS.

5.2 Or anization The corporate organization involved in environmental matters is depicted in Figure "5. 2-1.

5.3 Review and Audit 5.3.1 Review of implementation of the ETS shall be made by the Company Environmental Review Group (CERG) or, by. the Plant Nuclear Safety Committee (PNSC)., Secondary reviews shall be made by the Company Nuclear Review Board (CNRB)..

,t J 2 18 5.3.2 PNSC and CNRB membership and. responsibilities are described in Appendix.A, Technical Specifications. 4 i ~

5.3.3 Com an Environmental. Review.Grou CERG A. Function The Company Environmental Review Group (CERG) shall function to advise the Vice President, Environmental Department on 'all matters related to environmental quality.

R

1. Manager, Environmental Engineering - Chairman

2. Manager, Environmental Affairs 5-1.

EXECUTIVE V1CE PRESIDENT VICE PRESIDENT VICE PRESIDENT GROUP COMPANY NUCLEAR ENVIRONMENTAL NUCLEAR VICE REVIEW BOARD DEPARTMENT AFFAIRS PRESIDENT ENVIRONMENTAL MANAGER, VICE PRESIDENT I EW I QUALITY POWER l I I ASSURANCE RESOURCES I I I I I I I I I I

I l l I I MANAGER OF I I

I POWER RESOURCES- t I I I NUCLEAR I I

I I I I I I I I I I I I I PLANT I I I I MANAGER I

I I PLANT NUCLEAR I SAFETY COMMITTEE I

I 1

I I

FLORIDA POWER 6 LIGHT COMPANY CORPORATE ORGANIZATION - ENVIRONMENTAL AFFAIRS FIGURE 5.2-1 AUTHORITY COMMUNICATION

3. Senior Environmental Engineer, Environmental Engineering

4. Power Resources Test Group Supervisor

5. Environmental Department Life Scientist

6. Environmental Department Senior Prospect Coordinator

7. Plant Supervisor (Plant Involved)

8. Power Resources Administrative Assistant Nuclear C. Alternates Alternate members shall be appointed in writing by the CERG Chairman. No more than two alternates shall participate in CERG activities at any one time.

D. Meetin Fre uenc The CERG shall meet at least semiannually and as convened by the CERG Chairman or designated acting Chairman.

E. Quorum A quorum of the CERG shall consist of the Chairman, or designated acting Chairman and three members including alternates.

F. Res onsibilities

1. Review of all Environmental Department procedures required by Environmental Technical Specifications and changes thereto. Review of any proposed procedures or changes thereto as determined by the Plant Manager to aff'ect the environment.

Review results of the environmental monitoring programs prior to their submittal to the NRC.

Review of all proposed test and. experiments as determined by the Plant Manager to affect the environment.

J 4~ Review of all proposed changes,to the Environmental Technical Specifications.

5. Review of all proposed changes or modifications to plant systems or equipment as determined by the Plant Manager to affect the environment.

6. Review of investigation of violations of the Environmental Technical Specifications.

5-4

7. Performance of special reviews and investigations and reports thereon as required by the Chairman of the Company Nuclear Review Board.

G. ~Authot1t The Company" Environmental Review Group shall:

Recommend to the Vice President, Environmental Department, written approval or disapproval of the items considered under F.l through F.5 above.

H. Records The Company Environmental Review Group shall maintain written minutes of each meeting and copies shall be provided to the Vice President, Environmental Department, Vice President, Power Resources, and the Chairman of the Company Nuclear Review Board.

5.3.4 Periodic audits concerning the implementation of the ETS'shall be made as provided in the Quality Assurance Manual.

t 5.4 5.4.1 5.4.2 Action to be Taken When a if a Limitin Condition is Exceeded Limiting Condition is exceeded, action shall be taken as permitted by the applicable specification until the condition can be met.

Exceeding a Limiting Condition shall .be investigated by, the Company Environmental Review Group or by the Plant Nuclear Safety Committee.

5.4.3 All reviews and actions taken, with reasons therefor, shall be recorded and maintained as part of the permanent records.

5.4.4 Each instance whereby a Limiting Condition is exceeded shall be reported to the Company Nuclear Review Board.

5.5.5 A report for each occurrence shall be prepared as specified in Section 5.6.2.

5.5 Procedures 5.5.1 Detailed written procedures, including applicable check lists and instructions, shall be prepared and followed for activities involved in carrying out the environmental technical specifications. Procedures shall include sampling, data recording and storage, instrument calibration, measurements and analyses, and actions to be taken when limits are exceeded. Testing frequency of any alarms shall be included.

5-5 5.5.2 Plant operating procedures shall include provisions to ensure that plant systems and components are operated in compliance with the environmental technical specifications.

5.6 Re ortin Re uirements Routine Re orts 5.6.1.a Annual Non-Radiolo ical Environmental Monitorin Re ort A report on the environmental surveillance programs for the previous 12 months of operation shall be submitted to the Director of the Regional Office of Inspection and Enforcement with a copy to the Director of the Office of Inspection and Enforcement as a separate document within 90 days after January 1 of each year. In the'event that some of the results are not available within the 90 day'eriod, the report shall be submitted noting and explaining the'issing results. The missing data shall be submitted as soon as possible in a 'supplementary report. The period of the first report shall begin with the date of initial criticality. The report shall include summaries and interpretations of t'he'results of the non-radiological environ-mental surveillance activities (Section 3.0) and the environmental monitoring programs required by Limiting Conditions for Operation. This should also

'include 'a comparison with preoperational studiep, operational controls (as appropriate), and previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.

If harmful effects= or evidence of irreversible damage are detected by the monitoring, the licensee shall provide an analysis of the problem and a proposed course of action to*alleviate the problem.

5.6.1.b Annual Radiolo ical Environmental Monitorin Re ort A report on the radiological environmental surveillance programs for the previous 12 months of operation shall be submitted to the Director of the NRC Regional Office-(with a copy to the 'Director, Office of Nuclear Reactor Regulation) as a separate document within 90 days after January 1 of each year. The period of the first report shall begin with the date of initial criticality. The reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities .for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environ-ment. The reports shall also include the results of land use censuses required by the specifications. If -harmful effects or evidence of irreversible damage are detected by the monitoring, the licensee shall provide an analysis of the problem and a proposed course of action to alleviate the problem.

Results of all radiological environmental samples taken shall be summarized on an annual basis in a format similar to that indicated in Table 5.6.1-F. In the event that some results are not available within the 90-day period; the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

TABLE 5.6.1-F FORMAT FOR ENVIRONMENTAL RADIOLOCIGAL MONITORING PROGRAM

SUMMARY

NAME OF FACILITY ST. LUCIE PLANT UNIT 1 DOCKET NO. 50-335 LOCATION OF FACILITY ST. LUCIE COUNTY FLORIDA REPORTING PERIOD Type and Lower Limit All"Indicator Location with Hi hest Annual Mean Number of Medium or Pathway Total Number of a Locationsb Control Locations Nonroutine Sampled of Analyses Detection Mean (f) Name Mean (f)b Mean (f)b Reported (Unit of Measurement) Performed (LLD) Rangeb Distance and Direction Rangeb Rangeb Measurements c Air Particulates 3

(pCi/m ) Gross 8 416 0.003 0.08 (200/312)

(0.05-2.0) 5 miles 340'.10 Middletown (5/52)

(0.08-2.0) 0.08 (8/104)

(0.05-1.40).

Y -Spec.32 137C 0.003 0.05 (4/24} Smithville (2/4) <LLD 140BBa 0.003 (0.03-0.13) 0.03 (2/24) 2.5 miles 270'.08 160'odunk (0.03-013) 0.05 (2/4) 0.02 (1/8)

(0.01-0.08) 4.0 miles (0.01-0.08)

Sr 40 0.002 <LLD <LLD Sr 40 0.0003 <LLD <LLD Fish pCi/kg (dry weight) y -Spec. 8 Cs 80 <LLD <LLD - 90 (1/4)

Cs 80 <LLD <LLD <LLD 60Co 80 120 (3/4) River Mile 35 See column 4 <LLD (90-200) Podunk River Nominal Lower Limit of Detection (LLD) as defined in HASL-300 (Rev. 8/74) pp. D-08-01, 02, 03 at the 952 compliance level.

b Mean and range based upon detection measurements only. Fraction of detectable measurements at specified locations is indicated in parentheses (f}.

Nonroutine reported measurements as defined in Section 5.6.2.b.

4'-7 5.6.l.c Semiannual Radioactive Effluent Reiease Re ort A report on the radioactive discharges (Regulatory Guide 1.21, Rev. 1, June 1974) released from the site during the previous 6 months of operation shall

~ include the following:

Analyses of Effluent, releases shall be summarized on a quarterly basis and reported in a format similar to Tables 5.6.1-A, 8, C, and D.

Supplemental information shall be included covering topics similar to those itemized in Data Sheet 5.6.1-1.

Abnormal releases should be handled as batch releases for accounting purposes.

Solid wastes shall be summarized on a quarterly basis and reported in a format similar to that of Table 5.6.2-E.

The following information should be reported for shipments of solid waste and irradiated fuel transported from the site during the report period:

1. The semiannual total quantity in cubic meters and the semiannual total radioactivity in curies for the categories or types of waste.

a. Spent resins, filter sludges, evaporator bottoms;

b. Dry compressible waste, contaminated equipment, etc.;

Co Irradiated components, control rods, etc.;

d. Other (furnish description).

2. An estimate of the total activity in'the categories of waste in 1, above.

3. The disposition of solid waste shipments. (Identify the number of shipments, the mode of transport, and the destination.)

4. The disposition of irradiated fuel shipments. (Identify the number of shipments, the mode of transport, and the destination.) ~

5.6.2 Non-Routine Re orts 5.6.2.a Non-Radioactive .Effluent. Re orts h

A report shall be submitted in the event that: a) a limiting condition is exceeded (as specified in Section 2.0 Limiting Conditions), or an unusual or important event occurs that causes a significant environmental impact, that affects potential environmental impact from plant operation, or that has high public or potential public interest concerning environmental impact from plant operation. Reports shall be submitted under one of the report schedules described below.

5-8 TABLE 5. 6. 1-1 EFFLUENT AND WASTE DISPOSAL Supplemental Information Facility License

1. Regulatory Limits a~ Fission and activation gases:

b. Iodines:

ci Particulates, half-lives .8 days:

d ~ Liquid effluents:

2. Maximum Permissible Concentrations Provide the MPCs used in determining allowable release or concentrations.

a ~ Fission and activation gases:

b. Iodines:

co Particulates, half-lives >8 days:

d. Liquid effluents:

3. Average Energy Provide the average(E)of the radionuclide mixture in releases of fission and activation gases, if applicable.

4. Measurements and Approximations of Total Radioactivity Provide the methods used to measure or approximate the total radioactivity in effluents and the methods used to determine radionuclide composition.

a 0 Fission and activation .gases:

b. Iodines:

c ~ Particulates:

d. Liquide effluents:

5. Batch Releases Provide the following information relating to batch releases of radioactive materials in liquid and gaseous effluents.

5-9

a. Liquid

1. Number of batch releases:

2. Total time period or batch releases:

3. Maximum time period for a batch release:

4. Average time period for batch releases:

5. Minimum time period for a batch release:

6., Average stream flow during periods of release of effluent into a flowing stream:

b. Gaseous

1. Number of batch releases:

2 ~ Total time period for batch releases:

3 ~ Maximum time period for a batch release:

4 ~ Average time period for 'batch releases:

5. Minimum time period for a batch release:

6. Abnormal Releases

a. Liquid,

l. Number of, releases:

2. Total activity released:

b. Gaseous

1. Number of releases:

2. Total activity released:

5-10 TABLE 5. 6. 1-A GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES Unit A. Fission 6 activation gases

1. Total release

2. Average release rate for, period, pCi/sec

3. Percent of Mchnical specification limit B. Iodine s

1. Tbtal iodine-131

2. Average release rate for period pci/sec

3. Percent of Technical specification limit C. Particulates

1. Particulates with hal f-lives 8 days Ci

2. Average release rate for period pCi/sec

3. Percent of Technical specification limit

4. Gross alpha radioactivity D. Tritium

1. 'Xbtal release

2. Average release rate for period pci/sec

5-11 TABLE 5 ~ 6 ~ 1-B GASEOUS EFFLUENTS

~

QONZZNUOUS NODE Nuclides Released Unit Qu Q~~r Quarter Quarter

1. Fission gases kr ton-85 Cx E E E kr ton-85m C1 E E E E kr ton-87 ~ E E E E kr ton-88 '

E E E E xenon-133 E E E xenon-135 C1 E E E xenon-135m E E xenon-138 E E Others (s eczf ) E E E E E ~ E E E E E E E un'.dentzfxed E, E E E Total for erxod Cx E E E

2. Iodines a.odzne-131 E E E moderne-133 E E E E moderne-135 E E E E Total for eriod E E E

3. Particulates strontium-89 E strontium-9 C E E

E E '-. E cesium-134 E E E E cesium-137 E E E E barxum-lanthanum-140 E E E E Others s ecx.f E E E' E E E E unxdentx.fied E E E E

5-12 TABLE 5.6.j:-.C LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES Unit Quarter A. Fission and'ctivation products

1. Total release (not umlucLLng trz.thun, Ci . E ses, al ha)

2. Average luted concentration 1ICi/ml . E during period

3. Percent of applicable limit. E B. Tritium

1. Total release Ci E 2.

dur'od Average <<h.luted concentratI.on uCi/ml E C. Dissolved and entrained gases

l. Total release Ci . E

2. Average dhluted concentratI.on .-,

durin iod IIC1/ml E E

3. Percent of applicable E E D. Gross alpha radioactivity

1. Total release E . E F. Volure of dilution water used during period liters E E

5-13 TABLE, 5 6 ~ 1-D LIQUXD EFFLUENTS Nuclzdes Released strontium-8 E E E E strantxun-90 E E E E cesium-134 E E E .E oe un-137 'E E E E odme-131 E cobalt-58 E E E cobalt-60 E E E iran-S9 Ci E E E zinc-59 E Ci E E E manganese-54 E E E E

chxmo.un-51 E E E E

z 'un'. un-95 E ~ E E' Ci ~ ~ ,E E E. ~ E technetium-99m E E' E E barium-lantana@-140 E E E cerium-141 E E E E Other (specify)

E E Ci E E E E E E E E E E E E E E E ustified E E E E . E E ~ E xenon-133 E E . F E x engr E E E

5-14 TABLE 5.6.1-E SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSHE FOR BURDEN, OR DISPOSAL(Not irradiated fuel)

1. Type of waste Unit 6~nth Period

a. Spent resins, Q.lter sludges, evaporator m E bottcms, etc. Ci E

b. Dry oanpressxhle waste, contanunated m E etc. Ci E

c. Irradiated canponents, control m3 rods etc. Ci

d. Other (describe) E E

2. SOLID 0RSTE DISPOSITION Nunber of Shi ts Rxh of Trans rtation Destination B. IRRADIATED FUEL SHIPMENTS (Disposition)

Nunber of Shi ts Nxh of Tran rtation Destination

5-15

1. Prompt Reports Those events requiring prompt reports shall be reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone, telegraph, or facsimile transmission to the Director of the Regional Office of Inspection and Enforcement and within 10 days by a written report to the Director of the Office of Inspection and Enforcement.

2. 30-Day Reports Those events not requiring prompt reports shall be reported within 30 days by a written report to the Director of the Regional Office of Inspection and Enforcement with a copy to the Director of the Office of Inspection and Enforcement.

The reporting schedule for reports concerning limiting conditions shall be reported on the 30-day schedule. Reports concerning unusual or important events shall be reported on the prompt schedule.

Written 10-day and 30-day reports and to the extent possible the preliminary telephone, telegraph, or facsimile reports shall: a) describe, analyze, and evaluate the occurrence, including extent and magnitude of the impact, b) describe the cause of the occurrence and c) indicate the corrective action (including any significant changes made in procedures) taken to preclude repetition of the occurrence and to prevent similar occurrences involving similar components or systems.

The significance of an unusual or apparently important event with regard to environmental impact may not be obvious or fully appreciated at the time of occurrence. In such cases, the NRC shall be informed promptly of changes in the assessment of the significance of the event and a corrected report shall be submitted as expeditiously as possible.

5.6.2.b Radioactive Effluent Re orts Li uid Radioactive Wastes Re ort If the cummulative releases of radioactive materials in liquid effluents, excluding tritium and dissolved gases, should exceed one-half the design objective annual quantity during any calendar quarter, the licensee shall make an investigation to identify the causes of such releases and define and initiate a program of action to reduce such releases to the design objective levels. A written report of these actions shall be submitted to the NRC within 30 days from the end of the quarter during which the release occurred.

Gaseous Radioactive Wastes Re ort Should the conditions a), b), or c) listed below exist, the licensee shall, make an investigation to identify the causes of the release rates and define and

5-16 and initiate a program of action to reduce the release rates to design ob)ective levels. A written report of these actions shall be submitted to the NRC within 30 days from the end of the quarter during which the releases occurred.

a. If the average release rate of noble gases for the site during any calendar quarter exceeds one-half the design objective annual quantity.

b. If the average release rate per site of all radioiodines and radioactive materials in particulate form with half-lives greater than eight days during any calendar quarter exceeds one-half the design objective annual quantity.

c. If the amount of iodine-131 released during any calendar quarter is greater than 0.5 Ci/reactor.

Un lanned or Uncontrolled. Release Re ort Any unplanned or uncontrolled offsite release of radioactive materials in excess of 0.5 curie in liquid or in excess of 5 curies of noble gases or 0.02 curie of radioiodines in gaseous form requires notification. This notification must be made by a written report within 30 days to the NRC.

The report shall describe the event, identify the causes of the unplanned or uncontrolled release and report actions taken to prevent recurrence.

5.6.2.c Radiolo ical Environmental Surveillance Re orts If a confirmed measured level of radioactivity in an environmental medium exceeds ten times the control station value, a written report shall be submitted to the Director of the NRC Regional Office (with a copy to-the Director, Office of Nuclear Reactor Regulation) within 10 days after confirmation of the validity of the measured level. Confirmation shall be completed at the earliest time consistent with the analysis, but in any case, within 30 days. This report shall include an evaluation of any release conditions, environmental factors, or other aspects necessary to explain the anomalous result.

5.6.3 Chan es in Environmental Technical. S ecifications Request for changes in environmental technical specifications shall be sub-mitted to the Director of Nuclear Reactor Regulation for review and authorization. The request shall include an evaluation of the environmental impact of the proposed change.

5.7 Records Retention, Records and logs relative to the following areas shall be made and retained for the life of the plant:

a. Records and drawings detailing plant design changes and modifications made to systems and equipment as described in 5.3.3.F.S.

b. Records of all environmental surveillance data.

c. Records to demonstrate compliance with the limiting conditions in Section 2.

All other records and logs relating to the environmental technical specifica-tions shall be retained for five years following logging or recording. These shall include (but are not limited to) the following:

a. Details or any abnormal operating conditions having an effect on the environment, and actions taken to correct those conditions.

b. Maintenance activities to environment monitoring equipment, including but not limited to:

1) routine maintenance and component replacement,

2) equipment failures, ll

3) replacement of principal items of equipment.

c. Records of radioactivity levels in liquid and gaseous wastes released to the environment.

d. All reviews, including actions taken and reasons therefor, required in Sections 2, 3, and 4 of this specification.

4~ 4

SPECIAL CONDITIONS Li t Screen to Minimize Turtle Disorientation Australian pine or other suitable plants (i.e., native vegetation such as live oak, native figs, wild tamarind, and others) shall be planted and maintained as a light screen, along the beach dune line bordering the plant property to minimize turtle disorientation.

6-1

4 r

ML18096B437

Text

SUMMARY

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