ML17228B061

Amends 134 & 73 to Licenses DPR-67 & NPF-73,respectively, Correcting Editorial Errors,Removing Deleted Pages & Clarifying Confusing Statement & Eliminating Inconsistencies

ML17228B061
Person / Time
Site:	Saint Lucie
Issue date:	03/15/1995
From:	Matthews D NRC (Affiliation Not Assigned)
To:
Shared Package
ML17228B062	List: ML17228B061 ML17309A779
References
NUDOCS 9503170316
Download: ML17228B061 (46)

Text

<<00 Wp0 Cy t

O I

0 I

+0

~O

++*++

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON~ 0 C 2055$ 4001 LORIDA POWER 8I LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE P ANT UNIT NO.

1

~

MENDMENT TO FAC I

OPERAT NG LICENSE Amendment No.

134 License No.

DPR-67 The Nuclear Regulatory Commission (the Commission) has found that:

A.

B.

C.

D.

E.

The application for amendment by Florida Power

& Light Company, et al. (the licensee),

dated October 27,

1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

95031703ib

'3I50315

! t

)

PDR ADODK 05000335 P

PDR

a I

~ >

Tj

,1

2.

3.

Accordingly, Facility Operating License No.

DPR-67 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.(2) to read as follows:

(2)

Technical S eci ications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 134, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

@~M~

David B. Matthews, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 15, 1995

e Wl

TTACHMENT TO LICENSE AMENDMENT NO.

134 TO FACI ITY OPERATING LICENSE NO.

DPR-67 DOCKET NO. 50-335 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

emove Pa es IaIII IV V

VI VII IX X

XI XII XIII XIV XV 8 2-7 3/4 2-2 3/4 3-22 3/4 3-25 3/4 3-37 3/4 3-38 thru 3/4 3-40 3/4 3-44 3/4 3-45 3/4 3-46 thru 3/4 3-49 3/4 3-51 3/4 3-52 3/4 3-53 thru 3/4 3-56 3/4'-27 3/4 4-28 thru 3/4 4-57 3/4 7-10 3/4 7-11 and 3/4 7-12 3/4 7-32 thru 3/4 7-39a

'/4 11-1 3/4 11-2 thru 3/4 11-13 3/4 11-16 and 3/4 11-17 3/4 12-1 thru 3/4 12-12 8 3/4 7-1 8 3/4 ll-l thru 8 3/4 11-3 8 3/4 12-1 and 8 3/4 12-2 sert Pa es IaIII IV V

VI VII Ix X

XI XII XIII XIV XV 8 2-7 3/4 2-2 3/4 3-22 3/4 3-25 3/4 3-37 3/4 3-44 3/4 3-45 3/4 3-51 3/4 3-52 3/4 4-27 3/4 7-10 3/4 11-1 8 3/4 7-1 8 3/4 ll-l

/j I

tW.

INDEX DEFINITIONS SECTION g

PAGE

1. 0 DEFINITIONS 1

~ 1 Action.................

~

~

~

~

~

~

~

~ ~

~

~ ~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

1.2 Axial Shape Index............................................

1-1 1.3 Azimuthal Power Tilt.........................................

1-1 1.4 Channel Calibration..........................................

1-1 1.5 Channel Check...'.........

~

~

~

~

~

~

~

~

~

1

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

1 1

1.6 Channel Functional Test......................................

1-2 1.7 Containment Vessel Integrity.................................

1-2 1.8 Controlled Leakage....................................

.... 1-2 1.9 Core Alteration................................

~

~ ~

~ ~

~

~

~ ~

~

~

~

1 2

1.10

'Dose Equivalent I-131........................................

1-3 1.11 E Average Disintegration Energy..............................

1-3 1.12 Engineered Safety Features Response Time.....................

1-3 1.13 Frequency Notation.

~ ~

~

~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~

~ ~

~

~

~

~

1 3

1.14 Gaseous Radwaste Treatment System............................

1-3 1.15 Identified Leakage...........................................

1-4 1.16 Low Temperature RCS Overpressure Protection Range............

1-4 1.17 Member(s) of the Public......'................................

1-4 1.18 Offsite Dose Calculation Manual (ODCM)........

~

~

~

~

~

1 4

1.19 Operable - Operability.......................................

1-5 1.20 Operational Mode - Mode.....................

1-5 1.21 Physics Tests.......

~

~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~ ~

~

~

~

~

~

~

5 1.22 Pressure Boundary Leakage.........

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~ ~

~

1 5

ST.

LUGIE - UNIT 1

Amendment No. 27 82.Q 69

INOEX DEF'INITIONS SECTION PAGE 1.24 Purge - Purglngo.........oo.oe.o........o..ooooooo....

1.23 Process Control Program (PCP).............;..............

1-5 1-5 1.25 Rated Thermal Power.......................................

1-6 1.26 Reactor Trip System Response Time........................

1-6 1.27 eportable Event.........................................

R 1-6 1.28 Shield Building Integrity................................

1-6 1.29 S hutdown Margin..........................................

1.30 S t

>te Boundary.................................

1-6 1.31 ource Check.............................................

S 1-6 1.32 Staggered Test Bas>s.....................................

1-7 1.33 hermal Power............................................

T 1-7 1.34 Unidentified Leakage.....................................

1-7 1.35 Unrestricted Area........................................

1-7 1.36 Unrodded Integrated Radial Peaking Factor F...........

r'

"-.37 DELETED 1-7 1-7 TABLE 1.1 Frequency Notation....,.......,...,........,.....,....

TABLE 1.2 Operational Nodes.....,...................,....,.......

1-9 ST.

LUCIE UNIT 1 Ia A

d N

. SO-,SQ;M,Me.>>

3/4. 1.3 MOVABLE CONTROL ASSEMBLIES Full Length CEA Position Position Indicator Channels I

CEA Drop Time Shutdown CEA Insertion Limit Regulating CEA Insertion Limits

~

~

INDEX LIMITING CONDITIONS FOR OPERATION ANO SURVEILLANCE RE UIREMENTS SECTION 3 4.0 APPLICABILITY 3 4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 BORATION CONTROL Shutdown Margin - T

) 200'F I

Shutdown Margin - T,, a 200'F Boron Dilution Moderator Temperature Coefficient Minimum Temperature for Criticality.

3/4.1. 2 BORATION SYSTEMS Flow Paths

- Shutdown Flow Paths

- Operating Charging Pumps

- Shutdown Charging Pumps

- Operating Boric Acid Pumps

- Shutdown Boric Acid Pumps

- Operating Borated Water Sources

- Shutdown Borated Water Sources

- Operating PAGE 3/4 0-1 3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-4 3/4 1-5 3/4 1-7 3/4 1-8 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-13 3/4 1-14 3/4 1-15 3/4 1-16 3/4 1-18 3/4 1-20 3/4 1-20 3/4 1-24 3/4 1-26 3/4 1-27 3/4 1-28 ST.

LUCIE -

UNIT 1

Amendment No. 77, +Rh l34

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3 4.2 POW R DISTRIBUTION LIMITS PAGS 3/4.2.1 LINEAR HEAT RATE...................................

3/4 2-1 3/4.2.2 DELETED 3/4 2-6 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR F........

3/4 2-9 3/4.2.4 AZIMUTHAL POWER TILT T....................

g 3/4.2.5 DNB PARAMETERS...............................

...... 3/4 2-11

...... 3/4 2-13 3 4.3 INSTRU ENTATION 3/4.3. 1 REACTOR PROTECTIVE INSTRUMENTATION................. 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION INSTRUMENTATION.....................

SYSTEM

~

~

~ ~

~ ~

~

~

~

~

~

~

~

~

~ 3/4 3 9 3/4.3.3 MONITORING INSTRUMENTATION......................... 3/4 3-21 Radiation Monitoring...............................

Incore Detectors...........................

Seismic Instrumentation............................

Meteorological Instrumentation.....................

Remote Shutdown Instrumentation....................

3/4 3-21 3/4 3-25 3/4 3-27 3/4 3-30 3/4 3-33 Accident Monitoring Instrumentation...............

3/4 3-41 Explosive Gas Monitoring Instrumentation...........

3/4 3-50 3 4.4 REACTOR COOLANT SYSTE 3/4.4. 1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION...... 3/4 4-1 3/4.4.2 SAFETY VALVES SHUTDOWN...........................

3/4 4-2 3/4.4.3 SAFETY VALVES OPERATING........................ 3/4 4-3 ST.

LUCIE UNIT 1 IV

l' t

INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4.4.4 3/4.4.5 PRESSURIZERe

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

STEAM GENERATORS.......................................

PAGE I

3/4 4-4 3/4 4-5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.........................

3/4 4-12 3/4.4.7 Leakage Detection Systems.............

Reactor Coolant System Leakage........

HEMISTRY.............;................................

C 3/4.4.8 SPECIFIC ACTIVITY.................................

~....

3/4 4-12 3/4 4-14 3/4 4-15 3/4 4-17 3/4.4.9 PRESSURE/TEMPERATURE LIMITS............................

3/4 4-21 Reactor Coolant System................

Pressurizer...........................

3/4.4.10 STRUCTURAL INTEGRITY....................... -...........

3/4 4-2T'/44-2g 3/4 4-26 ASME Code Class 1, 2, and 3 Components.................

3/4 4-26 3/4.4.11.

3/4.4.12 ELETED ~ ~ ~ ~

~ ~ ~ ~

~ ~ ~ ~ ~ ~

~

~ ~

~ ~

~

~ ~ ~ ~

~

~

~

~

~

~ ~

~ ~ ~

~ ~

~

~

~

~

~

~ ~ ~ ~

~

~ ~

D PORV BLOCK VALVES......................................

3/4 4-56 3/4 4-58 3/4.4.13 POWER OPERATED RELIEF VALVES...........................

3/4 4-59 3/4.4.14 REACTOR COOLANT PUMP - STARTING........................

3/4 4-60 3/4.4.15 REACTOR COOLANT SYSTEM VENTS...........................

3/4 4-61 3 4.5 EMERGENCY CORE COOLING SYSTEMS ECCS 3/4.5.1 3/4.5.2 3/4.5.3 3/4.5.4 SAFETY INJECTION TANKS.................................

3/4 5-1 ECCS SUBSYSTEMS Tavg > 325 Fa ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~ 3/4 5 3 ECCS SUBSYSTEMS Tavg ( 325 F ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~

~ ~ ~

~

~ ~

~

o 3/4 5

7 REFUELING WATER TANK...................................

3/4 5-8 ST.

LUCIE-UNIT 1 Amendment No. gg, gg,,~

134

INDEX I'IMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.6.2 Asr Temperature............."."""....""

Containment Vessel Structural Integrity......

DEPRESSURI'ZATION.AND COOLING SYSTEMS.........

Containment Spray and Cool.ing Systems................

Spr ay Additive System........................"...".

3/4. 6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT VESSEL...................................

Containment Vessel Integrity.........................

C t onta>nment Leakage..................................

Containment Air Locks............................ "..

Inter nal Pressure.......... "....... ".".".". "."

3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-10 3/4 6-12 3/4 6-13 3/4 6-14 3/4 6-15 3/4 6-15 3/4 6-16a 3/4.6.3 3/4.6.4 3/4.6.5 3/4.6.6 CONTAINMENT ISOLATION VALVES.........................

COMBUSTIBLE GAS CONTROL..............................

Hydrogen Analyzers......... "........ """.

Electric Hydrogen Recombiners -

W...........

VACUUM RELIEF VALVES......."..................

SECONDARY CONTAINMENT................................

Shield Building Ventilation System...........

Shield Building Integrity....................

Shield Building Structural Integrity.........

3/4 6-18 3/4 6-23 3/4 6-23 3/4 6-24 3/4 6-26 3/4 6-27 3/4 6-27 3/4 6-30 3/4 6-31 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE................................

afety Valves.................................

S 3/4 3/4 7-1 7-1 3/4 3/4 A t

~

~

ctlvltyo

~

~ ~ ~ ~

~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~

~ ~

~

~ ~ ~ ~

~

~ ~

~ ~ ~

~ ~ ~ ~ ~, ~ ~ ~

~

~

~

~

~

Main Steam Line Isolation Valves..............

3/4 3/4 ST.

LUCIE - UNIT 1

VI Amendment Auxiliary Feedwater System............... "".

Condensate Storage Tank.......................

~ ~

~ ~ ~ ~ ~

7-4 7-6 7-7 I

7-9 ND Ãf. li1.IN~.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 5

PAGE 3/7.7.2 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 3/4.7.6 ULTIMATE HEAT SINK....................................

FLOOD PROTECTIONt ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~

~ ~ ~

~ ~

~

~ ~

~ ~ ~ ~

~

3/4 7-18 3/4 7-19 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATIONSYSTEM.............

3/4 7-20 3/4.7.8 ECCS AREA VENTILATIONSYSTEM...........................3/4 7-24 3/4.7.9 SEALED SOURCE CONTAMINATION........................... 3/4 7-27 3/4.?;10 SNUBBERS..............................................

3/4 7 3/4 '

ELECTRICAL POWER SYSTEMS 3/4.8.1

~ C ~

SOURCES

~ ~

~ ~ ~ ~

~ ~ ~

~

~

~ ~

~

~

~

~

~ ~

~ ~

~

~

~ ~ ~ t

~

~ ~

~

~

~

~

~

~

~

~

~

~

~

~ 3/4 8 1

A 0perattng.............................................

3/4 8 1

Shutdown..............................................

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.

~

~ ~ ~

~ ~ ~ ~ ~ o ~

~ i ~ ~ ~

~

~

~

~ ~ 3/4 8 10 D.C. Distribution - Shutdown..........................

3/4 8-13 ST.

LUCIE - UNIT 1

VII Amendment No. g$, Q, Q, 134

0I a ~<<y A F

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3 4.9 REFUELING OPERATIONS PAGE 3/4.9.1 3/4.9.2 BORON CONCENTRATION....................................

3/4 9-1 INSTRUMENTATION........................................,3/4 9-2 3/4.9.3 ECAY TIME.

D 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 SHUTDOWN COOLING AND COOLANT CIRCULATION...............

3/4.9. 9 CONTAINMENT ISOLATION SYSTEM...........................

3/4 9-8.

3/4 9 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 STORAGE............

3/4 9-12 3/4.9.13 SPENT FUEL CASK CRANE...

3/4 9-15 3/4.9.14 DECAY TIME - STORAGE POOL..............................

3/4 9-16 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIt(...

3/4 10-1 3/4.10.2 POWER DISTRIBUTION GROUP HEIGHT, INSERTION AND LIMITS

~

~ ~

~

~ ~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~ 3/4 10 2

3/4.10.3 3/4.10.4 fLETED i ~

~

~ ~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~ ~

~

~

~

~ o

~

~ ~ ~

~ ~

~

~

~

~

~

~

~

~

D ELETED o

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~ ~

~

~

~

~

~

~

~

~

~

~

~

D 3/4 10-3 3/4 10-4 3/4.10.5 CENTER CEA MISALIGNMENT................................

3/4 10-5 ST.

LUCIE - UNIT 1

VIII Amendment t<o. A, 2A. 25, 87. 6 0

I

BASES INDEX SECTION PAGE 3/4.0 APPLICABILITY............................ ~.............

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 3/4.1.2 BORATION CONTROL....................................

B 3/4 1-1 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....................................

8 3/4 2-1 3/4.2.2 DELETED

~ ~

~

~

~ ~

~

~

~

~

~

~ ~

~

~ ~

~

~

B 3/4 2 1

3/4.2.3 3/4.2.4 3/4.2. 5 DNB PARAMETERS B 3/4 2-2 TOTAL INTEGRATED RADIAL PEAKING FACTOR -

F.........

B 3/4 2-1 r

AZIMUTHAL POWEP, TILT................................

B 3/4 2-1 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES INSTRUMENTATION............................

B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION......................

B 3/4 3-1 ST.

LUCIE - UNIT 1

IX Amendment No. 27, gg,134

BASES INDEX SECTION 3 4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION....

3/4.4.2 and 3/4.4.3 SAFETY VALVES........................

N

~ ~ ~

~

~ ~

B

~ ~

~

~

~ e 8

PAGE 3/4 4-1

'/4 4-1 3/4.4.4 PRESSURIZER............................................

B 3/4 4-2 3/4.4.5 STEAM GENERATORS......................................

B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.........................

B 3/4 4-4 3/4.4.7 CHEMISTRY.............................................

B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY......................................

B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS............................

B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY...................................

B 3/4 4-12 3/4.4.11 DELETED................................................

8 3/4 4-13 3/4.4.12 PORV BLOCK VALVES.........'.............................

B 3/4 4-14 3/4.4.13 POWER OPERATED RELIEF VALVES and 3/4.4.14 REACTOR COOLANT PUMP - STARTING................................

B 3/4 4-15 3/4.4.15 REACTOR COOLANT SYSTEM VENTS...........................

B 3/4 4-15 3 4.5 EMERGENCY CORE COOLING SYSTEMS ECCS 3/4.5.1 SAFETY INECTION TANKS............;...................

B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS.............................

B 3/4 5-1 3/4.5.4 REFUELING WATER TANK (RWT)............................

B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6. 1 CONTAINMENT VESSEL.....................................

B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................

B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES...........................

8 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL................................

B 3/4 6-3 3/4.6.5 VACUUM RELIEF VALVES...................................

B 3/4 6-4 3/4.6.6 SECONDARY CONTAINMENT..................................

B 3/4 6-4 ST.

LUCIE - UNIT 1

X Amendment No. g7,

$5, Jig, H, PP>

INDEX BASES SECTION 3 4.7 PLANT SYSTEMS PAGE 3/4.7.1 3/4.7.2 TURBINE CYCLE...................................,...,.

STEAM'ENERATOR PRESSURE/TEMPERATURE LIMITATIONi 'i ~

~

~

~ ~ ~ ~ ~

~

~

~

~ ~

~

~ ~ ~

~

~

~

~

~

~ ~

~

~ ~

~

~ ~ i'

~

~

~ ~ ~ ~

~ ~

8 3/4 7-1 8 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................

8 3/4 7-4 3/4.7.4 INTAKE COOLING WATER SYSTEM...........................

8 3/4 7-4 3/4.7.5 3/4.7.6 ULTIMATE HEAT SINK....................................

FLOOD PROTECTION......................................

8 3/4 7-4 8 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATIONSYSTEM.............

3/4.7.8 ECCS AREA VENTILATIONSYSTEM..........................

8 3/4 7-4.

V 8 3/4 7-5 3/4.7.9 SEALED SOURCE CONTAMINATION...........................

8 3/4 7-5 3/4.7.10 SNUBBERS............

~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~

~ ~

~

~

~

8 3/4 7-5 3 4.8 ELECTRICAL POWER SYSTEMS................................

8 3/4 S-l 3 4.9 REFUELING OPERATIONS 3/4.9.1 3/4.9.2 3/4.9.3 BORON CONCENTRATION...................................

INSTRUMENTATION.......................................

DECAY TIME............

8 3/4 9-1 8 3/4 9-1 8 3/4 9-1 3/4.9.4 CONTAINMENT PENETRATIONS..............................

8 3/4 9-1 3/4.9.5 COMMUNICATIONS....'....................................

8 3/4 9-1 3/4.9.6 MANIPULATOR CRANE OPERABILITY..............;.....

8 3/4 9-1 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING............

8 3/4 9-2 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION..............

8 3/4 9-2 ST.

LUCIE - UNIT 1

XI Amendment No. g$, Q,g g, 134

BASES INDEX SECTION e

'PAGE 3/4.9.9 CONTAINMENT ISOLATION SYSTEM........................

B 3/4 9-2 3/4.9.10 and 3/4.9.11 MATER LEVEL - REACTOR VESSEL AND STORAGE POOL WATER LEVEL.................-..........

B 3/4 9>>2, 3/4'.9.12 3/4.9.13 3/4.9.14 FUEL POOL VENTILATION SYSTEM - FUEL STORAGE SPENT FUEL CASK CRANE.......................

DECAY TINE - STORAGE POOL...................

B 3/4 9-3 B 3/4 9-3 B 3/4 9-3 3 4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 3/4.10.2 SHUTDOWN MARGIN..................................,..

B 3/4 10-1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS

~ ~ ~

~

~

~ ~ ~

~ ~

~ ~ ~ ~

~ ~ ~

~ e ~ ~ ~

~

~

~ ~ ~ ~

~ ~ ~ ~ ~

~ ~

~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~

B 3/4 1 0 1

3/4.10.3 DELETED 3/4.10.4 DELETED 3/4.10.5 CENTER CEA MISALIGNMENT........................ "...

B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.-1 DELETED B 3/4 11-1 3/4.11.2.5 EXPLOSIVE GAS MIXTURE............................

B 3/4 11-4 3/4.11.2.6 GAS STORAGE TANKS..................................

B 3(4 11<<5 ST.

LUCIE - UNIT 1

XII Amendment No. b. 22,/27>> f ) I~<

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Excluslon Area............................................

5-1 Low Population Zone.....

Flood Control...........

~ ~

~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~

~ ~

~ ~

~ ~ ~ ~ ~ ~

~ ~

5-1

~

~ ~

~

~ ~

~ ~

~

~

~

~

~ ~

~

~ ~

~

~ ~ ~

~

~

~

~

~

~

~ ~

~

~ ~ ~ ~

5-1

5. 2 CONTAINMENT Conflguratl on......... " ~ ~.. " ~

~

Design Pressure and Temperature.

~

~

~ ~ ~ ~ ~ ~

~

~

~

~

~ ~

~

~

~

~

~

~

~ ~

~

~ ~ ~

~

~ ~

~

~

~

~ ~ ~ ~

~

~

~

~

~

~

~

~

~

~ ~

~

~ ~

~

~

5-4 Penetratlons..............................

~.-...

~.

~. "" ~ ~

5-4 5.3 REACTOR CORE Fuel Assemblies.........................................

Control Element Assemblies............'....................

5-5 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature...........................

5-5 Vol Ume ~ ~

~ ~

~

~ ~

~ ~

~

~ ~

~

~

~ ~ ~

~ ~

~ ~

~

~ ~

~

~

~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~

~ ~

~ ~

~

~

~ ~

~

~

~ ~

5.5 EMERGENCY CORE COOLING SYSTEMS............................

5-5 5.6 FUEL STORAGE C

~ ~ ~

i J.

rltlcalltgooo.o'oo

~ eo

~ soo

~ o.oo

~ oooooooo.eooieoo.oooeoooeoo 5-5 Dra l ilage

~

~ ~ ~

~

~ ~

~

~ ~ ~ ~

~

~

~ ~

~ ~ ~ ~

~

~ ~

~

~ ~

~ ~

~

~ ~ ~ ~ ~

~ ~

~ ~ ~

~ ~

~

~

~

~

~ ~

~

~ ~

5.7 SEISMIC CLASSIFICATION....................................

5-6 5.8 METEOROLOGICAL TOWER LOCATION.............................

5-6a 5.9 COMPONENT CYCLE OR TRANSIENT LIMITS.......................

5-6a T.

LUCIE - UNIT 1

XIII Amendment No. 134

IIIDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY...............................................

6-1 6.2 ORGANIZATION 6.2.1 ONSITE AND OFFSITE ORGANIZATION.............................

6-1 6.2.2 NIT STAFFoo ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~

U 6-2 6.2.3 SHIFT TECHNICAL ADVISOR.....................................

6-5

6. 3 UNIT STAFF UALIFICATIONS....

~

~

~

~

~

~

~

~

~

~

~

~ ~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

6 6 6.4 RAINING ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~ ~

~

~

~ ~

T 6-6 6.5 REVIEW AND AUDIT 6.5.1 FACILITY REVIEW GROUP unction....................................................

F omposition.................................

C 6-6 lternates..................................................

A 6-7' t

eetlng Frequency...........................................

6-7 uorum.....................................

Q 6-7 Responsibilities....

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~

~

6 7

Authorltyo ~

~

~

~

~

~

~

~

~

~

~

~

~

~

6-8 ecords..............

~o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

R 6-8 6.5.2 COMPANY NUCLEAR REVIEW BOARD FUnction ~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~ ~

6-9 Compos~tiont

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~

~

~ 6-9 lternates..................................................

A 6-9 onsultantso

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~

~

~

C 6-10 Meeting Frequency......

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

Quorum.

~

~

~

~ ~

~

~

~

~

~

~

~ I ~

~

~

~ ~

~

~

~

~

~

~

~

~ ~ ~

~

~

~ ~ I ~

~

~

~ 6-10 Revsewt

~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~ 6-10 Audits........

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~ ~

1

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~ ~

~

~ ~ ~

~

~

~ 6-11 uthorityt ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~

~

~ ~

~

A ecords......;..............................................

R 6-12 6-12 ST.

LUCIE - UNIT 1

XIV Amendment No. PP,N~ 134

INDEX ADMINISTRATIVE CONTROLS SECTION PAG 6.6 R

PORTABLE EVENT ACTION.................................

6-12 6.7 SAFETY LIMITVIOLATION..................................

6-12 6.8 PROCEDURES 0 PROGRAMS.................................

6-13 6.9 REPORTING R

UIREMENTS 6.9. 1 ROUTINE REPORTS.......................................

6-15b Startup Report.............................

Annual Reports.............................

Monthly Operating Reports................'..

Annual Radioactive Effluent Release Report.

Annual Radiological Environmental Operating 6-15 b

~

~

~

~

~

~

~

~

~

~

~

6 16 6-16a

~ ~

~ ~ ~ ~

~ ~

~ ~

~

6 17 Report....

6.-18 6.9.2 SPECIAL REPORTS.......................................

6-19

6. 10 RECORD RETENTION........................................

6-20 6.11 RADIATION PROTECTION PROGRA............................

6-21 6.12 HIG RADIATION AREA........................-.............

6-22

6. 13 PROCESS CONTROL PROGRAM.................................

6-23 6.14 OFFSITE DOSE CALCULATION MANUAL...........

~

~

~

~ ~

~ ~

~

6 23 ST.

LUGIE UNIT 1 XV d

d t d tdd;dttdd 134

P 4

'4'A I>

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 the'NBR limit.

The trip is initiated whenever the reactor coolant system pressure signal drops below either 1887 psia or a computed value as described below, whichever is higher.

The computed value is a function of the higher of bT 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 include appropriate allowances for equipment response time, calculational and measurement uncer-

tainties, and processing error.

A further allowance is included I

to compensate for the time delay associated with providing effective termina-tion of the occurrence that exhibits the most rapid decrease in margin to the DNBR limit.

As metric Steam Generator Transient Protective Tri Function ASGTPTF The ASGTPTF consists of Steam Generator pressure inputs to the TM/LP calculator, which causes a reactor trip when the difference in pressure between the two stdam generators exceeds the trip setpoint.

The ASGTPTF is designed to provide a reactor trip for those events associated with secondary system mal-functions which result in asymmetric primary loop coolant temperatures.

The most limiting event is the loss of load to one steam generator caused by a single main steam isolation valve closure.

The equipment trip setpoint and allowable values are calculated to account for instrument uncertainties, and will ensure a trip at or before reaching the analysis setpoint.

ST.

LUCIE - UNIT 1

8 2-7 Amendment No. gg, Pg, 44/. P. 134

LIMITING SAFETY SYSTEM SETTINGS BASES Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% 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 ensuing 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 Protec-tion System.

Rate of Chan e of Power-Hi h

The Rate of Change of Power-High trip is provided to protect the core'uring startup operations and its use serves as a backup to the administra-tively enforced startup rate limit.

Its trip setpoint does not correspond'o 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

B 2-8 Amendment No. 4 8

POWER DISTRIBUTION L I S

SURVEILLANCE REQUIREMENTS (Continued)

C.

Verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2, where 100 percent of maximum allowable power represents the maximum THERMAL POWER allowed by the following expression:

I MxN where:

1.

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

2.

N is the maximum as determined by allowable fraction of RATED THERMAL POWER the F

curve of Figure 3.2-3.

T r

4.2. 1.4 Incore Detector Monitorin S stems' The incore detector monitor-ing 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 of accumulated operation in MODE 1.

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.

A measurement-calculational uncertainty factor of 1.07, 2.

An engineering uncertainty factor of 1.03, 3.

A THERMAL POWER measurement uncertainty factor of 1.02.

PIf the incore system becomes inoperable, reduce power to M x N within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and monitor linear heat rate in accordance with Specification 4.2.1.3.

ST.

LUCIE UNIT 1 3/4 2-2 Amendment No. ~,

134

TABLE 3.3-6 RADIATION HONITORI G INS RUMENTATIO

~Sf~l~NT 1.. AREA MONITORS MINIMUM CHANNELS APPLICABLE OPERABLE MODES ALARM SETPOINT MEASUREMENT RANGE ACTION a.

Fuel Storage Pool Area b.

Containment (CIS) c.

Containment Area Hi Range 2.

PROCESS MONITORS a.

Containment 3

6 g 15 mR/hr

< 90 mR/hr 1, 2, 3

8 4

g 10 R/hr 10"- 10 mR/hr 13 1

10 mR/hr 16 1 10" R/hr 15 i.

Gaseous Activity.

RCS Leakage Detection ii.

Particulate Activity

'CS Leakage Detection b.

Fuel Storage Pool Area Ventilation System 1, 2, 3

8 4

Not Applicable 10 - 10 cpm 1, 2, 3 5 4 Not Applicable 10 - 10 cpm 6

i4 i.

Gaseous Activity ii.

Particulate Activity 10 " 10 pCi/cc 12 1 10 cpm 12

. *With fuel in the storage pool or building.

• • With irradiated fuel in the storage pool or whenever there is fuel movement within the pool or crane operation with loads over the storage pool.

• • • The Alarm Setpoints are determined and set in accordance with requirements of the Offsite Dose Calculation Manual.

~<

~

j 1

~

1

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

a.

At least 755 of all incore detector locations, and b.

A migimum of two quadrant symmetric incore detector locations per core quadrant.

An OPERABLE incore detector location shall consist of a fuel assembly contain-ing 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 0 By performance of a CHANNEL CHECK within 7 days prior to its use when required for:

1.

Recalibration of the excore axial flux offset detection sys tern, 2.

Monitoring the linear heat rate pursuant to Specification 4.2.1. 4, ST LUCIE - UNIT 1

3/4 3-25 Amendment No. gf, )Pi f87.134

Pages 3/4 3-38 through 3/4 3-40 (Amendment No. 115) have been deleted from the Technical Specifications.

The next page is 3/4 3-41.

ST.

LUGIE - UNIT 1

3/4 3-37 Amendment No.~

134

TABLE 3.3-11 Continued ACTION STATEMENTS ACTION 1

With the number of OPERABLE channels less than requ'fred by Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2-Mith position indication inoperable, restore the inoperable indicator to OPERABLE status or close the associated PORV block valve and remove power from its operator within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 3-ACTION 4-ACTION 5-With any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information once per shift to determine valve position.

With the number of OPERABLE Channels one less than the Total Number of Channels shown in Table 3.3-11, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a

Special Report to the Commission pursuant to the specification 6.9.2 within 30 days following the event outlining the action

taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirements of Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are feasible without shutting down or:

1.

Initiate an alternate method of monitoring the reactor vessel inventory; and 2.

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and 3.

Restore the Channel to OPERABLE status at the next scheduled refueling.

ST LUCIE UNIT 1

3/4 3-43 Amendment No. p/, 79.

In m

INSTRUMENT TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL CHECK CALIBRATION 1.

Pressurizer Water Level 2.

Auxiliary Feedwater Flow Rate 3.

Reactor Coolant System Subcooling Margin Monitor 4.

PORV Position Indicator 5.

PORV Block Valve Position Indicator 6.

Safety Valve Position Indicator 7.

Incore Thermocouples 8.

Containment Sump Water Level (Narrow Range) 9.

Containment Sump Water Level 10.

Reactor Vessel Level Monitoring System 11.

Containment Pressure R:

Pages 3/4 3-46 through 3/4 3-49 (Amendment No. 123) have been deleted from the Technical Specifications.

The next page is 3/4 3-50.

ST.

LUGIE - UNIT 1

3/4 3-45 134

TABLE 3. 3-13 EXPLOSIVE GAS MONITORING INSTRUMENTATION INSTRUMENT MINIMUM CHANNELS E

I A~II LT 1.

WASTE GAS DECAY TANKS EXPLOSIVE GAS MONITORING SYSTEM a.

Oxygen Monitors AB NOTATIO

• During waste gas system operation.

ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of this system may continue for up to 30 days provided samples of Oz are analyzed by the lab gas partitioner at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ST.

LUCIE - UNIT 1 3/4 3-51 Amendment No. 59,~ l

INSTRUMENT TAB 4.3-EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE CHECK HE K

CALEHH TT TEST

~HE HT EC 1.

WASTE GAS DECAY TANKS EXPLOSIVE GAS MONITORING SYSTEM a.

Oxygen Monitor

b. Oxygen Monitor (alternate)

Q(1)

Q(1)

AB E OTATI N

• During waste gas holdup system operation.

(1)

The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

l.

One volume percent

oxygen, balance nitrogen, and 2.

Four volume percent

oxygen, balance nitrogen.

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued Pages 3/4 4-28 through 3/4 4-55 (Amendment No. 90),

and Pages 3/4 4-56 through 3/4 4-57 (Amendment No. 80) have been deleted from the Technical Specifications.

The next page is 3/4 4&8.

ST.

LUGIE - UNIT 1

3/4 4-27 Amendment No. M, 134

'LANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION aaaaaaaaaaaaaaaaaaaa aaaaaaaaaaaaaaaaa aaaaaaaaaaaaaaaaa

= aaaaaaaaaaaa 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, be in HOT STANDBY within the next 6

hours.

MODES 2

and 3

- With one or both main steam isolation valve(s) inoperable, subsequent oper ation in MODES 2 or 3 may proceed provided the isolation valve(s) is (are) maintained closed.

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS aaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaaa a

4.7. 1.5 Each main steam line isolation valve that is open shall be demonstrated OPERABLE by verifying full closure within 6.0 seconds when tested pursuant to Specification 4.0.5.

ST.

LUCIE UNIT 1

3/4 7-9 Amendment No. %} ~~~

Pages 3/4 7-11 through 3/4 7-12 (Amendment No. 86) have been deleted from the Technical Specifications.

The next page is 3/4 7-13.

ST.

LUCIE - UNIT 1 3/4 7-10 Amendment No. ~ 134

Pages 3/4 11-2 through 3/4 11-13 (Amendment No. 123) have been deleted from the Technical Specifications.

The next page is 3/4 11-14.

ST.

LUCIE - UNIT 1 3/4 11-1 AMENDMENT NO.NN

~~~'34

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES I

The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to withinIllof its design pressure of l000 psia during the most severe anticipated system opera-tional transient.

The maximum relieving capacity is associated with a turbine trip from 100K RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e.,

no steam bypass to the condenser).

The specified valve liftsettings and relieving-capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure

Code, 1971 Edition and ASME Code for Pumps and Valves, Class II.

The total relieving capacity for all valves on all of the steam lines is 12.38 x 10'bs/hr which is 102.8 percent the total secondary steam flow of 12.04 x 10~ lbs/hr at 100% RATE) THERMAL POWER.

A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the'reduction in secondary system steam flow'and THERMAL POWER required by the reduced reactor trip settings of. the Power Level-High channels.

The reactor trip setpoint reductions are derived on the following bases:

For two loop operation SP

X x (106.5) where:

SP reduced reactor, trip setpoint in percent of RATED THERMAL POWER maximum number of inoperable safety valves per steam line ST.

LUCIE - UNIT 1

B 3/4 7-1

.~amendment

'No +g-, 134

PLANT SYSTEMS BASES 106.5

=

Power Level-High Trip Setpoint for two loop operation X

=

Total relieving capacity of all safety valves per steam line in 1bs/hour (6.192 x 106 1bs/hr. )

Y

=

Maximum relieving capacity of any one safety valve in lbs/hour (7.74 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 325'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 reprove reactor decay heat and reduce the RCS temperature to 325'F where the shutdown cooling system may be placed into operation fot'ontinued 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 325'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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with'team 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.

LUCIE - UNIT 1

B 3/4 7-2 Amendment No. EA.48

3 4. 11 RADIOACTIVE EF F I.VENTS BASES Pages 8 3/4 11-2 through B 3/4 11%

(amendment No, 123} have been deleted from the Technical Specifications.

The next page is 8 3/4 ll.4, ST.

LUCIE UNIT 1

B 3/4 11-1 Amendment No. 89, ~,

134