ML17215A654

Amend 8 to License NPF-16,revising Tech Specs for Cycle 2 Operation

ML17215A654
Person / Time
Site:	Saint Lucie
Issue date:	11/09/1984
From:	John Miller Office of Nuclear Reactor Regulation
To:
Shared Package
ML17215A655	List: ML17215A654 ML17215A657
References
NUDOCS 8411280395
Download: ML17215A654 (79)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FLORIDA POllER 5 LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO, FLORIDA AND FLORIDA MUNICIPAL POlJER AGENCY DOCKET NO. 50-389 ST.

LUCIE PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 8 License No. NPF-16 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Florida Power 8 Light Company, et al.,

(the licensee) dated June 4, 1984, 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; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

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; D.

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 E.

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.

84ii280395 84ii09-

'DR ADOCK 05000389

P

,PDR

2.

Accordingly, Facility Operating License No.

NPF-16 is amended by changes to the Technical Specifications as indicated in the attach-ment to this license amendment, and by amending paragraph 2.C.2 to read as follows:

2.

Technical S ecifications The Technical Specifications contained in Appendices A

and B, as revised 'through Amendment No.

8

, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifica-tions.

3.

This license amendment is effective as of the date of its issuance.

Attachment:

Changes to the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION James R. Miller, Chief Operating Reactors Branch ¹3 Division of Licensing Date of Issuance:

November 9, 1984

ATTACHMENT TO LICENSE AMENDMENT NO. 8 FACILITY OPERATING LICENSE NO.

NPF-16 OOCKET NO. 50-389 Remove and replace the following pages of the Appendix A Technical Specifica-tions with the attached pages.

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

The cor res-ponding overleaf pages are provided to maintain document completeness.

Remove Insert Remove Insert XXI XXII XXIII XXIV XXV 2-1 2-3 2-4 2-5 2-9 2-10 B 2-1 B 2-2 B 2-4 3/4 1-3 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-14 3/4 1-17 3/4 1-18 3/4 1-19 3/4 1-20 3/4 1-24 3/4 1-28 3/4 2-4 3/4 2-5 3/4 2-7 3/4 2-9 3/4 2-11 3/4 2-12 3/4 2-15 3/4 3-6 3/4 3-17 3/4 3-20 3/4 3-21 3/4 4-9 XXI XXII XXIII XXIV XXV 2-1 2-3 2-4 2-5 2-9 2-10 B 2-1 B 2-'2 B 2-4 3/4 1-3 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-14 3/4 1-17 3/4 1-18 3/4 1-19 3/4 1-19a 3/4 1-20 3/4 1-24 3/4 1-28 3/4 2-4 3/4 2-5 3/4 2-7 3/4 2-9 3/4 2-11 3/4 2-12 3/4 2-15 3/4 3-6 3/4 3-17 3/4 3-20 3/4 3-21 3/4 4-9 3/4 7-1 3/4 7-2 3/4 7-3 3/4 7-10 B 3/4 1-1 B 3/4 1-2 B 3/4 1-4 B 3/4 2-2 B 3/4 2-3 B 3/4 7-1 5-1 5-3 3/4 7-1 3/4 7-2 3/4 7-3 3/4 7-10 B 3/4 1-1 B 3/4 1-2 B 3/4 1-4 B 3/4 2-2 B 3/4 2-3 B 3/4 7-1 5 5-3

LIST OF 'FIGURES INDEX FIGURE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING.

PAGE 2-3

2. 2-1 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 1

(FRACTION OF RATED THERMAL POWER VERSUS QR2)................

2-7

2. 2-2 2.2-3

2. 2-4 B 2.1-1 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 2

(QR2 VERSUS Y1).

THERi%LL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1

(Y1 VERSUS A1).......

THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS QR1)

AXIAL POWER DISTRIBUTION FOR THERMAL MARGIN SAFETY LIMITSt

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2-8 2-9 2-10 B 2-2 3.1-1 3.1-1a

3. 1-2 3.2-1 3.2-2 3 ~ 2 3

4. 2-1 3.2-4 ALLOWABLE PEAK LINEAR HEAT RATE VS BUR/UP.

AXIAL SHAPE INDEX VS FRACTION OF MAXIMUM ALLOWABLE POWER LEVEL PER SPECIFICATION 4.2.1.3.........

ALLOWABLE COMBINATIONS OF THERMAL POWER AND F F

AUGMENTATION FACTORS VS DISTANCE FROM BOTTOM OF CORE......

3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6 AXIAL SHAPE INDEX OPERATING LIMITS WITH FOUR REACTOR COOLANT PUMPS OPERATING.

3/4 2-12 MINIMUM BORIC ACID STORAGE TANK VOLUME AND TEMPERATURE AS A FUNCTION OF STORED BORIC ACID CONGENTRATION............ 3/4 1-.15 ALLOWABLE TIME TO REALIGN. CEA VS INITIALF.................

3/4 1-19a CEA INSERTION LIMITS VS THERMAL POWER WITH FOUR REACTOR COOLANT PUMPS OPERATING. '.

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3/4 1-28 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMITS VERSUS PERCENT OF RATED THERMAL, POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY )

1 pCi/GRAM DOSE EQUIVALENT I-131...................

3/4 4-28 3.4-2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITATIONS FOR 0 TO 2

YEARS OF FULL POWER OPERATION........

3/4 4-31 ST.

LUCIE - UNIT 2 XXI Amendment No.8

1 LIST OF FIGURES Continued INDEX FIGIIRE 3.4-3

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PAGE REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITATIONS FOR 2

TO 10 YEARS OF FULL POWER OPERATION...................

3/4 4-32 3.4-4 4.7-1 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITATIONS FOR 10 TO 40 YEARS OF FULL POWER OPERATION............

SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST 3/4 4 33 i

3/4 7-25 B 3/4.4-1 NIL-DUCTILITYTRANSITION TEMPERATURE INCREASE AS A FUNCTION OF FAST (E>l MeV) NEUTRON FLUENCE (550'F IRRADIATION )

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B 3/4 4-10

5. 1-1

6. 2-1 SITE AREA MAP................

5-2 OFFSITE ORGANIZATION FOR FACILITY ORGANIZATION AND TECHNICAL SUPPORT...........

6-3 6.2-2 UNIT ORGANIZATION....

6-4 ST.

LUCIE - UNIT 2 XXII Amendment No. 8

LIST OF 3ABLES INDEX TABLE 1.2

2. 2-1 3.1-1 3.2-1 3.2-2 3.3-1

3. 3-2 FREQUENCY NOTATION......

'OPERATIONAL MODES.......

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS....

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MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR ST. LUCIE-2...................

DELETED...............

DNB MARGIN LIMITS.....

REACTOR PROTECTIVE INSTRUMENTATION..............

REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES.....

PAGE 1-8 1-9 2-4 3/4 1-17 3/4 2-11 I

3/4 2-15 3/4 3-2 3/4 3-6 4.3-1 3 ~ 3 3 3.3-4

3. 3-5 4.3-2 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............................................

3/4 3-8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.................

3/4 3-12 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES.

~. 3/4 3-17 ENGINEERED SAFETY FEATURES RESPONSE TIMES......

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3/4 3-19

-ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................

3/4 3-22 3.3-6 4.3-3 I

3 ~ 3 7

4.3-4 RADIATION MONITORING INSTRUMENTATION.........

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS SEISMIC MONITORING INSTRUMENTATION..............

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-25 3/4 3-28 3/4 3-33 3/4 3-34 3.3-8 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...

3/4 3-37, METEOROLOGICAL MONITORING INSTRUMENTATION............... 3/4 3-36 ST.

LUCIE-UNIT 2 XXIII Amendment No. 8

LIST OF TABLES Continued INDEX TABLE 3.3-9 4.3-6 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION.

REMOTE SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............

3. 3-10 ACCIDENT MONITORING INSTRUMENTATION..........

PAGE 3/4 3-39 3/4 3-40 3/4 3-42 4.3-7 3.3-11 3.3-12 4.3-8 3.3-13 4.3-9

4. 4-1 4.4-2 3.4-1 3.4-2 4.4-3 4.4 4

ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............

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FIRE DETECTION INSTRUMENTS..............

3/4 3-43 3/4 3-45 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION.... 3/4 3-49 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........................

3/4 3-51 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION... 3/4 3-54 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3/4 3-57 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION........................

3/4 4-16 STEAM GENERATOR TUBE INSPECTION.................

. 3/4 4-17 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES. ~........

3/4 4-21 REACTOR COOLANT SYSTEM CHEMISTRY..................

~ ~...... 3/4 4-23 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS 3/4 4-24 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM...........

3/4 4-27

4. 4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE.................

3/4 4-34

3. 6-1 3.6-2 CONTAINMENT LEAKAGE PATHS.........

CONTAINMENT ISOLATION VALVES 3/4 6-5 3/4 6-21 ST.

LUCIE-UNIT 2 XXIV Amendment No. 8

LIST OF TABLES Continued INDEX TABLE 3.7-1 PAGE MAXIMUMALLOWABLE LINEAR POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERA-TION WITH BOTH STEAM GENERATORS 3/4 7-2 4.7-0 4.7-1 3.7-3a STEAM LINE SAFETY VALVES PER LOOP.....

SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAMS

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SAFETY-RELATED HYDRAULIC SNUBBERS........

3/4 7-.3 3/4 7-8 3/4 7-26 3.7-3b SAFETY-RELATED MECHANICAL SNUBBERS.............

. 3/4 7-27 3.7-4 3.7-5 4.8-1 4'-2 3.8-1 FIRE HOSE STATIONS......................

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3/4 7-36 YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES.....

3/4 7-38 DIESEL GENERATOR TEST SCHEDULE.............................

3/4 8-8 BATTERY SURVEILLANCE REQUIREMENT..........................

3/4 8-12 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES..

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3/4 8-18 4.11-1 4.11-2 3.12-1 3.12-2 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM.... 3/4 11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM... 3/4 11-8 RADIOLOGICAL ENVIRONMEiNTAL MONITORING PROGRAM.............

3/4 12-3 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES......

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3/4 12-7 4.12-1 B 3/4.2-1 DELETED..

B 3/4 2-3 I

B 3/4 4-9 B 3/4.4-1 REACTOR VESSEL TOUGHNESS DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS.. 3/4 12-8 5.7-1 6.2-1 COMPONENT CYCLIC OR TRANSIENT LIMITS...

5-5 MINIMUM SHIFT CREW COMPOSITION -

TWO UNITS WITH TWO SEPARATE CONTROL ROOMS..

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LUGIE - UNIT 2 XXV Amendment No.8

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. 1 SAFETY LIMITS

2. 1. 1 REACTOR CORE DNBR

2. 1. l. 1 The combination of THERMAL POWER, pressurizer

pressure, and maximum cold leg coolant temperature shall not exceed the limits shown un Figure 2. 1-1.

APPLICABILITY:

MODES 1 and 2.

ACTION:

Whenever the combination of THERMAL POWER, pressurizer pressure and maximum cold leg coolant temperature has exceeded the limits shown on Figure 2. 1-1, be in.HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7. l.

PEAK LINEAR HEAT RATE

2. 1.'L.2 The peak linear heat rate of the fuel shall be maintained less than or equal to 22.0 kW/ft (value corresponding to centerline fuel melt).

APPLICABILITY:

MODES 1 and 2.

ACTION:

Whenever the peak linear heat rate of the fuel has exceeded 22.0 kW/ft (value corresponding to centerline fuel melt),

be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and

'omply with the requirements of Specification 6.7. 1.

REACTOR COOLANT SYSTEM PRESSURE

2. 1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.'PPLICABILITY:

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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7. l.

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, and comply with the requirements of Specification 6.7. l.

4T.

LUCIE - UNIT 2 2-1 Amendment No. 8

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 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 s'tatement requirement of Specification 3.3.

1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

'ST.

LUCIE - UNIT 2 2-2

580 560 UNACCEPTABLE OPER ATIOH CA CD

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0. 60

0. 80

l. 00

1. 20

1. 40

1. 60

1. 80

2. 00

2. 20 FRACTION OF RATED TIIEIRilAL POllER

TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS 3. Pressurizer Pressure - High 4. Thermal Margin/Low Pressure Four Reactor Coolant Pumps Operating 5. Containment Pressure - High 6. 7. Steam Generator Pressure Low Steam Generator Pressure (1) Difference High (Logic in TM/LP Trip Unit) FUNCTIONAL UNIT I 1. Manual Reactor Trip M 2. Variable Power Level - High Four Reactor Coolant Pumps Operating TRIP SETPOINT Not Applicable < 9.61K above THERMAL POWER, with a minimum setpoint of 15K of RATED THERMAL POWER, and a maximum of < 107.0X of RATED THERMAL POWER. < 2370 psia Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4. Minimum value of 1900 psia. < 3.0 Ps>g > 626.0 psia (2) < 120.0 ps>d ALLOWABLE VALUES Not Applicable < 9.63% above THERMAL POWER, and a minimum setpoint of 15K of RATEO THERMAL POMER and a maximum t of < 107.0X of RATED THERMAL POWER. < 2374 psia Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4. Minimum value of 1900 psia. < 3. l Pslg > 621.0 psia (2) < 132.0 psid O 8. Steam Generator Level Low > $9.5X (3) > 39.1X (3)

TABLE 2.2-1 (Continued REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS InM CTI M FUNCTIONAL UNIT 9. Local Power Oensity - High 10. Loss of Component Cooling Mater to Reactor Coolant Pumps-Low ll. Reactor Protection System Logic 12. Reactor Trip Breakers 13. Rate of Change of Power - High 14. Reactor Coolant Flow - Low 15. Loss of Load (Turbine) Hydraulic Fluid Pressure - Low TRIP SETPOINT Trip setpoint adjusted to not exceed the limit lines. of Figures 2.2-1 and 2.2-2. > 636 gpm"" Not Applicable Not Applicable < 2.49 decades per minute > 95.4X of. design Reactor Coolant flow with four pumps operating* > 800 psig ALLOMABLE VALUES Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2. > 636 gpm Not Applicable. Not Applicable < 2.49 decades.per minute > 94.9X of design Reactor Coolant flow with four pumps operating" > 800 psig Oesign reactor coolant flow with four pumps operating is 363,000 gpm. 10-minute time delay after relay actuation. O \\ CO

TABLE 2. 2-1 '(Continued) REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATION (1) Trip may be manually bypassed below 0.5X of RATED THERMAL POWER during testing pursuant to Special Test Exception

3. 10.3; bypass shall be automatically removed when the THERMAL POWER is greater than or equal to 0.5X of RATED THERMAL POWER.

(2) Trip may be manually bypassed below 705 psig; bypass shall be automatically removed at or above 705 psig. (3) X of the nar row range steam generator level indication. (4) Trip may be bypassed below 10-~X and above 15X of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 10-~X or < 15K of RATED THERMAL POWER. (5) Trip may be bypassed below 15K of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 15K of RATED THERMAL POWER.

1. 70 1.60 1.50 t

ERE.Mg A'gRj=.Qp'ggj ND-P =1400--x-g B--+-1-7".85-x--T '.--941 0 4 -0.8 ASI-1 +1 2'.40

1. 30

+ l 1.20 I+0".867 1.10 =',CO 1.00 -0.6 -0.4 -0.2 0.0 0.2 0.4 0.6 AXIAL SHAPE INDEX, Yl FIGURE 2.2-3 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1 Yl Versus Al ST. LUCIE-UNIT 2 2-9 Amendment No. 8

WHERE: Al x QRl QDNB AND P = 1400 x Q + 17.85 x T. - 9410 ] i 1.0 0.8 ~ = ~ 0 -- 0.85-.! I i I .95. I .95 QR1 0.6 0.4 0.2 4 0.2 0.4 0.6 0.8 1.0 1.2 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. LUGIE - UNIT 2 2-10 Amendment No. 8

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel clad-ding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centerline fuel melting will occur. 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 cladding surface temperature is 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 coeffi-cient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the CE-1 correlation. The CE-1 DNB correlation has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distributions. 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 opera- . tional transients, and anticipated transients is limited to 1.28. This value is derived through a statistical combination of the system parameter probability distribution functions with the CE-1 DNB correlation uncertainty. This value corresponds to a 95K probability at a 95K confidence level that DNB will not occur and is chosen as an appropriate margin to DNB.for all operating conditions. The curves 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 Cool-ant Pumps operating for which the minimum DNBR is no less than 1.28 for the family of axial shapes and corresponding radial peaks shown in Figure B Z.l-l. The limits in Figure 2.1-1 were calculated 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 112K of RATED THERMAL POWER is prohibited by the high power'evel trip set-point sepcifed in Table 2.2-1. The area of safe operation is below and to the left of these lines. The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure. The Thermal Margin/Low Pressure and Local Power Density Trip Systems, in conjunction with Limiting Conditions for Operation, the Variable Overpower Trip and the Power Dependent Insertion Limits, assure that the Specified Acceptable Fuel Design Limits on DNB and Fuel Centerline Melt are not exceeded during normal operation and design basis Anticipated Oqeration Occurrences. ST. LUGIE-UNIT 2 B 2-1 Amendment No. 8

m I 2.0 1.8 1.6 1.4 I-H 1.2 o 1.0 0.8 >C o 0.6 LJJ IV 0.4 C) 0.2 0.0 F~ = 1.67~ FT = 1.81 FR = 1.79 FT = 1.77 FT = 1.62 R 50 75 100 PERCENT OF ACTIVE CORE LENGTH FROM BOTTOM Figure 0 2.1-1 Axial power distribution for thermal margin safety limits

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES

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 contains in the reactor coolant from reaching the containment atmosphere.

The Reactor Coolant System components are designed to Section III, 1971 Edition including Addenda to the

Summer, 1973, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 11'2750 psia) of design pressure.

The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements. The entire Reactor Coolant System was hydrotested at 3125 psia to demonstrate integrity prior to initial operation. 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 R ACTOR TRIP SE POIN S The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the 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. ST. LUCIE - UNIT 2 B 2-3

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Variable Power Level-Hi h A Reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin/Low Pressure Trip. The Variable Power Level High.trip setpoint is operator adjustable and can be set no higher than 9.6V'bove 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 107.0X of RATED THERMAL POWER and a minimum setpoint of 15.0X 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 112%%uo of RATED THERMAL POWER, which is the value'used in the safety analyses. Pressurizer Pressure-Hi h The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than 1.28. The trip is initiated whenever the React'or Coolant System pressure signal drops below either 1900 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of 6T power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX'he 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 Specifica-tions 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 2.0X of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 3.0 F to compensate for potential temperature measurement uncertainty; and a further allowance of 91.0 psia to compensate for pressure measurement error and time delay associated with providing effective termination of the occurrence that exh'ibits the most rapid decrease in margin to the safety limit. The 91.0 psia allowance is made up of a 25.0 psia pressure measurement allowance and a 66.0 psia time delay allowance. ST. LUCIE " UNIT 2 Amendment No.8 The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2375 psia which is below the nominal lift setting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves. Thermal Mar in/Low Pressure

~ y ~ REACTIVITY CONTROL SYSTEMS" SHUTDOWN MARGIN - T LESS THAN OR E UAL TO 200 F LIMITING CONDITION FOR OPERATION 3.1.1;2 The SHUTDOWN MARGIN shall be greater than or equal'to 3.0%%uo delta k/k. APPLICABILITY: MODE 5. ACTION: 1 With the SHUTDOWN MARGIN less than 3.0X delta k/k, immediately initiate and continue boration at greater than-or equal to 40 gpm of a solution containing greater than or equal to 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE RE UIREMENTS

4. l. 1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 3.

O%%uo: delta k/k: a ~ b. Within 1 hour after detection of an inoperable CEA(s) and at least once per 12 hours 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 untrippable CEA(s). At least once per 24 hours by consideration of the following factors: l. 2. 3. 5. 6. Reactor coolant system boron concentration, CEA position, Reactor c'oolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration,and Samarium concentration. C. At least once per 24 hours, when the Reactor Coolant System is 'rained below the hot leg centerline, by consideration of the factors in 4. 1. 1.2b. and by verifying at least two charging pumps are rendered inoperable by racking out their motor circuit breakers. ST. LUCIE " UNIT 2 3/4 1-,3 Amendment No. 8

REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1.3 The flow rate of reactor coolant'o the reactor pressure vessel shall be > 3000 gpm whenever a reduction in Reactor Coolant System boron concentration 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 1 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 2 3/4 1-4

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 and capable of being powered from an OPERABLE emergency power source: 'a ~ A flow path from the boric acid makeup tank via either a boric acid makeup 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 or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE RE UIREMENTS

4. 1.2.

1 At least one of the above required flow paths shall be demonstrated OPERABLE: At least once per 7 days by 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. 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 2 3/4 1-7

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three bbr'on 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'cid makeup pump or a gravity feed connection, and a charging pump'o the Reactor Coolant System, and b. The flow path from the refueling water tank via a charging pump to. 'he 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 injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 3.0%%uo delta k/k at 200'F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE RE UIREMENTS 4.*1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE: a ~ b. C. d. At least once per 7 days by 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. 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 otherwi'se secured in position, is in its correct position.

At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on an SIAS test signal. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a delivers at least 40 gpm to the Reactor Coolant System. ST. LUCIE - UNIT 2 t 3/4 1-8 Amendment No.

~ ~ ~ REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - SHUTDOMN 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 pursuant to Specification

3. 1.2.

1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source, APPLICABILITY: MODES 5 and 6. ACTION: With no charging pump or high pressure safety injection pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE RE UIREMENTS

4. 1.2.3 At least the above required pump shall be demonstrated OPERABLE by verifying the charging pump develops a flow rate of greater than or equal to 40 gpm or the high pressure safety injection pump develops a total head of greater than or equal to 2854 ft when tested. pursuant to Specification 4.0.5.

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

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,',

3. and 4.

ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 3.0X delta k/k at 200 F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE RE UIREMENTS

4. 1.2.4. 1 At least two charging pumps shall be demonstrated OPERABLE by verifying that each pump develops a flow rate of greater than or equal to 40 gpm when tested pursuant to Specification 4.0.5.

4. 1.2.4.2 At least once per 18 months verify that each charging pump starts automatically on an SIAS test signal.

ST. LUCIE - UNIT 2 3/4 1"10 Amendment No. 8

REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 At least one boric acid makeup pump shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if only the flow path through the boric acid pump in Specification

3. l. 2. la. is OPERABLE.

APPLICABILITY: MODES 5 and 6. ACTION: With no boric acid makeup pump OPERABLE as required to complete the flow path of Specification

3. 1. ala.,

suspend all operations involving,CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE RE UIREMENTS

4. 1.2.5 The above required boric acid makeup pump shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump develops a

discharge pressure of greater than or equal to 90 psig when tasted pursuant to Specification 4.0.5. ST. LUCIE - UNIT 2 3/4 1-11

REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS " OPERATING LIMITING CONDITION FOR OPERAT'ION 3.1.2.6 At least the boric acid makeup pump(s) in the boron injection flow path(s) required OPERABLE pursuant to,Specification

3. 1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if tPe flow path through the boric acid pump(s) in Specification

3. 1. 2. 2a is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one boric acid makeup pump required for the boron injection flow path(s) pursuant to Specification 3.1.2.2a inoperable, restore the boric acid makeup pump to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 3.0X delta k/k at 200 F; restore the above required boric acid makeup pump(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE RE UIREMENTS

4. 1.2.6 The above required boric acid makeup pump(s) shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump(s) develop a

discharge pressure of greater than or equal to 90 psig when. tested pursuant to Specification 4.0.5. ST. LUCIE - UNIT 2 3/4 1-12 Amendment No. 8

REACTIVITY CONTROL SYSTEMS BORATED MATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 1.2.7 As a minimum, one of the fol.lowing boratgd water sources shall be OPERABLE:

a. Ohe boric acid makeup tank and at least one associated heat tracing circuit with a minimum contained volume of 4150 gallons of 8 weight percent boron. b. The refueling water tank with: l. A minimum contained borated water volume of 125,000 gallons, 2. A minimum boron concentration of 1720

ppm, and 3.

A solution temperature between 40'F and 120 F. APPLICABILITY: NODES 5 and 6. ACTION: With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE RE UIREMENTS

4. 1.2.7 The above required borated water source shall be demonstrated OPERABLE:

a 0 b. At least once 'per 7 days by: l. Verifyin'g the boron concentration of the water, 2. Verifying the contained borated water volume of the

tank, and 3.

Verifying the boric acid makeup tank solution temperature when it is the source of borated water. At least. once per 24 hours by verifying the RMT temperature when it is the source of borated water and the outside air temperature is outside the range of 40 F and 120'F. ST. LUCIE - UNIT 2 3/4 1-13

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 b. Each of the following borated water sources shall be OPERABLE: At least one boric acid makeup tank and at least one associated heat tracing circuit per tank with the contents of the tank in accordance with Figure 3.1-1, and The refueling water tank with: l. A minimum contained borated water volume of 417,100 gallons, 2. A boron concentration of between 1720 and 2100 ppm of boron, and 3. A solution temperature between 55'F and 100 F. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: a. b. With the above required boric acid makeup tank inoperable, restore the tank to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 3.0X delta k/k at 200'F; restore the above required boric acid makeup tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE UIREMENTS 4.1.2.8 a. b. Each borated water source shall be demonstrated OPERABLE: At least once per 7 days by: l. Verifying the boron concentration in the water, 2. Verifying the contained borated water volume of the water

source, and 3.

Verifying the boric acid makeup tank solution temperature. At least once per 24 hours by verifying the RWT temperature when the outside air temperature is outside the range of 55'F and 100'F. ST. LUCIE - UNIT 2 3/4 1-14 Amendment No. 8

TABLE 3.1-1 MONITORING FRE UENCIES FOR BACKUP BORON DILUTION DETECTION FOR ST. LUGIE-2 MODE Number of OPERABLE Charging Pumps* 0 1 2 3 12 hr 100 min 40 min 12 hr 130 min 50 min 25 min 30 min 5 (RCS level below hot leg centerline) 8 hr 8 hr 100 min 40 min 25 min 35 min Operation not Operation not allowed** allowed** 6 24 hr 220 min 95 min 55 min

• Charging pump OPERABILITY for any period of time shall constitute OPERABILITY for the entire monitoring frequency.
  • In MODE 5 with the RCS level below, the hot leg centerline, at least two charging pumps shall be verified to be inoperable by racking out their motor circuit breakers.

ST. LUCIE-UNIT 2 3/4 1-17 Amendment No. 8

REACTIVITY CONTROL SYSTEMS 3/4. 1.3 MOVABLE CONTROL ASSEMBLIES CEA POSITION LIMITING CONDITION FOR OPERATION 3.1.3.1 The CEA Block Circuit and all full-length (shutdown and regulating) CEAs which are inserted in the core, shall be OPERABLE with each CEA of a given group positioned within 7.0 inches (indicated position) of all other CEAs in its group. APPLICABILITY: MODES 1" and 2". ACTION: a 0 b. C. d. 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, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and bq in at least HOT STANDBY within 6 hours. With the CEA Block Circuit inoperable, within 6.hours either: 1. With one CEA position indicator per group inoperable take action per Specification

3. 1.3.2, or 2.

With the group overlap and/or sequencing interlocks inoperable maintain CEA groups 1, 2, 3 and 4 fully withdrawn and the CEAs in group 5 to less than 15/ insertion.and place and maintain CEA drive system in either the "Manual" or "Off" position, or 3. Be in at least HOT STANDBY. With more than one full-length CEA inoperable or misaligned from any other CEA in its group by more than 15 inches (indicated pos.ition), be in at least HOT'STANDBY within 6 hours. With one full-length CEA misaligned from any'ther CEQ in its group by more than 15 inches, operation in NODES 1 and 2 may continue, provided that the misaligned CEA is~ositioned within 15 inches of the other CEAs in its group in accordance with the time constraints shown in Figure 3.1-la. See Special Test Exceptions

3. 10.2,

3. 10.4, and 3;10.5.

ST. LUCIE " UNIT 2 3/4 1-18 Amendment No.

REACTIVITY CONTROL SYSTEMS ACTION e. g, (Continued) With one full-length CEA misaligned from any other CEA in its group by more than 15 inches beyond the time constraints shown in Figure 3.l-la, reduce power to < 705 of RATED THERMAL POWER prior to completing ACTION e.l or e.2. 1. Restore the CEA to OPERABLE status within its specified alignment requirements, or 2. Declare the CEA inoperable and satisfy SHUTDOWN MARGIN require-ment of Specification 3.1.1.1. After declaring the CEA inoper-able, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided:* a) Within 1 hour the remainder of the CEAs in the group with the inoperable CEA shall be aligned to within 7.0 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. b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours. With one or more full-length CEA(s) misaligned from any other CEAs in I its group by more than 7.0 inches but less than or equal to 15 inches, operation in MODES 1 and 2 may continue, provided that within 1 hour the misaligned CEA(s) is either: 1. Restored to OPERABLE status within its above specified alignment requirements, or 2. Declared inoperable and the SHUTDOWN MARGIN requirement of Specifica-tion 3.1.1.1 is satisfied. After declaring the CEA inoperable, opera-tion in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided: a) Within 1 hour the remainder of the CEAs in the group with the inoperable CEA shall be aligned to within 7.0 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. b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours. Otherwise, be in at least HOT STANDBY within 6 hours. With 'one full-length CEA inoperable due to causes other than addressed I by ACTION a.,

above, and inserted beyond the Long Term Steady State Insertion Limits but within its above specified alignment requirements, operation in MODES 1

and 2 may continue pursuant to the requirements of Specification 3.1.3.6. I

• Ifthe pre-misalignment ASI was more negative than -0.15, reduce power to <70!

of RATED THERMAL POWER or 705 of the THERMAL POWER level prior to the misalign-ment, whichever is less, prior to completing ACTION e.2.a) and e.2 '). ST. LUCIE-UNIT 2 3/4 1-19 Amendment No. B

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REACTIVITY CONTROL SYSTEMS ACTION: (Continued) h. With one full-length CEA inoperable due to causes other than addressed by ACTION a.,

above, but within its above specified alignment require-ments and either fully withdrawn or within the Long Term Steady State Insertion Limits if in full-length CEA group 5, operation in MODES 1

and 2 may continue. SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length CEA shall be determined to be within 7.0 inches (indicated position) of all other CEAs'in its group at least once per 12 hours 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 hours.

4. 1.3. 1.2 Each full-length CEA not fully inserted in the core 'shall be determined to be OPERABLE by movement of at least 7.0 inches in any one direction at least once per 31 days.

4. 1.3. 1.3 The CEA Block Circuit shall be demonstrated OPERABLE at least once per 31 days by a functional test which verifies that the circuit prevents any CEA from being misaligned from all other CEAs in its group by more than 7.0 inches (indicated position).

4. 1.3. 1.4 The CEA Block Circuit shall be demonstrated OPERABLE by a functional test which verifies that the circuit maintains the CEA group overlap and sequencing requirements of Specification

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:

"a. Prior to each entry into NODE 2 from MODE 3, except that such verification need not be performed more often than once per 31 days, and b. At least once per 6 months. The licensee shall be excepted from compliance during the initial startup test program for an entry into MODE 2 from NODE 3 made in association with a measurement of power defect. ST. LUGIE - UNIT 2 3/4 1-20 Amendment No.8

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS - SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 1.3.3 At least one CEA position indicator channel shall be OPERABLE for each shutdown or regulating CEA not fully inserted.

APPLICABILITY: MODES 3", 4,~ and 5". ACTION: With less than the above required position indicator channel(s)

OPERABLE, immediately open the reactor trip breakers.

SURVEILLANCE RE UIREMENTS 4.1.3.3 Each of the above required CEA position indicator channel(s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months. With the reactor trip breakers in the closed position. ST. LUCIE - UNIT 2 I 3/4 1"23

REACTIVITY CONTROL SYSTEMS CEA DROP TIME 'I LIMITING CONDITION FOR OPERATION

3. 1.3.4 The individual full-length (shutdown and regulating)

CEA drop time, from a fully withdrawn position, shall Ue less than or equal to 2.7 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches.its 90%%uo insertion position with: a. T greater than or equa'1 to 515'F, and b. All reactor coolant pumps operating. APPLICABILITY: MODES 1 and 2. ACTION: a ~ b. With the drop time of any full-length CEA determined to exceed the above limit: l. If in MODE 1 or 2, be in at least HOT STANDBY within 6 hours, or 2. If in MODE 3, 4, or 5, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2. With the CEA drop times within limits but determined at less than full reactor coolan. 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 ~ b. C. For all CEAs following each removal and installation of the reactor vessel

head, For specifically affected individuals CEAs following any main-tenance on or modification to the CEA drive system which could affect the drop time of those specific

CEAs, and At least once per 18 months:

ST. LUGIE " UNIT 2 3/4 1-24 Amendment No. 8

REACTIVITY CONTROL SYSTEMS ACTION: (Continued) C. With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits for intervals greater than 5 EFPD per 30 EFPD interval or greater than 14 EFPD per calendar year, either: 1. Restore the regulating groups to within the Long Term Steady State Insertion Limits within 2 hours, or 2. Be in at least HOT STANDBY within 6 hours. SURVEILLANCE RE UIREMENTS

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 hours except during time intervals when the PDIL Auctioneer Alarm Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours.

The accumulated times during which the regulating CEA groups are inserted beyond the Long Term Steady State Insertion Limits but within the Power Dependent Insertion Limits shall be determined at least once per 24 hours. ST. LUCIE - UNIT 2 3/4 Z-27

0.90 'c-O.8O 0.70 0.60 o 0.50 UJ 0.40 C) 0.30 C) I 0.20 0.10 C0 C Cl0 C Ol C Rg e,g ill Q f 0 o. jU O C lA v a clIll o.C Q io 0 g C0 ~4 C g S Q IAao0 (9 LA EO C0 O~ N C0 Qltl C .o Pv 0 U STEADY STATE INSERTION LIMIT STEADY STATE 'N S E RTION LIMIT j g j I I j j o I j . il ..I I .:I 4.ONG TERM-~ -SHOAT TERM-C0 L Ol C g Cl EO Pl o. 0 C9 C4 Cl - 5" 0 20 '0 60: 80'00 0 20 40 (136)(108.8)(81.6) (54') (27 2) (0) (136)(108.8)(81.6) 4 60 80 100 0 20 40 60 80 . 100 (54.4) (27.2) (0) (136)(108.8)(81.6) (54.4) (27.2) (0) 2 0 20 40 60 80 100 0 20 40 60 80 100 (136) (108.8) (81.6) (54.4) (27. 2) (0) (136) (108.8) (81.6) (54.4) (27.2) (0) %CEA INSEIITION (INCHES CEA WITHDRAWNI Fiqure 3.1-2 CFA Insr rtinn I imits vs. THFRHAI POIIFR with Fn11r Reactor (;nnlant PIm)nS Anerztinn

15.0 t ~ 'I ~ t' 1 ~ t ~ ~ ~ ~ Y W~ 14.0 ~l I K QW WQ, XQ z Cl Y0 W ~ Q WW 12.0 0 r ~ ~ ~ ~ ~ ~ ~ ~ ~ I' I ' '4 ~ I t ~ U ~ ~ t ~ I ACCEPTABLE t ~ ~ OPERATION NACCEPTABLE OPER ATION t ~ ~ I ~ t ~ t ~ ~ ~ t 11.0 BOL CYCLE LIFE EOL Figure 3.2-1 Allowable peak linear heat rate vs burnup ST. LUCIE - UNIT 2 3/4 2-3

1.1 1.0 .UNACCEPTABLE OPERATION 'EGION 0.9 C) CL UJ CXl 0.8 (-0,,10,.0,88) (0.15, 0.88',) I -(--0;3; 0.72.) 0.7 ACC EPTAOL- 'BPERSTIOH . *REGION (0;3,'; 0.72) W 0.5 d -0.6 -0.4 -0.2 0.0 0.2 PERIPHERAL AXIAL SHAPE INDEX 0.4 0.6 FIGURE 3.2-2 AXIAL SHAPE INDEX VS FRACTION OF MAXIMUMALLOWABLE POWER LEVEL PER SPECIFICATION 4.2.1.3 ST. LUCIE - UNIT 2 3/4 2-4 Amendment No. $,8

FIGURE 3.2-3 ALLOWABLE CQl1BINATIONS OF THEg%L POWER AND F, Fr'y 1.2 ] UNACCE BLE ll i P. RA%ION REG I ION 1.0 1.)),'i.vs, j.. cUay (Z..7j., 0.~) ~ ~ 0.8 W 'F LIH T CU OPE ATION EBlON YE CCEPJABLE R t i{z; i I i ~ 0.8 83(,0. 85) I I l 0.6 1.65

1. 70

1. 75 MEASURED F Fr'y 1.80

1.10 3II ~ I I I ~ I= I ~ ~ I I-I h ~ 0I-O z0I-I-Z U 1.06 I ~- "~ I" II '1 ~ \\ I ~ I I 3 ~ i I ~ t * ~ I ~ ~ ~ ~ ~ I ~ I (74.4,1.036)

• t

(90.5,1 t ~

~ ~ ~ , (106. 5,1.046); ~ r .~(134 118.6,1.050)'-:

===. 7,1.0 54) ~ ~ i@i '. 11I* ~ I ~ ~ ~ ~ ~ ~ "(62.3,1 032 1.02 ~ I ,:(46 2,1 025) ~ 'I I ~ ~ h' ~ ~ Wtr I t h h1-

(30.2,1 019)

I ~ h 18;1,1.014) I ~ ~ \\ I I'3 ~ 0 (2.0,1.004) '.'0 I ~ 40 ~ ~ ~ 'I I 60 ~ '\\ ~ I 80 1h ~] 120 '40 D(STANCE FROM BOTTOM OF CORE, INCHES Figure 4.2-1 Augmentation factors vs distance from bottom of core

POWER DISTRIBUTION LIMITS 3/4,2.2 TOTAL PLANAR RADIAL PEAKING FACTORS - F LIMITING CONDITION FOR OPERATION 3.2.2. The calculated value of Fx shall be limited to < l.75. T APPLICABILITY: MODE 1". ACTION: With Fx > 1.75, within 6 hours either: T a. Reduce THERMAL POWER to'ring the combination of THERMAL POWER and Fx to within the limits of Figure 3.2-3 and withdraw the full T length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification

3. 1.3.6; or b.

Be in HOT STANDBY. SURVEILLANCE RE UIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 F -shall be calculated by the expression F = F (1+T ) when T T xy xy q Fx is calculated with a non-full core power distribution analysis code and shall be calculated as F = Fx when calculations are performed with a full T xy xy core power distribution analysis code. F shall be determined to be within xy its limit at the following intervals: a. Prior to operation above 70K of RATED THERMAL POWER after each fuel

loading, b.

At least once per 31 days of accumulated operation in MODE 1, and c. Within 4 hours if the AZIMUTHAL POWER TILT (T ) is > 0.03. q See Special Test Exception

3. 10.2.

ST. LUCIE - UNIT 2 3/4 2-7 Amendment No. B

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued) 4.2.2.3 F shall be determined each time a calculation of F is required by T xy using the incore detectors to obtain a power distribution map 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 between 15K and 85% of full core height and shall .exclude regions influenced by grid effects. 4.2.2.4 T shall be determined each time a calculation of F is vade using T.. xy" a non full core power distribution analysis code. The value of T used in q this case to determine F shall be the measured value of T xy q'T. LUCIE - UNIT 2 3/4 2-8

i'OWER DISTRIBUTION LIMITS TOTAL INTEGRATED RADIAL PEAKING FACTOR - F LIMITING CONDITION FOR OPERATION 3.2.3 The calculated value of Fr, shall be limited to < 1.70. T APPLICABILITY: MODE 18. ACTION: With F > 1.70, within 6 hours either: Tr a. Be in at least HOT STANDBY, or b. Reduce THERMAL POWER to bring the combination of THERMAL POWER and F to within the limits of Figure 3.2-3 and withdraw the full-length r CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THERMAL POWER limit determined from Figure 3. 2-3 shall then be used to establish a revised upper THERMAL POWER level limit on Figure 3.2-4 (truncate Figure 3.2-4 at the allowable fraction of RATED THERMAL POWER determined by Figure 3,2-3) and subsequent operation shall be maintained within the reduced acceptable operation region of Figure 3.2-4.'URVEILLANCE RE UIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 Fr shall be calculated by the expression F = F (1+7 ) when F T Tr r q r is calculated with a non-full core power distribution analysis code and shall be calculated as F =. F when calculations are performed with a Tr: r full core power distribution analysis code. F shall be determined to r be within its limit at the following intervals: a. Prior to operation above 70K of RATED THERMAL POWER after each fuel loading. b. At least once per 31 days of accumulated operation in MODE 1, and c. Within 4 hours if the AZIMUTHAL POWER TILT (T ) is > 0.03. q See Special Test Exception 3. 10.2. ST. LUGIE " UNIT 2 3/4 2-9 Amendment No. 8

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued) r 4.2.3.3 F shall be determined each time a calculation of F is required by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing reactor coolant pump combination. 4.2.3.4 T shall be determined each time a calculation of Fr is made using T a non-full core power distribution analysis code. The value of T used to T q determine Fr in this case shall be the measured value of T q'T. LUCIE - UNIT 2 3/4 2-10

THIS PAGE INTENTIONALLY LEFT BLANK ST. LUCIE - UNIT 2 3/4 2-11 Amendment No. 8

~ ~ 1.0 UNACCEPTABLE OPERATION REGION (-0..08, 1.00) (0.15, 1.00) 0.9 0.8 UJ o 0.7 8 0.6 -0.3.W.~) ACCEPTABLE OP ERATION REGAIN (oa,.o.ao) 0.5 0.4 -0.6 -0.4 -0 ' 0.0 0.2 PERIPHERAL AXIAL SHAPE INDEX Yl 0.4 0.6 FIGURE 3. 2-4 AXIAL SHAPE INDEX OPERATING LIMITS WITH FOUR REACTOR COOLANT PUMPS OPERATING ST. LUGIE UNIT 2 3/4 2-12 Amendment No. f, 8

PARAMETER TABLE 3.2-2 DNB MARGIN LIMITS FOUR REACTOR COOLANT PUMPS OPERATING Cold Leg Temperature (Narrow Range) Pressurizer Pressure Reactor Coolant Flow Rate AXIAL SHAPE INDEX 535 F' T < 549'F 2225 psia~" P < 2350 psia" PZR- > 363,000 gpm Figure 3.2-4 Applicable only if power level > 70K RATED THERMAL POWER. Limit not applicable during either a THERMAL POWER ramp increase in excess of 5X of RATED THERMAL POWER or a THERMAL POWER step increase of greater than 10K of RATED THERMAL POWER. ST. LUCIE - UNIT 2 3/4 2"15 Amendment No. 8

TABLE 3. 3-1 Continued) ACTION STATEMENTS ACTION 2 (Continued) 6. Cold Leg Temperature 7. Hot Leg Temperature Variable Power Level - High (RPS) Thermal Margin/Low Pressure (RPS) Local Power Oensity - High (RPS) Variable Power Level - High (RPS) Thermal Margin/Low Pressure (RPS) Local Power Oensity - High (RPS) ACTION 3 ACTION 4 ACTION 5 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes. Verify compliance with the SHUTOOWN MARGIN requirements of Specifica-tion'. l.l. 1 or 3. 1. 1.2, as applicable, within I hour and at least once per 12 hours thereafter. 'ith the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the reactor trip breakers of the inoperable channel are placed in the tripped condition wjthin 1 hour, otherwise, be in at least HOT STANOBY within 6 hours;

however, one channel may be bypassed for up to 1 hour, provided the trip breakers of any inoperable channel are in the tripped condition, for surveillance testing per Specification 4.3. l. l.

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement restore. the inoperable channel to OPERABLE status within 48 hours or open the. reactor trip breakers within the next hour. ST. LUCIE - UNIT 2 3/4 3-5

TABLE 3.3-2 lC:nM M FUNCTIONAL UNIT 1. Manual Reactor Trip 2. Variable Power Level - High

RESPONSE

TIME Not Applicable < 0.40 second"'"* REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES 3. Pressurizer Pressure - High 4. Thermal Margin/Low Pressure < 1.15 < 0.90 seconds second"* 5. Containment Pressure High 6. Steam Generator Pressure - Low 7. Steam Generator Pressure Difference - High 8. Steam Generator Level Low 9. Local Power Density High < l.l5 seconds < 1.15 seconds < 1.15 seconds < 1.15 seconds < 0.40 second*'"" O CO

I M m FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUHENTATION TRIP VALUES M 0 1. SAFETY INJECTION (SIAS) a. Hanual (Trip Buttons) b. Containment Pressure High c. Pressurizer Pressure Low d. Automatic Actuation Logic 2. CONTAINMENT SPRAY (CSAS) a. Hanual (Trip Buttons) b. Containment Pressure -- High-High c. Automatic Actuation Logic 3. CONTAINHENT ISOLATION (CIAS) a. Hanual CIAS (Trip Buttons) b. Safety Injection (SIAS) c. Containment Pressure High d. Containment Radiation - High e. Automatic Actuation Logic 4. HAIN STEAH LINE ISOLATION a. Hanual (Trip Buttons) b. Steam Generator Pressure Low c. Containment Pressure - High d. Automatic Actuation Logic Not Applicable 3.5 pslg > 1736 psia Not Applicable Not Applicable < 5.40 psig Not Applicable Not Applicable Not Applicable < 3.5 psig < 10 R/hr Not Applicable Not Applicable > 600 psia < 3.5 psig Not Applicable Not Applicable < 3.6 psig > 1728 psia Not Applicable Not Applicable < 5.50 psig Not Applicable Not Applicable Not Applicable < 3.6 psig < 10 R/hr Not Applicable Not Applicable > 567 psia < 3.6 psig Not Applicable

I A M m FUNCTIONAL UNIT TRIP VALUE ALLOWABLE VALUES TABLE 3. 3-4 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES M CONTAINMENT SUMP RECIRCULATION (RAS) a. Manual RAS (Trip,Buttons) b. Refueling Water Storage Tank Low c. Automatic Actuation Logic 6. LOSS OF POWER Not Applicable 5.67 feet above tank bottom Not Applicable Not Applicable 4.62 feet to 6.24 feet above tank bottom Not Applicable 4J I CO a. b. (1) 4. 16 kV Emergency Bus Undervoltage (Loss of Voltage) (2) 480 V Emergency Bus Undervoltage (Loss of Volta'ge) (1) 4. 16 kV Emergency Bus Undervoltage (Degraded Voltage) > 3120 volts > 360 volts > 3848 volts with a 10-second time delay > 3120 volts > 360 volts > 3848 volts with a 10-second time delay d. e. Feedwater Header High hP (2) 480 V Emergency Bus Undervoltage (Degraded Voltage) AUXILIARYFEEDWATER (AFAS) a. Manual (Trip Buttons) b. Automatic Actuation Logic c. Steam Generator hP-High SG 2A82B Level Low > 432 volts Not Appli cable Not Applicable < 180.0 psid > 20.6X '< 100.0 psid > 432 volts Not Applicable Not Applicable < 187.5 psid > 20.0X < 107.5 psid

TABLE 3.3-5 ~ ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS 1. Manual a ~ b. C. SIAS Safety Injection (ECCS) Containment Isolation- .(') Shield Building Ventilation System Containment Purge Valve Isolation Containment Fan Coolers CSAS Containment Spray Iodine Removal CIAS Containment Isolation Shield Building Ventilation System Containment Purge Valve Isolation'ot Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable d. MSIS Main Steam Isolation Feedwater Isolation e. RAS Containment Sump Recirculation AFAS Auxiliary Feedwater Actuation Feedwater Isolation Not Applicable ~ Not Applicable Not Applicable Not Applicable Not Applicable ST. LUCIE UNIT 2 3/4 3-19

TABLE 3. 3-5 Continued) ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS 2. Pressurizer Pressure-Low a ~ b. C. d. Safety Injecti on (ECCS) Containment Isolation Shield Building Ventilation System Containment Fan Coolers Charging Flow 3P Pk/20 Pk* 21 75'8/1 1 758 Jc < 26.0~/10.0~" < 24. 15~/ll. 15" < 330.00~/180.00~ 3. Containment Pressure-Hi h a. Safety Injection (ECCS) b. Containment Isolation c. Shield Building Ventilation System d. Containment Fan Coolers e. Feedwater Isolation f. Main Steam Isolation 3P PA/20 PAR < 21.75~/11.75"" < 26.0'/10.0"'

24. 15"/ll.15 "~

< 5.15"/5.15"" 6 75)k/6 75AR a. Containment Spray/Iodine Removal 5. Containment Radiation-Hi h a. Containment Isolation b. Shield Buildirg Ventilation System" 6. Steam Generator Pressure-Low a. Feedwater Isolation b. Main Steam Isolation 7. Refuel'in Mater Stora e Tank-Low a. Containment Sump Recirculation < 25.65~/11.15"~ < 26.75'/16.75" < 32.75"/16.75~~ < 5.15/5.l5 ' 6.75/6.75 " < ill.15"/101. 15"" 8. 4.16 kV Emer enc Bus Undervolta e (Loss of Volta e) a. Loss of Power (4.16 kV) < 14 b. Loss of Power (480 V) < 14 9.

4. 16 kV Emer enc Bus Undervolta e

(De raded Volta e) a. Loss of Power (4. 16 kV) b. Loss of Power (480 V) < 12 < 22 ST. LUCIE - UNIT 2 3/4 3"20 Amendment No.8 I

I ~ TABLE 3. 3-5 Continued) ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

T'IME IN SECONDS 10. Steam Generator Level-Low a. Auxi 1 iary Feedwater b. Feedwater Isolation < 120"/120*" ( 5 l5A/5 l5)k* Feedwater Header DP a. Auxiliary Feedwater Feedwater Isolation < 120"/120** 5 15"/5 15 12. Steam Generator b,P a. Auxiliary Feedwater b. Feedwater Isolation < 120"/120~" 5 ]5/c/5 ]5AR NOTE:

Response

time for Motor-Dr iven and Steam-Dri ven Auxi 1 iary Feedwater Pumps on all AFAS signal starts 120. 0 TABLE NOTATION Diesel generator starting and sequence loading delays included.

Response

time limit includes movement of valves and attainment of pump or blower discharge pressure. Diesel generator starting and sequence loading delays not included. Offsite power available.

Response

time limit.includes movement of valves and attainment of pump or blower discharge pressure. Containment Isolation response time is applicable to the valves specified (1) in Specification 3.6.3. ST. LUCIE - UNIT 2 3/4 3-21 Amendment No. 8

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAION SURVEILLANCE RE UIREMENTS m FUNCTIONAL UNIT 1. SAFETY INJ a. Manua ECTION (SIAS) 1 (Trip Buttons) b. Containment Pressure High c. Pressurizer Pressure - Low d. Automatic Actuation Logic CHANNEL CHECK N. A. S S N. A. CHANNEL CALIBRATION N.A. R R N. A. CHANNEL FUNCTIONAL TEST R M M(l), SA(2) MODES FOR, WHICH SURVEILLANCE IS RE UIRED 1, 2, 3, 4 1, 2, 3 1, 2, 3 1,2,3,4 2. CONTAINMENT SPRAY (CSAS) a. Manual (Trip Buttons) b. Containment Pressure-- High High c. Automatic Actuation Logic N.A. S N. A. N. A. R N.A. M M(l), SA(2) 1, 2, 3, 4 1,2,3 1,2,3,4 3. CONTAINMENT ISOLATION (CIAS) Manual CIAS (Trip Buttons) b. Safety Injection SIAS c. Containment Pressure High d. Containment Radiation - High e. Automatic Actuation Logic 4. MAIN STEAM LINE ISOLATION a. Manual (Trip Buttons) b. Steam Generator Pressure Low c. Containment Pressure High d. Automatic Actuation Logic N.A. N. A. S S N.A. N. A. S S N. A. N. A. N. A. R R N.A. N.A. R R N. A. R R M M M(1), SA(2) R M M M(l), SA(2) 1,2,3,4 1, 2, 3, 4 1, 2, 3 1, 2, 3 1, 2, 3, 4 1 2 3 1,2,3 1 2 3 1,2,3,4 5. CONTAINMENT SUMP RECIRCULATION (RAS) a. Manual RAS (Trip Buttons) b. Refueling Water Storage Tank - Low c. Automatic Actuation Logic N. A. S N. A. N.A. R N. A. 'M M(l), SA(2) N.A. 1,2,3 1,2,3

REACTOR COOLANT SYSTEM 3/4. 4. 3 RESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a minimum water level of greater than or equal to 27K indicated level and a maximum water level of less than or equal to 68K indicated level and at least two groups of pressurizer heaters capable of being powered from 1E buses each, having a nominal capacity of at least 150 kW. APPLICABILITY: MODES 1, 2, and 3. ACTION: 'a ~ With one group of the above required pressurizer heaters inoperable, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE RE UIREMENTS 4.4.3.1 The pressurjzer Rater volum'e shall be determined to be within its limits at least once per 12 hours. 4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified to be at least 150 kW at least once per 92 days. 4.4.3.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by verifying that on an Engineered Safety Features Actuation test signal concurrent with a loss of offsite,power: a. the pressurizer heaters are automatically shed from the emergency power sources, and b. the pressurizer heaters can be reconnected to their respective buses manually from the control room. ST. LUCIE - UNIT 2 3/4 4-9 Amendment No. 8

REACTOR COOLANT SYSTEM 3/4.4;4 PORV BLOCK VALVES LIMITING CONDITION FOR OPERATION 3.4.4 Each Power Operated Relief Valve (PORV) Block valve shall be OPERABLE. No more than one block valVe shall be open at any one time. APPLICABILITY: MODES 1, 2, and 3. ACTION: a. With. one or more block valve(s) inoperable, within 1 hour either restore the block valve(s) to OPERABLE status or close the block valve(s) and remove power from the block valve(s); otherwise, be,in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. b. With both block valves

open, close one bloCk valve within 1 hour, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

c. The provisions of specification 3.0.4 are not applicable. SURVEILLANCE RE UIREMENTS 4.4.4 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of Action a..or b. above. ST. LUCIE - UNIT 2 3/4 4-10

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: a. With 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 hours, 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 hours and in COLD SHUTDOWN within the following 30 hours. b. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 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 4.7-0, in accordance with Section XI of the ASME Boi.ler and Pressure Vessel

Code, 1974 Edition.

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

TABLE 3.7-1 MAXIMUMALLOWABLE POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS Maximum Number of Inoperable Safety Valves on An 0 eratin Steam Generator Maximum Allowable Power Level-High Trip Setpoint Percent of RATED THERMAL POWER 92. 8'6.3

VALVE NUMBER TABLE 4.7-0 STEAM LINE SAFETY VALVES PER LOOP LIFT SETTING 1% ORIFICE SIZE Header A Header B a ~ b. C. d. e. g. h. 8201 8202 8203 8204 8209 8210 8211 8212 8205 8206 8207 8208 8213 8214 8215 8216 1000 psia 1000 psia 1000 psia 1000 psia 1040 psia 1040 psia 1040 psia 1040 psia 16 in. 2 16 in. 2 16 in. 16 in'6 in. 16 in. 2 16 in. 2 16 in.

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING.CONDITION FOR OPERATION

3. 7. 1. 2 At least three independent steam generator auxi liary feedwater pumps and associated flow paths shall be OPERABLE~ with:

a. b. Two feedwater

pumps, each capable of being powered from separate OPERABLE emergency

busses, and One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3. ACTION: a 0 b. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours. c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status. SURVEILLANCE RE UIREMENTS 4.7.1.2 a. Each auxiliary 'feedwater pump shall be demonstrated OPERABLE: At 1 east once per 31 days by: 1. Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to 1270 psig on recirculation flow. 2. 3. Verifying that the turbine-driven pump develops a discharge pressure of greater than or equal to 1260 psig on recirculation flow when the secondary steam supply pressure is greater than 50 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3. ~ 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.

The Auxiliary Feedwater System automatic initiation system shall be completely installed and OPERABLE prior to initial criticality. ST. LUCIE - UNIT 2 3/4 7-4

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION '3..7. 1.5 Each main steam line isolation valve shall be OPERABLE'PPLICABILITY: MODES 1, 2, 3 and 4. ACTION: MODE 1 MODES 2, 3-and 4 With one main steam line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 24 hours. With one main steam line isolation valve inoperable, subseqent operation in MODES 2, 3 or 4 may proceed provided: a. The isolation valve is maintained closed. b. The provisions of Specification

3. 0. 4 are not applicable.

Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE RE UIREMENTS 4.7. 1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by: a. Part-stroke exercising the valve at least once per 92 days, and b. Verifying full closure within 5.6 seconds on any closure actuation signal while in HOT STANDBY with T > 515 F during each reactor shutdown except that verification o7 full closure within 5.6 seconds need not be determined more often than once per 92 days. ST. LUCIE - UNIT 2 3/4 7-9

~ 4 1 PLANT SYSTEMS MAIN FEEDWATER LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7. 1. 6 Each main.feedwater line isolation valve shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: MODE 1 MODES 2, 3-and 4 With one main feedwater line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 24 hours. With one main feedwater line isolation valve inoperable, subseqent operation in MODE 2, 3, or 4 may proceed provided: a. The isolation valve is maintained closed. b. The pr ovisions of Specification 3.0.4 are not applicable. Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the foll'owing 24 hours. SURVEILLANCE RE UIREMENTS 4.7. 1.6 Each main feedwater line isolation valve shall be demonstrated OPERABLE by: Part-stroke exercising the valve at least once per 92 days, and Verifying full closure within 5.15 seconds on any closure actuation signal while in HOT STANDBY with Tavg > 515'F during each reactor shutdown except that verification of full closure within 5.15 seconds need not be determined more often than once per 92 days. ST. LUCIE - UNIT 2 3/4 7-10 Amendment No. B

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 wi 11 be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T The most restrictive avg'ondition occurs at EOL, with T at no load operating temperature, qnd is avg associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 5.0X delta k/k is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. At earlier times in core life, the minimum SHUTDOWN MARGIN required for the most restric-tive conditions is less than 5.0X hk/k. With T less than or equal to 200'F, avg the reactivity transients resulting from any postulated accident are minimal and a 3X delta k/k SHUTDOWN MARGIN provides adequate.protection. 3/4. l.l. 3 BORON DILUTION A minimum flow rate of at least 3000 gpm provides adequate

mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System.

A flow rate of at least 3000 gpm will circulate an equivalent Reactor Coolant System volume of 10,931 cubic feet in approximately 26 minutes. The reactivity'hange rate associated with boron concentration reductions will therefore be within the capability of operator recognition and control. 3/4. l. 1.4 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurances that the coef-ficient will be maintained within acceptable values throughout each fuel cycle. ST. LUCIE - UNIT 2 B 3/4 1-1 Amendment No. 8

REACTIVITY CONTROL SYSTEMS BASES. 3/4. 1. 1: 5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 515'F. This limitation is required to ensure (1) the moderator temperature coefficient is within its analyzed temperature

range, (2) the protective instrumentation is within its normal operating

range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor pressure vessel is above its minimum RTNDT temperature.

3/4. 1. 2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include (1) borated water

sources, (2) charging

pumps, (3) separate flow paths, (4) boric acid makeup

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 boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 3.OX delta k/k after xenon decay and cooldown to 200 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires boric acid solution from the boric acid makeup tanks in the allowable concentrations and volumes of Specification

3. 1. Z. 8 or 72,000 gallons of 1720 ppm - 2100 ppm borated water from the refueling water tank.

With the RCS temperature below 200 F one injection system is acceptable without single fai lure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable. The boron capability required below ZOO~F is based upon providing a 3X delta k/k SHUTDOWN MARGIN after xenon decay and cooldown from 200'F to 140'F. This condition requires either 4,150 gallons of 1720 ppm - 2100 ppm borated water from the refueling water tank or boric acid solution from the boric acid makeup tanks in accordance with the requirements of Specification

3. 1. Z. 7.

The contained water volume limits includes allowance for water not available because of discharge l.ine location and other physical characteristics. The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. The limits on contained water volume and boron concentration of the RWT also ensure a pH value of between

7. 0 and 11. 0 for the solution recirculated within containment after a

LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. ST. LUCIE - UNIT 2 B 3/4 1-2 Amendment No. B

REACTIVITY CONTROL SYSTEMS

BASES, 3/4. l. 2. 9 BORON DILUTION The simultaneous use of the boronometer and RCS sampling at intervals dependent upon the MODE and the number of OPERABLE charging pumps to monitor the RCS boron concentration provides diverse and redundant indications of an inadvertent boron dilution.

This wil] allow detection with sufficient time for termination of the boron dilution event before a complete loss of SHUTDOWN MARGIN occurs. 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 main-

tained, and (3) the potential effects of CEA misalignments 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 design criteria are met. \\ The ACTION statements applicable to a stuck or untrippable CEA, to two or more inoperable CEAs and to a large misalignment (greater than or equal to 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 untrippab',e CEA, the loss of SHUTDOWN MARGIN. For small misaliqnments (less than 1.5 inches) of the CEAs, there is (1) a small effect on the tame-dependent long-term power distributions relative to those used in generating LCOs and LSSS setpoints, (2) a small effect on the available SHUTDOWN MARGIN, and (3) a small effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with small misalignments of CEAs permits a 1-hour time interval during which attempts may be made to restore the CEA to within its alignment requirements. The 1-hour time limit is sufficient to (1) identify causes of a misaligned

CEA, (2) take appropriate corrective action to realign the

CEAs, and (3j minimize the effects of xenon redistribution.

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

REACTIVITY CONTROL'YSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued} 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. This distribution may, in turn, have a significant effect on (1) the available SHUTDOWN MARGIN, (2) the time- 'ependent long-.term power distributions relative to those used in generating LCOs and LSSS setpoints, and (3) the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt 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 advers'e 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. The r~quirement to reduce power in certain time limits depending upon the previous Fr is to eliminate a potential nonconservatism for situations when a CEA has been declared inoperable. A worst-case analysis has shown that a DNBR SAFDL violation may occur during the second hour after the CEA misalignment if this'equirement is not met. This potential DNBR SAFDL violation is eliminated by limiting the time operation is permitted at full-'power before power reductions are required. These reductions will be necessary once the deviated CEA has been declared inoperable. This time allowed to continued operation at a reduced,, power level can be permitted for the following reasons: 2. 3. 4 5. The margin calculations which support the Technical Specifications are based on a steady-state radial peak of F = 1.7 0. r When the actual Fr < 1.70, significant additional margin exists. T This additional margin can be credited to offset the'increase in F with time that can occur following a CEA misalignment. This increase in Fr is caused by xenon redistribution. T The present analysis can support allowing a misalignment to exist for up to 63 minutes without correction, if the initial F< 1.54. ST. LUCIE - UNIT 2 8 3/4 1-4 Amendment No. 8

\\ 3/4.2 POWER DISTRIBUTION LIMITS s BASES 0 3/4;2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a

LOCA, the peak temperature of the fuel cladding will not exceed 2200'F.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provides adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3. 2-2. In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made: (1) the CEA insertion limits of Specifications

3. 1.3.5 and 3. 1 ~ 3.6 are satisfied (2) the flux peaking augmentation factors are as shown in Figure 4.2-1, (3) the AZIMUTHAL POWER TILT restrictions of Specifica-tion 3.2.4 are satisfied, and (4) the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2. 2.

The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1. The setpoints for these alarms include allowances, set in the conservative directions, for (1) flux peaking augmentation factors as shown in Figure 4.2-1, (2) a measurement-calculational uncertainty factor of 1.062, (3) an engineering uncertainty factor of 1.03, (4) an allowance of 1.01 for axial'uel densification and thermal expansion, and (5) a THERMAL POWER measurement uncertainty factor of 1.02. 3/4.2.2 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING FACTORS - F AND F AND AXIMUTHAL POWER TILT - T The limitations on F and T are provided to ensure that the assumptions used in the analysis for establishing the Linear Heat Rate and Local Power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. The limitations on F'nd T T. are provided to ensure that the assumptions used ia the analysis establishing the DNB Margin LCO, the Thermal Margin/Low Pressure LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. If F F or T exceed their basic limitations, operation may continue under xy' q the additional restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the ST. LUCIE - UNIT 2 B 3/4 2-1 I

POWER DISTRIBUTION LIMITS BASES assumptions used in establishing the 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 requi rement that the measured value of Tq be mutiplied by the calculated values of F and F to determine F and F is applicable only T r xy r xy when F and F are calculated with a non-full core power distribution analysis Xy code. When monitoring a reactor core power distribution, F or F with a full r xy core power distribution analysis code the azimuthal tilt is explicitly accounted for as part of the radial power distribution used to calculate Fx and Fr'he Surveillance Requirements for verifying that F F and T are xy' q within their limits provide assurance that the actual values of F F and T xy' q do not exceed the assumed values. Verifying F and F after each fuel T Tr loading prior to exceeding 75% of RATED THERMAL POWER provides additional assurance that the core was properly loaded. 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 tran'sient and safety analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DHBR of > 1 ~ 28 throughout each analyzed transient. l The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the 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-hour basis. ST. LUCIE - UNIT 2 B 3/4 2-2 Amendment No. 8

THIS PAGE INTENTIONALLYLEFT BLANK ST. LUCIE - UNIT 2 B 3/4 2-3 Amendment No. 8

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1100 psia) of its design pressure of 1000 psia during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser). The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Vessel

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

The to]al relieving capacity for all valves on all of the steam lines is 12.49 x 10 lbs/hr which is 103.8% of the total secondary steam flow of 12.03 x 10 lbs/hr at 100% RATED THERMAL POWER. A minimum of two 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 set-point reductions are derived on the following bases.: For two loop operation: SP = ( ) x (107.0) - 0.9 where: SP 107.0 0.9 reduced reactor trip setpoint in percent of RATED THERMAL POWER maximum number of inoperable safety valves per steam line Power Level-High Trip Setpoint for two loop operation Equipment processing uncertainty Total relieving capacity of all safety valves per steam line in lbs/hour (6.247 x 106 lbs/hr) Maximum relieving capacity of any one safety valve in lbs/hour (7.74 x 106 lbs/hr) ST. LUCIE - UNIT 2 B 3/4 7-1 Amendment No. 8

PLANT SYSTEMS BASES 3/4.7. 1.2 AUXILIARY-FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss-of-offsite power. Each electric-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 3ZO gpm at a pressure of 1000 psia to the entrance of the steam generators. The steam-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 500 gpm at a pressure of 1000 psia to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350~F when the shutdown cooling system may be placed into operation. 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 to maintain the Unit 2 RCS at HOT STANDBY conditions for 4 hours followed by an orderly cooldown to the shutdown cooling entry temperature (350'F). The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. The actual water requirements are 149,600 gallons for Unit 2 and 125,000 gallons for Unit 1. Included in the requi red volumes of water are the tank unusable volume of 9400 gallons and a conservative allowance for instrument error of 21,400 gallons. ST. LUCIE - UNIT 2 8 3/4 7-2

5. 0 DESIGN FEATURES

5. 1 SITE EXCLUSION AREA

5. l. 1 The exclusion area shall be as shown in Figure 5. 1-1.

ll LOW POPULATION ZONE

5. 1.2 The low population zone shall be as shown in Figure

5. 1-1.

I

5. 2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel building of cylindrical

shape, with a dome roof and having the following design features:

a. Nominal inside diameter = 140 feet. b. Nominal inside height = 232 feet. c. Net free volume = 2.506 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.

5. 2. 1. 2 SHIELD BUIL'DING a.

Minimum annular sp'ace'= 4 feet. b. Annulus nominal volume = 543,000 cubic feet. c. Nominal outside height (measured from top of foundation mat to the top of the dome).= 228.5 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 steel reactor containment building is designed and shall be maintained for a maximum internal pressure of 44 psig and a temperature of 264 F. ST. LUCIE - UNIT 2 5-1 Amendment No,.

I I Vl 'W 42 bttg ln ~e O 42 r a4 Ul Ule OO <i lU Cl O 222 2a 0 e. FP a L's PROPERTY LINE , l/ 8L IND CR Cl r CVO Ol r ~ TI ~ 40 C r nt 0 "rrr1l 'e, ro I( Inset Detail C. +E'EK rl ~ I Q) l 0'Pearl I EXCLUSION AREA (0.97m2) ANO LOW POPULATION 20NE (I mi) UNIT I UNIT 2 'OTES:

1) L-Liquid Radwaste Release Point 2)

Due to the scale of the Figure the Exclusion Area Radius (0.97 mile) and the Low Population 2one (1 mile) are shown as being the same size. 3) The following are unrestricted areas: a) All waterways (Herman Bay, Big Mud

Creek, Indian River) with the exception of the intake and discharge canal.

b) Ocean front beaches c) State Road AIA rp ~ I/2 0 SCALE IN MILES 5-2 ST. LUCIE - UNIT 2

I f;I~~~

FpaL s PROPERTY LINE \\ ~ 0 Vl Ul Cl 0 o O 4 s ~I FLORIDA POWER 8 LIGHT COMPANY ST. LUCIE PLANT SITE AREA MAP FIGURE S.i-i

DESIGN fEATURES

5. 3 REACTOR CORE FUEL ASSEMBLIES e

5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing 236 fuel and poison rod locations. All fuel and poison <ods are clad wi th Zircaloy-4. Each fuel rod shall have a nominal active fuel length o 136.7=inches and contain approximately 1700 grams uranium. The initial core loading shall have a maximum enrichment of 2.73 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.70 weight percent U-235. CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain gl full-length control element assemblies and no part-length control element assemblies. 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE

5. 4.

1 The Reactor Coolant System is designed and shall be maintained: a. b. C. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, For a pressure of 2485 psig, and For a temperature of 6500F, except. for the pressurizer which is 700 F. ST. LUCIE " UNIT 2 5-3 Amendment No.8

DESIGN FEATURES VOLUME 5,4,2 The total water and steam volume of the reactor coolant system is 10,931 + 275 cubic feet at a nominal T of 572'F. 5.5 METEOROLOGICAL TOWER LOCATION 5.5; 1 The meteorological tower shall be located as shown on Figure 5.1-1. 'I 5.6 FUEL STORAGE CRITICALITY 5.6.1 ~ a. The spent fuel storage racks. are designed and shall be maintained with: 1 A k ff equi val ent to 1 ess than or equal to 0. 95 when f1 ooded witk unborated water, which includes a conservative allowance of 0.024 ak ff for Total Uncertainty. eff 2. A nominal 8.96 inch center-to-center distance between fuel assemblies placed in the storage racks. 3. A boron concentration greater than or equal to 1720 ppm. Region I can be used to store fuel which has a U-235 enrichment less than or equal to 4.5 weight percent. Region II can be used to store fuel which has achieved sufficient burnup such that storage in Region I is not required. The initial enrichment vs. burnup requirements of Figure 5.6-1 shall be met prior to storage of fuel assemblies in Region I.I. b. The new fuel storage racks are designed for dry storage of unirradiated fuel assemblies having a U-235 enrichment less than or equal to 4.5 weight percent, while maintaining a k +f of less than or equal to 0.98 under the most reactive conditifh. DRAINAGE 5.6.2 The spent fuel storage 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 storage pool is designed and shall be maintained with a storage capacity limited to no more than 1076 fuel assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. ST. LUCIE-UNIT 2 5-4 Amendment No ~}}