ML17206A579

Amend 28 to License DPR-67,changing Tech Specs to Increase Min Required Volume of Water in Refueling Water Tank

ML17206A579
Person / Time
Site:	Saint Lucie
Issue date:	12/14/1978
From:	Reid R Office of Nuclear Reactor Regulation
To:
Shared Package
ML17206A580	List: ML17206A579 ML17206A581 ML17206A582
References
NUDOCS 7901020043
Download: ML17206A579 (24)

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4y*y4 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20566 FLORIDA POWER 8

LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

28 License No.

DPR-.67 l.

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

A.

The applications for amendment by Florida Power 8 Light Company (the licensee) dated February 24 and 27, 1978, comply 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 applications, the provisions of the Act, and the rules and regulations of the Comission; 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 coIIIIon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CPR Part 51 of the COIIission's regulations and all applicable requirements have been satisfied.

voodoo>oo't>

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C 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated 'in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.

DPR-67 is hereby amended to read as follows:

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Technical S ecifications I

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 28, are hereby incorporated in the license.

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

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications (Jd 8, +.c ober t M.

Re s d, Chi e Operating Reactors Branch F4 Division of Operating Reactors Date of Issuance:

December 14, 1978

ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. DPR-67 DOCKET NO. 50-335 Replace the following pages of tbe Appendix '!A" Technical Specifications with the enclosed pages.

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

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

~Pa es V

1~7 3/4 1-18 3/4 4-23a 3/4 4-'23b 3/4 4-23c 3/4 5-3 3/4 5-7 3/4 5-8 B 3/4 1-2 B 3/4 4-2 B 3/4 7-2

LIMITING CONDITION FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4.4.4 PRESSURIZER...............................

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

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE....

Leakage Detection Systems,........

Reactor Coolant System Leakage....

4-5 3/4 4-12

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3/4 4-12

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~ ~ 3/4 3/4.4.7 CHEMISTRY...............,..............................

3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY.................

3/4.4.9 PRESSURE/TEMPERATURE LIMITS.......

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~ 3/4 4-17 4-21 Reactor Coolant System.....................;..........

3/4 4-21 Pressurszer...........................................

3/4 4-25 3/4.4.10 STRUCTURAL INTEGRITY..............

Safety Class 1 Components.........

Safety Class 2 Components.:.......

Safety Class 3 Components.........

3/4.4.11 CORE BARREL MOVEMENT..............

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4-53 4-56 3 4.5 EMERGENCY CORE COOLING SYSTEMS ECCS) 3/4.5.1 3/4.5.2 3/4.5.3 3/4.5.4 ECCS SUBSYSTEMS - T

> 325'F....

ECCS SUBSYSTEMS - T

< 325'P....

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

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3/4 5-1 ST.

LUCIE - UNIT 1 Amendment No. 28

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3 4.6 CONTAINMENT SYSTEMS 3/4.6.1 3/4.6.2 3/4.6.3 3/4.6.4 3/4,6.5 3/4.6.6 Air Temperature......................

Containment Vessel Structural Integri DEPRESSURIZATION AND COOLING SYSTEMS.

Containment Spray System.............

Spray Additive System................

Containment Cooling System...........

CONTAINMENT ISOLATION VALVES.........

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

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Hydrogen Analyzers...................

Electric Hydrogen Recombiners W...

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

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SECONDARY CONTAINMENT............-...................

Shield Building Ventilation System...

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

Shield Building Structural Integer ity.

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CONTAINMENT VESSEL.....................

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

Containment Leakage...........................

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

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Internal Pressure.................

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3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-10

'3/4 6-12 3/4 6-13 3/4 6-14 3/4 6-15 3/4 6-15 3/4 6-16a 3/4'-17 3/4 6-18 3/4 6-23 3/4 6-23 3/4 6-24 3/4 6-26 3/4 6-27 3/4 6-27 3/4 6-30 3/4 6-31 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE....................

afety Valves...................-"

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3/4 3/4 7-1 7-1 Auxiliary Feedwat'er System........

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

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,~ o ~ ~ ~ ~ ~ ~ oo ~ o ~ ~ oooo' Main Steam Line Isolation Valves..

Secondary Water Chemistry.........

ST.

LUCIE - UNIT 1

VI

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3/4 3/4 3/4 3/4 3/4 Amendment 7-4 7-6 7-7 7-9 7-10 No. 27

TABLE 1.1 OPERATIONAL NODES MODE 2.

STARTUP 3.

HOT STANDBY 4.

HOT SHUTDOWN

> 0.99

< 0.99

< 0.99 REACTIVITY CONDITION, K ff eff POWER OPERATION

> 0.99

%RATED THERMAL POWER*

5%

50/

AVERAGE COOLANT TEMPERATURE

> 325'F

> 325'F

> 325'F 325'F T

200'F 5.

COLD SHUTDOWN 6.

REFUELING""

< 0.99

< 0.95

<,200'F

< 140'F xc u

>ng ecay heat.

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

ST.

LUCIE - UNIT 1

1-7 Amendment No.

TABLE 1.2 FRE UENCY NOTATION NOTATION D

SA S/U N.A.

FRBRUENCY At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

At least once per 7 days.

At least once per 31 days.

At least once per 92 days.

At least once per 6 months.

At least once per 18 months.

Prior to each reactor startup.

Not applicable.

ST,.

LUCIE - UNIT 1 1-8

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TEMPERATURE( F)

'REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 At least two of the following three borated water sources shall be OPERABLE:

a

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Two boric acid makeup tanks and-one associated heat tracing circuit with the contents of the tanks in accordance with Figure 3.1-1, and The refueling water tank with:

l.

A minimum contained volume of 401,800 gallons of water, 2.

A minimum boron concentration of 1720 ppm, 3.

A maximum solution temperature of 100'F, 4.

A minimum solution temperature of 55'F when in NODES 1 and 2, and 5.

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

APPLICABILITY:

NODES 1, 2, 3 and 4.

ACTION:

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

', 'SURVE'IL'LANCE RE UIREMENTS 4.1.2.8 At least two borated water sources shall be demonstrated OPERABLE:

a.

At least onceper 7 days by:

1.

Verifying the boron concentration in each water source, T. LUCIE - UNIT 1 3/4 1-18 Amendment No. 28

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MINIMUMPRESSURE TEMPERATURE FoR. )1) 1!ltrlll~tfittj):1l'jttttfi.+adjt G CRITICALOPERATION OF CORE

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NOTE 1 REACTOR VESSEL BELTLINEMATERIAL INITIALRTNDT 5 F

NOTE I REACTOR VESSEL BELTLINEMATERIAL

5 YEAR RTNDTSHIFTEP NOTE I REMAINiNGPRESSURE BOUNDARY MAXIMUMRTNDT 50 4

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FIGURE 3.4t2a Amendment No. 77,$ 8 3/4 4-23a ST.

LUCIE - UNIT 1 Reactor Coolant System Pressure Temperature Limitations for up to 5 Years of Full Power Operation

III Ili}

I I:li i:::::."I::: ': MINIMUMPRESSURE TEMPERATURE FOR

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100 200 300 400 500 INDICATEDREACTOR COOLANTTEMPERATURE Tc.

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'I FIGURE 3.4.2b Reactor Coolant System Pressure Temperature Limitations for up to 10 Years of Full Power OperatIOn ST.

LUCIE - UNIT 1 3/4 4-23b, Anendment No. 77 ~8

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-iNOTE 1 REACTOR VESSEL BELTUNE MATERIAL INITIALRTNoT ~ 5 F

NOTE 2-REACTOR VESSEL BELTLINEMATERIAL 40 YEAR RTIIoT SHIFT

NOTE 3 REMAININGPRESSURE BOUNDARY XIMUMRTIIoT E0 F 4T 44

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-'.'AXIMUMPRESSURE =

..FOR SDC OPERATION MINIMUMPRESSURE TEMPERATURE FOR:

iCRITICALOPERATION OF CORE

~iNON4;RITICAL OPERATION OF CORE 0

100 200 300

'00 INDICATEDREACTOR COOLANTTEMPERATURE T~

F FIGURE 3.4.2c React'or Coolant System Pressure Temperature Limitations for up to 40 Years of Full Power Operation ST.

LUCIE - UNIT 1

3/4 4-23c Amendment 'No. 77,

TABLE 4.4-5 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE SPECIMEN REMOVAL INTERVAL 8 years 2.

16 years 3.

23 years 4.

30 -years 5.

6.

35 years 40 years

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T

> 325'F LIMITING CONDITION FOR OPERATION 3..5.2 Two independent ECCS subsystems shall be OPERABLE with each sub-system comprised of:

a.

One OPERABLE high-pressure safety injection (HPSI) pump (one ECCS subsystem shall include HPSI pump A and the second ECCS subsystem shall include either HPSI pump B or C),

b.

One OPERABLE low-pressure safety injection pump, and c.

An independent OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal.

APPLICABILITY:

MODES 1, 2 and 3*.

ACTION:

a.

b.

With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In the event the ECCS is actuated

'and injects water into. the Reactor Coolant System, a Special Report shall. be prepared and submitted to the Commissi.on pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

"Wit pressurizer pressure

> 1750 psia.

ST.

LUGIE - UNIT 1

3/4 5-3 Amendment No. $8'

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number 1.

V-3659 Valve Function 1,

Mini-flow isolation Valve Position 1.

Open 2.

V-3660 2.

Mini-flow isol ati on 2.

Open b.

At least once per 31 days on a

STAGGERED TEST BASIS by:

l.

Verifying that each high-pressure safety injection pump:

a)

Starts (unless already operating) from the control room.

b)

Develops a discharge pressure of > 1138 psig on recirculation flow.

c)

Operates for at least 15 minutes.

2, Verifying that each low-pressure safety injection pump:

a)

Starts (unless already operating) from the control room.

b)

Develops a discharge pressure of > 175 psig on recirculation flow.

c)

Operates for at least 15 minutes.

3.

Verifying that upon a recirculation actuation signal, the containment sump isolation valves open.

4.

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

ST.

LUCIE - UNIT 1

3/4 5-4

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T

< 325'F av LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE high-pressure safety injection pump, and b.

An OPERABLE flow path capable of taking suction from the refueling water tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a recirculation actuation signal.

APPLICABILITY:

MODES 3~ and 4.

ACTION; a

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b, With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20, hours.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Coranission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

SURVEILLANCE RE UIREMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

• With pressurizer pressure

< 1750 psia ST. LUCIE - UNIT 1

3/4 5-7 Amendment No.28;

EMERGENCY CORE COOLING SYSTEMS REFUELING WATER TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank shall be OPERABLE with:

a.

A minimum contained volume 401,800 gallons of borated

water, b.

A minimum boron concentration of 1720 ppm, c.

A maximum water temperature of 100'F, d.

A minimum water temperature of 55'F when in MODES 1 and 2, and e.

A minimum water temperature of 40'F when in MODES 3 and 4

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30.hours.

SURVEILLANCE RE UIREMENTS 4.5.4 The RWT shall be, demonstrated OPERABLE:

a.

At least once per 7 days by:

l.

Verifying the water level in the tank, and 2.

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWT temperature.

T. LUCIE - UNIT 1

3/4 5-8 Amendment No.

3 4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3 4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions,

2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition, SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T v The most restrictive condition occurs at EOL, with Tave at no o

d operating temperature an,d is associated with a postulaated steam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 3.3X ak/k is required to control the I

reactivity transient.

Accordingly, the SHUTDOWN MARGIN required by Specification

3. l.l. 1 is based upon this limiting condition and is con-sistent with FSAR accident analysis assumptions.

For earlier periods during the fuel cycle, this value is conservative.

With Tav

< 200'F, the reactivity transients resulting from any postulated accident are minimal and a

1% ak/k shutdown margin provides adequate protection.

3/4.1.1.3 BORON DILUTION AND ADDITION A minimum flow 'rate of at least 3000 GPM provides adequate

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

A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System. volume,of 11,400 cubic feet in approximately 26 minutes.

The reactivity change rate associated with boron concentration changes will be within the capability for operator recognition and control.

Amendment No.

3 4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT MTC The limiting values assumed for the MTC used in the accident and transient analyses were

+ Oe5 x 10 4 hk/k/'F for THERMAL POWER levels

< 70K of RATED THERMAL POWER,

+ 0.2 x 10 4 hk/k/'F for THERMAL POWER Tevels

> 70K of RATED THERMAL and - 2.2 x 10 4 ak/k/'F at RATED THERMAL POWER.

Therefore, these limiting values are included in this specification.

Determination, of MTC at the specified conditions ensures that the maximum positive and/or negative values of the MTC will not exceed the limiting values.

ST.

LUCIE - UNIT 1

B 3/4, 1-1

REACTIVITY CONTROL SYSTEMS BASES 3 4.1.1:5 MINIMUM TEMPERATURE FOR CRITICALITY The MTC is expected to be slightly negative at operating conditions.

However, at the beginning of the fuel cycle, the MTC may-be, slightly positive at operating conditions and since it will become more positive at lower temperatures, this specification is provided.to restrict reactor operation when T is significantly below the normal operating temperature.

'I 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 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.

I The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of 1.0X ak/k after xenon decay and cooldown to 200'F.

The maximum boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 7,925 gallons of 8.0X boric acid solution from the boric acid tanks or 13,700 gallons of 1720 ppm borated water from the refueling water tank.

'he requirements for a minimum contained volume of 401,800 gallons of borated water in the refueling water tank ensures the capability for borating the RCS to the desired level.

The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4.

Therefore, the larger volume of borated water is specified here too.

With the RCS temperature below 200'F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restric-tions prohibiting, CORE ALTERATIONS and positive reactivity change'n the event the single injection system becomes inoperable.

ST.

LUCIE - UNIT 1 B 3/4 1-2 II Amendment No. g7, 28:

3 4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS

, The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients.

STARTUP and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for the Power Level-High, Reactor Coolant Flow-Low, and Thermal Margin/Low Pressure trips have been reduced to their specified values.

Reducing these trip setpoints ensures that the DNBR will be maintained above 1.30 during three pump operation and that during two pump operation the core void fraction will be limited to ensure parallel channel flow stability within the core and thereby prevent premature DNB.

A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure consi-erations require plant cooldown if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

I 3 4.4.2 and 3 4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

Each safety valve is designed to relieve 2 x 10 lbs per hour of saturated steam at. the valve setpoint.

The relief capacity of a single safety valve is adequate to.

relieve any over pressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating shutdown ooling loop, connected to the RCS, provides overpressure relief capa-bility and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.

The combined relief capacity of these valves is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while' perating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e.,

no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.

j. LUCIE - UNIT 1

B 3/4 4-1

REACTOR COOLANT SYSTEM BASES SAFETY VALVES Continued Demonstration of the safety valves'ift settings will occur only during shutdown and will be performed in a'ccordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel

Code, 1974 Edition.

3 4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accoranodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and power operated relief valve against water relief.

The power operated relief valve and steam bubble function to relieve RCS pressure during all design transients.

Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

3 4.4.5 STEAM GENERATORS One OPERABLE steam generator provides sufficient heat removal capa-bility to remove decay heat after a reactor shutdown.

The requirement for two steam generators capable of removing decay heat, combined with the requirements of Specifications 3.7.1.1, 3.7.1.2 and 3.7.1.3 ensures adequate decay heat removal capabilities for RCS temperatures greater than 325'F if one steam generator becomes inoperable due to single failure considerations.

Below 325'F, decay heat is removed by the shutdown cooling system.

The Surveillance Requirements for i.nspection of the steam generator tubes ensure that the structural integrity of this portion of the'RCS will be maintained.

The program for inservice inspection of steam

'enerator tubes is based on a modification of Regulatory Guide 1.83, Revision 1, Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken, ST.

LUCIE - UNIT 1

B 3/4 4-2 amendment No. 28

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 its design pressure of 1025 psig during the most severe anticipated system opera-tional transient.

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

no steam bypass to the condenser).

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

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

The total relieving capacity for all valves on all of the steam lines is 11.91 x 10~ lbs/hr which is 106.7 percent the total secondary steam flow of 11.17 x 10'bs/hr at 100/

RATED THERMAL POWER.

A minimum of 2 OPERABLE safety valves per steam generator ensures that s'ufficient relieving capaci ty is available for removing decay heat.,

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

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

For two loop.operation SP

X x (106.5) where:

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

LUCIE - UNIT 1

B 3/4 7-1

PLANT SYSTEMS BASES

$ 06.5

=

Power Level-High Trip Setpoint for two loop operation X'

'otal relieving capacity of all safety valves per steam line in lbs/hour-(5.95 x 10 1bs/hr.}

Naximum relieving capacity of any one safety valve in lbs/hour (7.44 x 10s lbs/hr.)

3 4.7.1.2 AUXILIARYFEEDNATER PUMPS The OPERABILITY of the auxiliary feedwater pumps ensures that the Reactor Coolant System can be cooled down to less than 325'F from normal operating conditions in the event of a total loss of off-site power.

Any two of the three auxiliary feedwater pumps have the required capacity to provide sufficient feedwater flow to remove reactor decay heat. and reduce the RCS temperature to 325'F where the shutdown cooling system may be placed into operation for continued cooldown.

3 4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minim'um water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less tban 325'F in the event of a total loss of off-site power.

The minimum water volume is sufficient'o

. maintain the RCS at HOT STANDBY conditions for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with'team discharge to atmosphere.

3 4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that

='the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.

The dose

.calculations for an assumed steam line rupture include the effects of a coincident 1.0 GPN primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power.

These values are consistent with the assumptions used in the accident analyses.

ST.

LUCIE - UNIT 1

B 3/4 7-2 Amendment No.

28