Text

Amend 48 to License DPR-67,changing License Condition 2.C. (1) & Tech Specs to Authorize Operation at 2700 Mwt
	ML17212B209
Person / Time
Site:	Saint Lucie
Issue date:	11/23/1981
From:	Eisenhut D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17212B210	List: ML17212B209; ML17212B212; ML17212B214;
References
	NUDOCS 8112210031
	Download: ML17212B209 (46)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C, 20555 1.

The A.

B.

C.

D.

E.

FLORIDA POWER 5 LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE PLANT UNIT NO.

1

'MENDMENT TO FACILITY OPERATING LICENSE Amendment No. 48 License Ho.

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

The applications for amendment by Florida Power 5 Light Company (the licensee) dated November 14, 1980 and'September 28, 1981 comply with the standards and requirements of the Atomic

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

8i i22i003i 8i ii23 PDR ADOCK 08000338 P

PDR

2.. Accordingly, Facility Operating License No.

DPR-67 is amended by changes to the Technical Specifications as indicated in the Attachment to this license amendment, and by amending paragraphs 2.C(1) and 2;C(2) to read as follows:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2700 megawatts (thermal), provided that the construction items, preoperational

tests, startup tests, and other items identified in Enclosure 1

to this license have been completed as specified in Enclosure l.

Enclosure 1 is an integral part of, and is hereby incorporated in this license.

(2)

Technical S ecifications The Technical Specifications contained in Appendices A andB,

-as revised through Amendment No. 48, are hereby incorporated in the license.

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

3.

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

OR THE NUCLEAR REGULATORY COMMISSION arre 1

Oivision of '4 sen ut, irector icensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Oate of Issuance:

November 23, 1981

ATTACHMENT TO LICENSE AMENDMENT NO 48 FACILITY OPERATING LICENSE NO.

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

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

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

~Pa es l-l 2-2 2-7 2-8 2-9 B 2-1 B 2-3 B 2-4 B 2-5 B 2-7 B-2-8 3/4 1-3 3/4 1-10 3/4 1-18 3/4 1-30 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-8 3/4 2-9 3/4 2-14 3/4 2-15 B 3/4 1-1 B 3/4 1-2 B 3/4 2-2 B 3/4 4-1 B 3/4 7-1 B 3/4 7-2

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

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

RATED THERMAL POWER 1.3 RATED THERMAL PO'WER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWt.

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

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

OPERABLE -'PERABILITY 1.6 A system, subsystem, train, componen't or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation,

controls, normal and emergency electrical power

sources, cooling or seal. water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or'device to perform its function(s ) are also capable of performing their related support function(s).

REPORTABLE OCCURRENCE 1.7 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specificat'ions 6.9.1.8 and 6.9.1.9.

ST.

LUCIE - UNIT 1, Amendment No. g3, $) 4 8

.DEFINITIONS.

CONTAINMENT VESSEL' INTEGRITY 1.8 CONTAINMENT VESSEL INTEGRITY shall exist when:

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

a.

Capable of being closed by an OPERABLE containment'utomatic isolation 'valve system, or b.

Closed by manual valves, blind flan'ges, or deactivated automatic valves secured in their closed position except as provided in Table 3.6-2 of Specification 3.6.3.1, 1.8.2 All containment vessel equipment hatches are closed and

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

CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the. channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBRATION may be performed by any series of seauen-tial, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

ST.

LUCIE - UNIT 1

1-2

I 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer

pressure, and maxi-mum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1.

APPLICABILITY:

MODES 1 and 2.

ACTION:

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

REACTOR COOLANT SYSTEM PRESSURE

'2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY:

MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1

and 2

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

MODES 3, 4 and 5

Whenever the Reactor Coolant System pressure has exceeded 2750 ps'ia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

ST.

LUCIE - UNIT 1

2-1

600 580 560 0

I f S00

~ 520 '

~

500 480 UNACCEPTABLE OPERATION REACTOR OPERATION LIMITED TO LESS TIIAN 5gP'F BY ACTUATION OF TIIE HAIN STEAM LINE SAFETY VALVES.

VESSEL FLOW LESS MEASUREMENT UNCERTAINTIES 370,000 GPM FOR PRE-CLAD COLLAPSE OPERATION ONLY LIMITS CONTAIN NQ ALLOWANCE FOR INSTRUMENT ERROR OR FLUCTUATIONS VAI.ID FOR AXIAL SIIAPES AND INTEGRATED ROD RADIAL PEAKING FACTORS LESS TIIAN OR EQUAL TO THOSE ON FIGURE B 2.1-1 ACCEPTABLE OPERATION Ig

.l~

Ig g

1iP 5

~ 5o l)5 O

O OOO UNACCEPTABLE OPERATION 460 0

0.20 0.40 0.60 0.80 1.00

'1.20 1;40 1.60

'.1.80 2.00 FRACTION OF RATED TIIEmL POWER Figure 2.1-1 REACTOR CORE TIIERHAL MARGIN SAPETY LIMIT POUR REACTOR '.

COOLING PUMPS OPERATINO

1.4 1.2 UNACCEPTABLE OPERATION (0.0, 1.17),

UNACCEPTABLE OPERATION 1.0

(-.145, 1.0)

(0.2, 1.0) 0.8 QRZ

(-0.4, 0 70)

(0.4,0 70) 0.6 ACCEPTABLE OPERATION 0.4 0.2 0.0

-0.5

-0.4

'0.2 0.0 0.2 AXIAL SHAPE INDEX, Yl 0.4 0.6 FIGURE 2.2-2 Local Power Density-High Trip Setpotnt Part 2(QR Versus Yl)

ST.

LUCIE UNIT 1 2-7 Amendment No.'7, gg, 4 8

Al FUNCTION 1.4 POLAR 2061 'l

~

QR1 + 15.85 TIN 8950 1.3 1.2 1.0-0.6>>0.5

-0.4

-0.3

-0.2

-0.1 0.0 0.1 0.2 0.3 0.4 0.5 0,6 AXIAL SHAPE INDEX, Yl FIGURE 2.2-3 Thermal Margin/Low Pressure Trip Setpoint ST.

LUCIE - UNIT 1 2-8 Amendment.'(p,

)7 4 8

PVAR 2061

~ Al QRl + 15,85 TIN 8950 QRl FUNCTION 1.2 1.0

(. 972,. 972) 0.8

(.781,

.863)

QR1 0.6 0.4 (0,.235}

0.2 0

0.2 0.4 0.6 0.8 1.

0'RACTION OF RATED THERMAL POWER

. 1.2 FIGURE 2.2-4 Thermal Margin/Low Pressure Trip Setpoint Part 2 (7raction of BATED THER'04~R Versus QR1) l ST.

LUCIE - UNIT 1 2-9 Amendment No. g7, 48

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding 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 coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have

'een related to DNB through the CE-1 correlation.

The CE-1 DNB correlat'ion has been developed to predict the DNB 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 operational transients, and anticipated transients is limited to 1.23.

This value corresponds to a

95 percent probability at a

95 percent confidence level tha-DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curv'es of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the minimum DNBR is no less than 1.23 for the family of axial shapes and corresponding radial peaks shown in Figure B 2.1-1.

The limits in Figure 2.1-1 were 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 112% of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in Table 2.1-1.

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

ST.

LUGI E - UNIT 1

B 2-1 Amendment No. gj, 48

2.0 1.6 z0I-D 1.2 o

V O

K W

O '0.0 X

0.4 ROD RAOIAL P 1.44 1.44 1.51

. 1.56 1.56 EAK 0

0 10 20 30 40 50 60 70 80 90 100 PERCENT OF ACTIVE CORE LENGTH FROM BOTTOM Figure B2.1-1 Axial Power Distribution for Thermal Margin Safety Limits

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

The reactor protective system in combination with the Limiting Condi-tions.-for Operation, is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature,

pressure, and thermal power level that would result in a DNBR of less than 1.23 and preclude the existence of flow instabilities.

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

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant componen'ts which permits a

maximum transient pressure of 110%

(2750 psia) of design pressure.

The Reactor Coolant System piping, valves and fittings, are designed to, ANSI B 31.7, Class I which permits a maximum transient pressure of 110/ (2750 psia) of component design pressure.

The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

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

ST.

LUCIE - UNIT 1

B 2-3 Amendment No.48

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter.

The Trip Values have been selected to ensure that the reactor co're and reactor coolant system are prevented from exceeding their safety limits.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that 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.

Power Level-Hi h

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

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

Operator action is required to increase the trip setpoint as THERMAL POWER is increased.

The trip setpoint is automatically decreased as THERMAL POWER decreas'es.

The trip setpoint has a maximum value of 107.0/ of RATED THERMAL POWER and a minimum 'setpoint of 15Ã of RATED THERMAL POWER.

Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual THERMAL POWER level at which a trip would be actuated is 112% of RATED THERMAL POWER. which is consistent with the value used in the safety,.

analysis.

Reactor Coolant, Flow-Low The Reactor Coolant Flow-Low trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant flow.

Provisions have been made in the reactor protective system to permit operation of the reactor at reduced power if one or two ST.

LUCIE - UNIT 1

B 2-4 Amendment No.

$7, g 8

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow-Low (Continued) reactor coolant pumps are taken out of service.

The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in ccnsideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.23 under normal operation and expected transients.

For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip setpoints, the Power Level-High trip setpoints, and the Thermal Margin/Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually iset to the desired two-or three-pump position.

Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below l. 23'uring normal operational transients and ariticipated transients when only two or three reactor coolant pumps are operating.

Pressurizer Pressure-Hi h

The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without,.reactor trip.

This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.

Containment Pressure-Hi h

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

team Generator Pressure-Low The'team Generator Pressure-Low trip provides protection against an xcessive rate of heat extraction from the steam generators and sub-equent cooldown.of the reactor.cool.ant.

The setting of. 600 psia is ufficiently below the full-load operating point of 800 psig so as not ST.

LUCIE - UNIT 1

B 2-5 Amendment No,

LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Pressure-Low (Continued)

I to interfere with normal operation, but still-high enough to provide the required protection in the event of excessively high steam flow.

This setting was used with an uncertainty factor of + 22 psi in the accident analyses.

Steam Generator Water Level - Low The Steam Generator Water Level-Low trip provides core protection

'y preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded due to loss of steam generator heat sink.

The specified setpoint provides allowance that there will be sufficient 'water inventory in the steam generators at the time of trip to provide a margin of more than 10 minutes before auxiliary feedwater is required.

Local Power Densit -Hi h The local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a consequence of axial power maldistributions.

A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2.--The AXIAL SHAPE INDEX is calculated from the upper and lower ex-core neutron detector channels.

The calculated setpoints are generated's a function of THERMAL POWER level with the allowed CEA group position being inferred from the THERMAL POWER level.

he trip is automatically bypassed below 15 percent power.

The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment per-itted for continuous operation are assumed in generation of the set-oints.

In addition, CEA group sequencing in accordance with the pecifications 3.1.3.5 and 3. 1.3.6 is assumed.

- Finally, the maximum insertion of CEA banks which can 'occ'fir during. any. anticipated operational ccurrence prior to a Power Level-High trip is assumed.

ST.

LUCIE - UNIT 1

B 2-6 Amendment No. 27

LIMITING.SAFETY SYSTEM SETTINGS BASES Thermal Mar in/Low Pressure The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than 1.23.

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

The computed value is a function of the higher of aT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX.

The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function.

In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.

Finally, the maximum insertion of CEA banks which can occur duriqg any anticipated operational occurrence prior to a

Power Level-High trip is assumed.

The Thermal Margin/Low Pressure trip setpoints include appropriate allowances for equipment response time, calculational and measurement uncer-

tainties, and processing error.

A further allowance of 30 psia is incladed to compensate for the time delay associated with providing effective termina-tion of the occurrence that exhibits the most rapid decrease in margin to the DNBR limit.

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

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

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

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

ST.

LUCIE - UNIT 1

B 2-7 Amendment No. g7, ii8. 4 8

LIMITING SAFETY SYSTEM SETTINGS BASES Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 155 of RATED THERMAL POWER.'his trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the

'ervice life.of these valves.

No credit was taken in the accident analyses for operation of this trip.

Its functional capability at.the specified trip setting is required to enhance the overall reliability of the Reactor Protec-tion System.

Rate of Chan e of Power-Hi h

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

Its trip setpoint does not correspond to a Safety Limit and no 'credit was taken in the accident analyses for operation of this trip.

Its functional capability at the specified trip setting 'is required to enhance the overall reliability of the Reactor Protection System.

ST.

LUCIE - UNIT 1

8 2-8 Amendment No. 4 8

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T < 200'F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be:

> 2.0X hk/k, and in addition with the Reactor Coolant System drained below the hot leg centerline, one charging pump shall be rendered inoperable.*

APPLICABILITY:

MODE 5.

ACTION:

If the SHUTDOWN MARGIN requirements cannot be met, immediately initiate and continue boration at

> 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN requirements of Specification 3.1.1.2 shall be determined:

a 0

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

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

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

l.

2.

3.

4.

5.

6.

Reactor coolant system boron concentration, CEA position, Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.

c.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , when the Reactor Coolant System is drained below the hot leg centerline, by consideration of the

.factors in 4.1;1.2.b and by verifying at least one charging pump is rendered inoperable.*

Breaker racked-out.

ST.

LUCIE - UNIT 1

3/4 1-3 Amendment No. < 8

REACTIVITY'ONTROL SYSTEMS BORON DILUTION LIMI TING: CONDITION:FOR: OPERATION, 3.1.1.3 The flow rate of reactor coolant to the reactor pressure vessel shall be

> 3000 gpm whenever a reduction in Reactor Coolant System boron concentat7on is being made.

APPLICABILITY: ALL MODES.

ACTION:

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

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

SURVEILLANCE'.RE UIREMENTS 4.1.1.3 The flow rate of reactor coolant to the reactor pressure vessel shall be determined to be

> 3000 gpm within one hour prior to the start of and at least once per hour during.a reduction in the Reactor Coolant System boron concentration by either:

a.

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

Verifying that at least one low pressure safety injection pump is in operation and supplying 3QOO gpm to the reactor pressure vessel.

ST.

LUCIE - UNIT 1

3/4 1-4

REACTIVITY..CONTROL 'YSTEMS SURVEILLANCE REQUIREMENTS b.

At least once per 31 days by verifying that each valve (manual,.

power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

~

ST.

LUCIE - UNIT 1

3/4 '1-9

REACTIVITY CONTROL

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

a.

Two flow paths from the boric.acid makeup tanks via either a

boric acid pump or a gravity feed connection, and a charging pump to the Reactor Coolant System, and b.

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

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With only one of the above required boron injection flow paths t'o the Reactor Coolant System OPERABLE, restore at least two boron i.njection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or make the reactor subcritical within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and borate to a

SHUTDOWN MARGIN equivalent to at least 2A hk/k at 200'F; restore at least two flow paths to OPERABLE statuswithin the next 7

days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

a

~

At least once per 7 days by:

1.

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

ST.

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

n ID

~ gg CO n2 C

n 3 o'~o 0

Nn+

UJ x o

mC O 'D o-x 0(

o o O C CD R

CD D) o o-2'a PtC CO 0

37lll U

tD0 O

A U

O0zOlllz 0z C)

C)

CD CI I

~ ~ I t

I I.'; I

)II/

~ Jl TEMPERATURE ( F) o ID MINIMUMBORIC ACIDMAKEUPTANKVOLUME,(GALLONS)

'4 O

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 ~

b.

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,'.

A maximum solution temperature of 100'F.

4.

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

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

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

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

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

a.

At least once per '.days by; l.

Verifying the boron concentration in each water source, ST.

LUCIE - UNIT 1

3/4 1-18 Amendment No.

Pg, g g

REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS (Continued)

LIMITING CONDITION FOR OPERATION C.

With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Power Dependent Inser-tion Limits for intervals

> 5 END per 30 EFPD interval or

> 14 EFPD oer calendar year, except during operations pursuant to the provisions of ACTION items c.

and d. of Specification

3. 1.3.1, either:

1.

Restore the regulating groups to within the Long Term Steady State Insertion Limits within two hours, or 2.

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

SURVEILLANCE REOUIREMENTS 4.1.3.6 The.position of each regulating CEA group shall be determined to be within the Power Dependent Insertion Limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the PDIL Auctioneer Alarm Circuit is inoperable",

then verify the individual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The accumulated times during which the regu-lating CEA groups are inserted between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ST.

LUCIE - UNIT 1

3/4 1-29

1. 00 0.90 0.80

'0.70 g

0.60

> 0.50 ws Oa: 0.40 0 r H

". 0.30 0.20

~82,

.70)

(68,.75) 1 l

I

).LONG TERM

) STEADY STATE INSERTION LIMIT SHORT TERM STEADY STATE INSERTION LIMI POWER DEPENDENT INSERTION LIMIT 0'. 10 0

GROUPS 0

27 55 82 109 137 3

0 '.

27 55 82 109 137 0

27 55 82 109 137'EA INSERTION (INCllES) t Figure 3.1-2 CEA Insertion Limits vs TllERMAL POWER with 4 Reactor Coolant Pumps Operating

16.0 UNACCEPTABLE OPERATION o

CO C4

+

6

+

o

15. 0

14. 0 ACCEPTABLE OPERATION 15.0-13.0 BOL EOL CYCLE LIFE FIGURE 3.2-1 Allowable Peak Linear Heat Rate vs Burnup ST.

LUCIE - UNIT 1 3/4 2-3 Amendment No. lj, gg, g8

1.2

~ 1.0 O

~

~ 0.9 O

~ 0.8

~ 0.7 o

O 5 0.6 REGION OF UNACCEPTABLE OPERATION

(-0. 3,0.58)

REGION OF

, ACCEPTABLE OPERATION

(-0.05, 0.82)

'(0.15, 0.82)

REGION OF

~ UNACCEPTABLE OPERATION (0.3, 0.58) 0.5 0.4

-0.6

-0.4

-0.2 0.0 0;2 0.4 0.6 PERIPHERAL AXIAL SHAPE INDEX FIGURE 3.2-2 AXIAL SHAPE INDEX vs. Fractioa of Maximum Allowable Power Level Per Specification 4.2.1.3 ST. LUCIE UNIT 1 3/4 2-4 Amendment No. gj, gg, 4 g

1. 10

1. 08 g 1.06 O

$ 1.04 134.7,1.071 118.6,1.067 o 106.5,1.063

~

4 90.5, 1.057

~ 74.4, 1.050 f

62.3, 1.045 46.2, 1.035

l. 02 418.1, 1.017 O

V

~0 1.00

~ 2.0, 1.004 0

14 28 42 56'0 84 98 DISTANCE FROM BOTTOM OF CORE, INCHES FIGURE 4.2-1 112 125 140 lb OO AUGMENTATION FACTORS vs DISTANCE FROM BOTTOM OF CORE

POWER DISTRIBUTION LIMITS TOTAL PLANAR RADIAL PEAKING FACTOR - F LIMITING CONDITION FOR OPERATION 3.2;2 The calculated value of Fx, defined as Fx

= Fx (1+T '), shall be T

T limited to 1.70.

APPLICABILITY'ODE 1*.

ACTION:

With F 1.70, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

xy a.

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 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 xy yy xy q

in non-LOAD FOLLOW OPERATION and by the expression F

= 1.03 F

(1+T

)

T when in LOAD FOLLOW OPERATION.

Fx shall be determined to be within its limit at the following intervals:

a.

Prior to operation above 70 percent of RATED THERJSL POWER after each 'fuel loading, b.

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

Within four hours if the AZIMUTHAL POWER TILT (T } is

> 0.03.

See Special Test Exception 3.10.2.

ST.

LUCIE - UNIT 1

3/4 2-6

'Amendment No.. g7, Pg,4 8

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued 4.2.2.3 F

shall be determined each time a calculation of F

. is xy 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.

This determination shall be limited to core planes between'54 and 85K 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 T

required and the value of T used to determine F

shall be the measured T

value of T ST.

LUCIE - UNIT 1

3/4 2-7 Amendment No.

1.1 1.0

l. 7) 1.0)

UNACCEPTABLE OPERATION REGION (1.78,0.9)

O g

0.8 CJ 0.7 ACCEPTABLE OPERATION REGION 0.6 1..70

l. 71 1.72

l. 73 1.74 1.75 Heasured F>, Fl T

T 1.76 1.77

1. 78 FIGURE 3.2-.3 Allowable Combinations Of Thermal Power And F<,

Fxy

POWER DISTRIBUTION LIMITS TOTAL INTEGRATED RADIAL PEAKING FACTOR -

F LIMITING CONDITION FOR OPERATION 3.2.3 The calculated value of Fr, defined as Fr =

F (1+T ), shall be T

T limited to ( 1.70.

r' r

q'PPLICABILITY:

MODE 1*.

CTION:

With Fr > 1.70, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

T a.

Be in at least HOT STANDBY, or b.

Reduce THERMAL POWER to bring the combination of THERMAL POWER and F

to witin the limits of Figure 3.2-3 and withdraw the full length 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 allowabl.e fraction of RATED THERMAL POWER determined by Figure 3.2-3) and subsequent opera-tion shall be maintained within the reduced acceptable operation

'region of Figure 3.2-4.

URVEILLANCE REQUIREMENTS

.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F

shall be calculated by the expres'sionTF

=

F (1+T

) when in T

non-LOAD F5LLOW OPERATION and by )he expression F

1.05 F

(7 + T

)

hen in LOAD FOLLOW OPERATION.

F shall be deterKined to be within its limit at the following intervals. r a.

Prior to operation above 70 percent of RATED THERMAL POWER

'fter each fuel loading, b.

At least once per 31 days of accumulated operation in MODE 1, and Within four hours if the AXIMUTHAL POWER TILT (T ) is

> 0.03.

See Special Test Exception 3. 10.2.

ST.

LUCIE - UNIT 1

3/4 2-9 Amendment No. l7 P ',

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued 4.2.3.3 F

shall be determined each time a calculation of F is required.

r 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 4.2.3.4 T

shall be determined each time a calculation of F is required r

and the value of T used to determine F shall be the measured value of T

.q'T.

LUGIE - UNIT 1

3/4 2-10 Amendment No. 27

POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3'.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a.

Cold Leg Temperature b.

Pressurizer Pressure c.

Reactor Coolant System Total Flow Rate d.

AXIAL SHAPE INDEX APPLICABILITY:

MODE 1.

ACTION:

W With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to (

5'A of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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

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

ST.

LUGIE - UNIT 1

3/4 2-13 Amendment No.

27

TABLE 3.2-1 DNB MARGIN LIMITS Parameter Cold Leg Temperature Pr essurizer Pressure Reactor Coolant Flow Rate AXIAL SHAPE INDEX Four Reactor

, Coolant Pumps

~0erati n

< 549'F

. > 2225 psia*

> 370,000 gpm Figure 3.2-4 Limit not applicable during either a

THERMAL POWER ramp increase in excess of 5%%d of RATED THERMAL POWER or a THERMAL POWER step increase of greater than 10Ã of RATED THERMAL POWER.

wa ST; LUCIE - UNIT 1

3/4 2-14 Amendment No. 7,4 p

~

1.1 1,0 0.9 O

I 0.8 p

0.7 O

OM 0.6 UNACCEPTABLE OPERATION REGION

(-0 3, 0.70)

(-0.08,1.00)

ACCEPTABLE OPERATION REGION

(.15 (0.3, 0.70) 1.00)

UNACCEPTABLE OPERATION REGION 0.5 0.4

-0. 4

-0.4

-0.2 0.0 0.2 0.4 0.6 PERIPHERAL AXIAL SHAPE L%)EX (Yl)

FIGURE 3.2-4 AXIAL SHAPE INDEX Operating Limits With 4 Reactor Coolant Puinps Operating ST.

LUCIE - UNIT 1 3/4 2>>15 amendment No.

R7s g2 4 8

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

'ubcritical 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 Tavg The most restrictive condition occurs at EOL, with Tave at no load operating temperature, and is associated'with a postulated steam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 5.0g b,k/k is required.to control the 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 Tavq ( 200'F, the reactivity transi ent resulting from a boron dilution event with a partially drained Reactor Coolant System requir es a

2~ b,k/k SHUTDOWN MARGIN and restrictions on charging pump operation to provide adequate protection.

A 2X ak/k SHUTDOWN MARGIN is 1.0X nk/k conservative for Mode 5 operation. with total RCS volume present, however LCO 3.1,1.2 is written co~servatively for simplicity.

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'ill be gradual during boron concentration changes in the Reactor Coolant System.

A flow rate of at least

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

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

wl

'3 4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT MTC Amendment No. g7, PPJ.:i" B 3/4 l-l The l.imiting values assumed for the MTC used in the accident and transient analyses were

+ 0.5 x 10 4 ak/k/'F for THERMAL POWER levels

( 70Ã of RATED THERMAL POWER,

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

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

Therefore,'hese 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

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.

l'/4.1.2 BORATION SYSTEMS

'I 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) borat'ed 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'ystems 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 all operating conditions of 2.0X ak/k after xenon decay and cooldown to 200'F.

The maximum boration capability requirement occurs at EOL from full, power equi librium xenon conditions and requires 7,925 gallons of S.OX boric acid solution from the boric acid tanks or 13,700 g'allons of 1720 ppm borated water from the refueling water tank.

The 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.-

wj With the RCS temperature below 200'F, one injection system is acceptable wi.thout single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restric-tions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

ST.

LUCIE - UNIT 1

B 3/4 1-2 Amendment No.

g7, g8, 8.".:

3/4. 2 POWER DISTRIBUTION LIMITS BASES 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, provide adequate monitor ing 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 an'd 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 Specification 3.2.4 are satisfied, and 4) the TOTAL PLANAR RADIAL PEAKING FACTOR does'ot 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 1.07,* 3) an engineering uncertainty factor of 1.03, 4) an allowance of 1.01 for axial fuel 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 PEAKENG FACTORS -

F AND F

'AND AZIMUTHAL POWER TILT - T The limitations on Fx and T are provided to ensure that the assump-T tions used in the analysiPyfor esfablishing the Linear Heat Rate and Local Power Density - High LCOs'and'LSSS setpoints remain valid dur'ing operation at theTvarious allowable CEA group insertion limits.

The limitations on F and T

are provided to ensure that the assumptions An uncertainty factor of l. 10 applies when in LOAD FOLLOW OPERATION.

ST.

LUCIE - UNIT 1

B 3/4 2-1

'Amendment No. g7, 32

POWER DISTRIBUTION LIMITS BASES used in the analysis establishing the DNB Margin LCO, and Thermal Margin/Low Pressure LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. If F F

or T exceed xy' q

their basic limitations, operation may continue under the additional

'estrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the 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, sub-sequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The value of T that must be used in the equation F

=

F

.(1'+ T

)

q xy xy.

q and F

=

F (1+T

) is the measured tilt..

r r

q The surveillance requirements for verifying'hat F

F and T are.

T T

within their limits provide assurance that the actual values of Fx, F

T T.

and T

do not exceed the assumed values.

Verifying F and F

after T

T q

each fuel 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 tPe transient and accident analyses.

The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum bNBR of 1.2B throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is'ufficient 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 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

ST.

LUCIE - UNIT 1

B 3/4 2-2 Amendment No. gP, g 8

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 1.23 during all normal operations and anticipated transients.

STARTUP and POWER OPERATION may be initiated and may proceed with one or two'eactor 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 1.23 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'hereby prevent premature DNB.

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

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer

'code safety valves operate to prevent the RCS from being pressuri zed above its Safety Limit of 2750 psia.

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

The.relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected-to the RCS, provides overpressure relief capa-bilityy 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

'imit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED 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'rip on the loss of turbine) and also assuming no operation of the pressur izer power operated relief valve or steam dump valves.

ST.

LUCIE - UNIT 1

B 3/4 4-1 Amendment No.48

REACTOR COOLANT SYSTEM BASES SAFETY YALVES Continued Demonstration of the, safety valves'ift settings will o'ccur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel

Code, 1974 Edition.

3 4.4.4 PRESSURIZER A steam'ubble in the pressurizer ensures that. the RCS is not a

hydraulically solid system and is capable of accommodating pressure surges during op'eration.

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.

The required pressurizer.

heater capacity if capable of mainhining natural circulation subcooling.

Operability of the heaters, which are powered by. a diesel generator

bus, ensures ability to maintain pressure control even with loss of offsite power.

3 4.4.5 STEAN 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'.7.1.1, 3.7.1.2 and 3,7.1.3 ensures adequate decay heat removal capabi'lities 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 cooli.ng system'.

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

The program for i.nservice inspection of steam generator tubes is based oo a modification of Regulatory Guide 1.83, Revision 1, Inservice inspection o'f 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 i nspection 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. R.

37

~

J

~

I ~

~

I

~

J

~

t

~

I

~

I

~

~ I 0

~

I

~ i

~

I ~

~

J J

~

J

~

I I

~

J

~

J

~

4

~

+

~

~ ~

~

I

~

I ~

~

0

~

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~

0

~

~ ~

~

~ I

~

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~

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PLANT SYSTEMS BASES 106. 5 Power Level-High Trip Setpoint for two loop operation Total relieving capacity of all safety valves per steam line in lbs/hour (6.192 x 10< lbs/hr.)

Maximum relieving capacity of any one safety valve in lbs/hour (7.74 x 10s lbs/hr.)

3 4.7.1.2 AUXILIARY FEEDWATER PUMPS The OPERABILITY of'he 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 operotion for continued cooldown.

3/4.7.1. 3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank vij.th the min'imum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 325'F in the event of a total.

loss of off-site power.

The minimum water volume is sufficient.to maintain the RCS at HOT STANDBY conditions for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with steam discharge to atmosphere..

3/4.7.1.4 ACTIYITY The limitations on secondary system specific activity.ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a.steam line rupture.

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

These values are consistent with the assumptions used

$ n the accident analyses.

ST.

LUCIE - UNIT 1

B 3/4 7-2 endment No.. gA, 48

ML17212B209

Text

Attachment:

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