ML17353A503

Proposed Tech Specs,Increasing Allowed Rated Thermal Power from 2,200 Mwt to 2,300 Mwt

ML17353A503
Person / Time
Site:	Turkey Point
Issue date:	12/18/1995
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17353A502	List: L-95-245, Application for Amends to Licenses DPR-31 & DPR-41,revising Definition of RTP from 2,200 Mwt to 2,300 Mwt.Proprietary TR WCAP-13719,Rev 2 & Nonproprietary TRs WCAP-13718,Rev 2 & WCAP-14276,Rev 1 Encl.Proprietary TR WCAP-13719 Withheld ML17353A503 ML17353A504 ML17353A505 ML17353A506
References
NUDOCS 9512270309
Download: ML17353A503 (83)

Text

ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS Marked Up Techn9.ca1 Specification Pages:

L9.cense Cond3.ti.ons 1-5 2-2 2-4 2-5 2-7 2-8 2-9 2-10 3/4 2-4 3/4 2-11 3/4 2-16 3/4 3-23 3/4 3-26 3/4 3-27 3/4 4-7 3/4 4-8 3/4 4-31 3/4 4-32 3/4 4-33 3/4 5-5 3/4 6-14 3/4 7-2 3/4 7-6 3/4 7-7 3/4 7-11 3/4 7-17 6-20 6-20a B 2-1 B 2-7 B 3/4 2-1 B 3/4 2-4 B 3/4 2-8 B 3/4 4-2 B 3/4 4-8 B 3/4 4-9 B 3/4 6-3 B 3/4 7-2 B 3/4 7-3 B 3/4 7-4 CORE OPERATING LIMITS REPORT 9512270309 951218

)

PDR ADOCK 05000250

)

P PDR

U dated throu h Amendment 175 Dated 8 8 95 DPR-31 Page 2

3.

A.

C.

D.

Pursuant to Section 104b of the Atomic Energy Act of

1954, as amended (the Act), and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess,

use, and operate the facility as a utilization facility at the designated location on the Turkey Point site; Pursuant to the Act and 10 CFR Part 70, to receive,

possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis

Report, as supplemented and amended; Pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive,

possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor

startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration,

.and as fission detectors in amounts as required; Pursuant to the Act and 10 CFR Part 30 to receive,

possess, and use at any time 100 millicuries each of any byproduct material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; Pursuant to the Act and 10 CFR Part 40 and 70 to receive,

possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical

form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; Pursuant to the Act and 10 CFR Parts '30 and 70, to possess, but not

separate, such byproduct and special nuclear materials as may be produced by the-operation of Turkey Point Units Nos.

3 and 4.

This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations; 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the

rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified below:

A.

Maximum Power Level lh pit tt th 1 dt t

th 1 tittyt~

power levels not in excess of megawatts (thermal).

g 3oo

0

~

~C'e$ ~ p)g9 C

~P

U dated throu h Amendment 169 dated 8 8 95 DPR-41 Page 3

C.

Pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive,

possess, and use at any time any, byproduct,,source and special nuclear material as sealed sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required:

D.

Pursuant to the Act and 10 CFR Part 30 to receive,

possess, and use at any time 100 millicuries each of any byproduct material without restriction to chemical or physical form, for sample analysis or instrument calibration; E.

Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; F.

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units No.

3 and No. 4.

This license shall be deemed to contain and is subject to the conditions specified in the following Commission Regulations in 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50 and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the

rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A.

Haximum Power Level The reactor shall not be made critical until the tests described in the applicant's letter of April 3,

1973, have been satisfactorily completed.

Thereafter, the applicant is authorized to oper te the facility at reactor core power levels not in excess of megawatts thermal.

B.

Technical S ecifications g Paw The Technical Specifications contained in Appendix A, as revised through Amendment No.

169 are hereby incorporated in the license.

The Environmental Protection Plan contained in Appendix B is hereby incorporated into the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

0

<za C. ~

NT NS UA RANT POWER T

T RATIO 1.23 QUADRANT POWER TILT RATIO shall be the. ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

AT T

RMA P

W 1.24 RATED THERMAL POW R shall be a total reactor core heat transfer rate to the reactor coolant of gÃ0't.

~RPORT48 E

VillT t ~13 1.25 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

SHUTDOWN MARG 1.26 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.27 The SITE BOUNDARY shall mean that line beyond which the land or property is not owned,

leased, or otherwise controlled by the licensee.

SOLIDIFICAT ON 1.28 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK 1.29 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGER D T ST BASIS 1.30 A STAGGERED TEST BASIS shall consist of:

a.

b.

A test schedule for n systems, subsystems,

trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

TURKEY POINT - UNITS 3

& 4 1-5 AMENDMENT NOS.157AND 15

665 655 4$ PSlA I

2250 PS UNACCEPTABLE OPERATlON 1825 PSIA 615 575 0.0 Ll 02 03 OA DS 0.6 0.7 O.b 0.9 1.0 1.1

.2 5u B&717U7C Y'I-lE

&,-YwAt HC9 K< G,re<

l aR, 7H1$

KlQ,MA.Q POWER (FRACTlON OF NOMlNAL)

FIGURE 2. 1-1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION TURKEY POINT - UNITS 3 4 4 2-2 AMENDMENT NOS.>>7 ANO>>2

670 660 I

I I

I I'455 PS I

I i

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1 I

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I UNACCEPTABLE OPERATlQN 650 640

~ 630 i-620 610 600 It I

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'400 PS I

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2000 I

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1.1 1.2 POWER {FRACTlONOF NOMINAL)

Figure 2.1-1 Reactor Core Safety Limit-Three Loops in Operation

QEAC 0 T

S S

U 0

S 0

S UNC 0 AL U

1.

Manual Reactor Trip ALLOWA8LE

~AU N.A N.A.

S 0

2.

Power Range, Neutron Flux a.

High Setpoint b.

Low Setpoint 3.

Intermediate

Range, Neutron Flux

<112.0X of RTP**

<28.0X of RTP'*

<31.0X of RTP'.*

<109X of RTP**

<25X of RTP**

<25X of RTP**

4.

Source

Range, Neutron Flux

<1.4 X 10 cps

<10 cps m

C) m 5.

Overtemperature hT 6.

Overpower hT 7.

Pressurizer Pressure-Low S.

Pressurizer Pressure-High 9.

Pressurizer Water Level-High 10.

Reactor Cool ant Flow-Low See Note 2

See Note 4

21817 psig

<2403 psig

<92.2X of instrument span

88. 8 of loop des gn flow*

See Note 1

See Note 3

>1835 psig

<2385 psig

<92X of instrument span

>90X of loop design flow*

+

(o%

Cl

(/l ll.

Steam Generator Water Level Low-Low

• Loop design flow 49-5ee pm

'* RTP Rated Thermal Power

~%. of narrow range instrument span 8 /5('f narrow range instrument span Revised Thermal Design Procedure (RTDP) Technical Spec9.fication change under review by the NRC (L-95-131 and L-95-250)

TABLE 2.2-ontinued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 12.

Steam/Feedwater Flow Mismatch Coincident With Steam Generator Water Level-Low ALLOWABLE VALUE Feed Flow <23.9%

below rated Steam Flow of narrow range instrument span TRIP SETPOINT Feed Flow <20%

below rated Steam Flow of narrow range instrument span 13.

Undervoltage - 4.16 kV Busses A and B

14.

Underfrequency - Trip of Reactor Coolant Pump Breaker(s)

Open 15.

Turbine Trip a.

Auto Stop Oil Pressure b.

Turbine Stop Valve Closure 16.

Safety Injection Input from ESF 17.

Reactor Trip System Interlocks a.

Intermediate Range Neutron Flux, P-6

>69% bus voltage

>55.9 Hz

>42 psig Fully Closed***

N. A.

>6.0x10 " amps

>70% bus voltage

>56.1 Hz

>45 psig Fully Closed***

N.A.

Nominal lxlO" amp

• • • Limit switch is set when Turbine Stop Valves are fully closed.

Revised Thermal Design Procedure (RTDP) Technical Specification change under review by the NRC (L-95-131 and L-95-250)

TABLE 2.2-1 Continued NOTE 1:

OVERTENPERATURE hT hT 1 + T~S 1 + r>S Nere:

hT 1 + TIES 1 + TzS 1+ v3S hTD K2 4-1+v S 1 + t5S

~4

~S

(

1

) 5 ETO (Kg K2 ~1 +

S) [T (

y

~

T J + K (p p,)

f pg))

1 + t'3S 1 + ~sS~

1 + ~,S Measured hT by RTD Instr mentation Lead/Lag coiipensator on measured hT; r> Os, t'2 Os Lag coepensator on aeasured hT; t3 Os Indicated hT at RATED THERllhL PNER I. 2+

0 0-.0%07/

F; C

~ >. < l 7 The function generated by the lead-lag coepensator for T dynaaic coapensation; Tiae constants utilized in the lead-lag caapensator for T

, r4 ~ 25s, T5 < 3 S~

aVg>

4 hverage teaperature, F;

1 + T6S K3 Lag coepensator on aeasured Tag., t6 Os 577 574-.z. F (Hoeinal T at RATED THERNLL POMER);

psig;

~.eo

)

Pressurizer

pressure, psig; Revised Thenna1 Design Procedure (RTDP) Technica1 Specification change under review by the NRC (L-95-131 and L-95-250)

TABLE 2.2-1 Continued TABLE NOTATIONS Continued NOTE 1:

(Continued) pt 5

2235 psig (Noainal RCS operating pressure);

Laplace transfora operator, s-NOTE 2:

and f~ (hI) is a function of the indicated difference between top and bottoe detectors of the peer range neutron ion chaabers; with gains to be selected based on Neasured instr~nt response during plant startup tests such that:

+>

(1)

For qt " q between -

and fq (hI) = 0, where qt and qb are percent RATED THERNL PNER in the top and bottoa halves of the core respectively, and q

+ qb is total THERNL PNER in percent of RATED THERRAL PNER; w/

(2)

For each percent that the nagnitode of qt - nh exceeds~ the 4T Trip Setpoint shall be autoaatically reduced by of its value at RATED THERNL PNER; and Q.O cs 4o (3)

For each percent that the aagnituck o

qt - q exceeds

+

the hT Trip Setpoint sha11 be autoaatically reduced by of its value at RATED THERNL ONER.

/.

The channels aaxiam trip setpoint shall not exceed its computed setpoint by aore than of instr~nt span.

o. <+%

Revised Thermal Design Procedure (RTDP) Technical Specification change under review by the NRC (L-95-131 and L-95-250)

TABLE 2.2-1 Continued TABLE NOTATIONS Continued NOTE 3:

OVERPNER AT hT 1+x S

1

<hT fK "K

'T" K

[T

) "T"]- f2 (hI))

( > 10 s, As defined in Hote 1, 1

+

Revised Thermal Design Procedure (RTDP) Technical Speci& cation change under revie~ by the NRC (L-95-131 and L-95-250)

TABLE 2.2-1 Continued TABLE NOTATIONS Continued NOTE 3:

(Continued)

=~'F for T > T" 0 for T<T",

As defined in Note 1, O.OOI (

NOTE 4:

avg 0

(leMLV)

As defined in Note 1, and 577, 2 f

( N c>~ 'r a.l To.v~

a.4 kA;74f0 WHIRR.rnht 7c mph'.)

fq (hI)

=

0 for all hI The c nnel's aaxiaua trip setpoint shall not exceed its computed trip setpoint by more than of instrument span.

Revised Thermal Design Procedure (RTDP) Technical Specification change under review by the NRC (L-95-131 and L-95-250)

POW R'TR UTION IM TS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR -

F (Z)

L M T NG COND T ON FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

L Fq(Z)

< ~[F ]

X [K(Z)] for P > 0.5 p

F~(Z)

< ~[F ]

X [K(Z)] for P < 0.5 o ~ 5 (i~Smw) p

),'; y

~+

RA~L~

where:

[F ]

~~w't c

~~<+~++

~<<( A'eO Rated Thermal Power Co g.+

oP<RAMtMC LImt'YS CPo av.

[F~]

The Measured

Value, and K(Z) for a given core height, is specified in the K(Z) curve, defined in the CORE OPERATING LIMITS REPORT.

A~PCAB: MODE 1

KIL9H:

b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced power limit required by ACTION a.,

above; THERMAL POWER may then be increased provided F~(Z) is demonstrated through incore mapping to be within its limit.

With the measured value of F (Z) exceeding its limit:

0 a.

Reduce THERMAL POWER at least IX for each I/ Fq(Z} exceeds F~(Z}

M L

within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may pro-ceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower Oelta-T Trip Setpoints (value of K4) have been reduced at least 1X for each 1X F~(Z) exceeds the.F~(Z);

L and TURKEY POINT - UNITS 3 5 4 V

3/4 2-4 AMENDMENT NOS.

AND

I 4

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f t

p, 'r~

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4 ~

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~ gag ~,fg

~ ~ Ik ~ h

~ 8 Q-flag (

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Cap~ (0

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I

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 F

H shall be limited by the following relationship:

F~

( ~[1.0 +~(l-P)j, 7-'Tf'here:

(IVSCLV)

F +~

a.4 QP 'TKD

+g <41 41+cJ I W

+ke C.~ Q~

MP W~~ Ti',

THERMAL POWER APPLICABILITY:

MODE 1.

b.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limit, verify through incore flux mapping that F~ has been restored to within the above limit, or reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

't c.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2.

and/or b., above; subsequent POWER OPERATION may proceed provided that F~ is demonstrated, through incore flux mapping, to be within the. limit of acceptable operation prior to exceeding the following THERMAL POWER leveis:

1.

A nominal 50K of RATED THERMAL POWER, 2.

A nominal 75X of RATED THERMAL POWER, and 3.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95% of RATED THERMAL POWER.

ACTION:

N y w~

, With F>H exceeding its limit:

\\

a 5 5pc<l I~

a.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

cvpggÃTi&6,

\\<

1.

Restore F~ to within the above limit, or 2.

Reduce THERMAL POWER to less than 50K of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55K of RATED THERMAL POWER within the next 4 hourp.

TURKEY POINT - UNITS 3 8

4 3/4 2-11 AMENDMENT NOS. 137AND 132

POWER DISTRIBUTION LIMITS 3/4.2.5 ONB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related following limits:

parameters shall be maintained within the a.

Reactor Coolant System T

b.

Pressurizer Pressure c.

Reactor Coolant System Flow APPLICABILITY:

MODE l.

ACTION:

< 576HPF

> 2209 sig", and gpm Cgism f OO &

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

SURVEILLANCE RE UIREMENTS

( st.)

l<SQJaw A,'YWAcHm~~~ (H~g

)

4.2. 5.1 rs.

y ~4.2.

Q2 The RCS flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.5 3

l<SCg.W ATTAC.Hm<~~ ( Iqggg)

( ss.)

"Limit not applicable during either a THERMAL POWER ramp in excess of SX of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10K of RATED THERMAL POWER.

TURKEY;POINT - UNITS 3 8 4 3/4 2-16 AMENDMENT NOS.137 AND 132

(,-)

Revised Thermal Desi.gn Procedure (RTDP) Techni.cal Specification change under review by the NRC (L-95-131 and L-95-250)

INSERT TO TS SURVEILLANCE RE UIREMENTS 4. 2. 5. 1 4. 2. 5. 4 4.2.5.1 Reactor Coolant System Tavg and Pressurizer Pressure shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 RCS flow rate shall be monitored for degradation at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.3 4.2.5.4 The RCS flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

After each fuel loading, and at least once per 18 months, the RCS flow rate shall be determined by precision heat balance after exceeding 90%

RATED THERMAL POWER.

The measurement instrumentation shall be calibrated within 90 days prior to the performance of the calorimetric flow measurement.

The provisions of 4.0.4 are not applicable for performing the precision heat balance flow measurement.

• Note to reviewer:

The above insert is identical to the wording currently under review by the NRC for the Revised Thermal Design Procedure Technical Specification change (L-95-131 and L-95-250).

FUNCTIONAL UNIT

'ABL 3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ALLOWABLE VALUE TRIP SETPOINT 1.

Safety Injecti on (Reactor Trip, Turbine Trip, Feedwater Isolation, Control Room Ventilation Isolation, Start Diesel Generators, Containment Phase A Isolation (except Manual SI),

Containment Cooling Fans, Containment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Water) a.

Manual Initiation b.

Automatic Actuation Logic c.

Containment Pressure High d.

Pressurizer Pressure Low e.

High Differential Pressure Between the Steam Line Header and any Steam Line.

f.

Steam Line FlowHigh N.A.

N.A.

<4.5 psig

>1712 psig

<114 psig

<A function defined as follows: A h corres-ponding to~

steam flow at OX load increasing linearly from 20X load to a val e co responding to steam flow at fu l load.

N.A.

N.A.

<4.0 psig

>1730 psig

<100 psi

<A function defined as follows: A hP corres-ponding to 40X steam flow at OX load in-creasing linearly from 20X load to a value sponding to steam flow at full oad.

FUNCTIONAL UNIT TABLE 3.3-ontinued)

ENGINEERED SAFETY FE RES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ALLOWABLE VALUE TRIP SETPOINT 4.

Steam Line Isolation (Continued) b.

Automatic Actuation Logic and Actuation Relays c.

Containment Pressure High-High Coincident with:

Containment Pressure High d.

Steam Line FlowHigh N.A.

<22.6 psig

<4.5 psig

<A function defined as

-follows: A corres-ponding to steam flow at 0% load increasing linearly from 20X load to a val e

corresponding to ~%

steam flow at full oad.

N.A.

<20.0 psig

<4.0 psig

<A function defined follows: A hP corres-ponding to 40X steam flow at OX load in-creasing linearly from 20X load to a val corresponding to steam flow at full load.

lt+%

Coincident with:

Steam Line Pressure Low 01 T,,Low 5.

Feedwater Isolation a.

Automatic Actuation Logic and Actuation Relays b.

Safety Injection

>588 psig

>542.5'F N.A.

See Item l. above for all Safety Injection Allowable Values.

>614 psig

>543'F N.A.

See Item I. above for all Safety Injection Trip Setpoints.

dd I

dd

~GgQR SA INS U

U S

CUA 0

0 S

0 S

UC OA U

5.

Feedwater Isolation (Continued) c.

Steam Generator Water Level High-High 6.

Auxiliary Feedwater (3)

ALLOWABLE PAL~.E

<81.9X of narrow range instrument span

<BOX of narrow range instrument span a.

Automatic Actuation Logic and Actuation Relays b.

Steam Generator Water Level Low-Low N.A.

of narrow range instrument 8.I S'pan.

N.A.

narrow range instrument span.

(o%

c.

Safety Injection d.

Bus Stripping e.

Trip of All Hain Feedwater Pump Breakers 7.

Loss of Power See Item l. above for all Safety Injection Allowable Values.

See Item 7. below for all Bus Stripping Allowable Values.

N.A.

See Item I; above for all Safety Injection Trip Setpoints.

See Item 7. below for all Bus Stripping Trip Setpoints.

N.A.

a.

4.16 kV Busses A and B

(Loss of Voltage)

N.A.

N.A.

Rev3.sed Therma1 Des'.gn Procedure (RTDP) Technica1 Spec3.f Scat'.on change under rev9.ew by the NRC (L-95-131 and L-95-250)

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN LIHITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pr urize e safety valve shall be OPERABLE" with a lift setting of 2485 psig e APPLICABILITY:

MODES 4 and B.

9 nL /

ACTION:

With no pressurizer Code safety valve OPERABLE, immediately suspend all opera-tions involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE RE UIREMENTS 4.4.2.1 No additional requirements other than those required by Specification 4.0.5.

(. iwSWle.r U 0 L<eS 4-e 4~

WLu.'S ~

~eVe.

~~4y~, ~4s

+k.~+

o ~e

~i ~4w "While in NODE 5, an equivalent size vent pathway may be used provided that the vent path~ay is not isolated or sealed.

""The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

TURKEY POINT - UNITS 3 8i 4 3/4 4-7 AMENOHENT NOS,137ANO 132

REACTOR COOLANT SYSTEH OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressuri e

Code afety valves shall be OPERABLE with a lift setting of 2485 psig APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

With one pressurizer Code safety valve inoperable, either restore the inoper-able valve to OPERABLE 'status within 15 minutes 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 at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREHENTS 4.4.2.2 No additional requirements other than those required by Specification 4.0.5.

's-left"

/-

"The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

TURKEY POINT - UNITS 3 a 4 314 4-8 AMENDMENT NOS 137AND 132

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL'IRCUMFERENTIALWELD INITlAL RTNPT1 10'F IR SERVICE PERIOD: W EFPY RT NDT 1/

THICKNESS HEATUP RA ES:

UP TO 60 F/HR RTNDT m >/< THICKNESS ~ 200.4'F NOTE:

NO MARGINS ARE GIVEN FOR POSSIBLE INSTRUMENT ERRORS.

LEAK TEST LIMIT 17SO 15CO 1250 C) 1000 I

750 HR RATIO CRITICALITY LIMIT fBASED ON INSERVlCE HYDRO-STATlC TEST MP TURE 380oF FOR S RVICE OD UP TO ACCEPTAB OPERATION 0

SO 100 ISO 200 250 3N ZO 400 450 S00

<Am) TEMPERATURE (g FIGURE 3.4-2 TURKEY POINT UNITS 3 4 4 TURKEY POINT - UNITS 3 4 4 REACTOR COOLANT SYSTEM HEATUP LI ITATIONS (60 F/hr) - APPLICABI.E UP TO EFPY i 'I 3/4 4-31 AMENDMENT NOS.137 AND 132

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: QRCUMFERENTIAL WELD INITIALRTNDT'D 19 SERVICE PERIOD:~20 PT RTNDT ~ I/O THICKNESS ~ 2'P HEATUP RATES:

UP TO 100 F/HR RT NDT ~ ~/4 THICKNESS ~ 200,4 NOTE:

NO MARGINS ARE GIVEN FOR POSSIBLE INSTRUMENT ERRORS.

LEAK TEST LIMIT 1750 1500 1250 0

~0te K

HEATUP RATES UP TO 100'F HR I

ON CRlTICALITY LIMIT IfBASED ON INSERVICE HYDRO-STATIC TEST TEMPERATURE (380 F) FOR THE SERVICE PERIOD UP TO EFPY ACCEPTABLE OPERATION ACCEPTABI.E OPERATION 0

50 100 ISO 200 250,XO 350 400 450 500 INDICATED TEMPERATURE (F)

FIGURE 3.4-3 TURKEY POINT - UNITS 3 8 4 TURKEY POINT UNITS 3 4 4 REACTOR COOLANT SYSTEM HEATUP LI ITATIONS (1000F/hI') - APPLICABLE UP TO EFPY 3/4 4-32 AMENDMENT NOS.137 AND l32

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD INITIAL RTNDT'0'F RTNPT ~1/4 THICKNESS 2m F SERVICE PERIOD:

RTNPT ~3/4 THICKNESS ~ 200o4 F COOLDOWN RATE:

P TO 100'F/HR NOTE:

NO MARGINS ARE GIVEN FOR POSSIBl.E INSTRUMENT ERRORS, I

I I

s I

I

'l75Q t/l i5N N

CC a

1250 l000 UNACCEPTABLE OPERATION COOLDOWN RATES F/HR 0

40 60 100 0

0 SO 100 ISO 200 250 XO QO 400 450 5N INDE TEMPERATURE ('Fj FIGURE 3.4-4 TURKEY POINT UNITS 3 4 4 TURKEY POINT - UNITS 3 4 4 REACTOR COOLANT SYSTEM COOLPOW L

ITATIONS (1004F/hr) - APPLICABLE UP TO EFPY 3/4 4-33 NENPHENT NOS.l37ANP 132

EMERGENCY CORE COOLING SYST96 SURVEILLANCE RE UIREHENTS 4.5.2 Each ECCS component and flow path shall be deaonstrated 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 by control rooa indication that the following valves are in the indicated positions with power to the valve operators reaovcd:

Valve Nusber Valve Function Valve Position 864A and 8 Supply froa fNST to ECCS Open 862A and 8

%ST Supply to RHR puIIps Open 863A and 8 RHR Recirculation Closed 866A and 8 H.H.S.I. to Hot Legs Closed HCV-758*

RHR HX Outlet Open To perait teaporary operation of these valves for surveillance or maintenance

purposes, power say be restored to these valves for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

At least once per 31 days by:

1)

Verifying that the ECCS piping is full of water by venting the ECCS puap casings and accessible discharge piping, 2)

Verifyi that each valve (aanual, power-operated, or autoaatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, and 3)

Verifying that each RHR Puep develops the indicated differential pressure, applicable to the operating conditions in accordance with Figure 3.5-1 when tested pursuant to Specification 4.0.5.

c.

At leist once per 92 days by:

1)

Verifying that each SI pulp develops the indicated differential pressure applicable to the operating conditions when tested pursuant to Specification 4.0.5.

'I pulp psid at a aetered flowrate > 300 gpe (normal al yaent or Unit 4 SI puaps aligned to Unit 3 NST), or psid at a aetered flowrate > 28O gpe 7Un t 3 SI pumps aligned to Unit 4 NST).

"Air Supply to HCV-758 shall be verified shut off and sealed closed once per 31 days.

TURKEY POINT - UNITS 3 4 4

'3/4 5-5 NEl8%NT NOS.138 ANO 133

CONTAINMENT SYSTEMS EMERGENCY CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 Three emergency containment cooling units shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTTON:

With one of the above required emergency containment cooling units inoperable restore the inoperable cooling unit to OPERABLE status ithin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With two or more of the above required emergency containment cooling units inoperable, restore at least two cooling units 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 the next 6

hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Restore

~ all of the above required cooling units to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.6.2.2 Each emergency containment cooling unit shall be demonstrated OPERABLE:

a.

At least once per 31 days by starting each cooler unit from the control room and verifying that each unit motor reaches the nominal operating current for the test conditions and operates for at least 15 minutes.

b.

At least once per 18 months by:

1)

Verifying that each-ttaA star automatically on a safety injection (SI) test signal, an 2)

Verifying a cooling water flow rate of greater than or equal to 2000 gpm to each cooler.

QWo t wev.~e~

~~4M'Lh ~<~5

~nile y M~A 0 S TURKEY POINT - UNITS 3

8 4

3/4 6-14 AMENDMENT NOS. 137AND 132

TABLE 3.7-1 HAXINUM ALLOWABLE OWER LEVEL WITH STEAN L NE SAFETY VALVES OURING THREE LOOP OPERATION NXIMUN NUNBER OF INOPERABLE SAFETY VALVES ON ANY OP 8

8 E

TOR SLXINUtl ALLOWABLE POWER LEVEL C

OF RA THE A

OWE 2

14 VA E NUMB S

A

~LJI

~Lo C

TABB

. 7-2 SA V

S ING ORIFICE SIZE i*

JHEKJILNH 3%

1.

RV1400 RV1405 RV1410 2.

RV1401 RV1406 RV1411 3.

RV1402 RV1407 RV1412 4.

RV1403 RV1408 RV1413 1085 psig 1100 psig 1115 psig 1130 psig 16 16 16 16

~~+)~) p $ 5 QpaS Q.~

w< 44K I

/~

We 4 4'~~

g i~~

llsW~

(~

~~biz

• The l)ft setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

TURKEY POINT - UNITS 3 K 4 3l4 7-2

~ENDgENT NOS 172 AND 166

PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7. 1.3 The condensate storage tanks (CST) system shall be OPERABLE with:

0 posite Unit in MODES 4 5 or 6

~nrr indicated water volume storage tanks.

0 osite Unit in MODES 1 2 or

~indicated water volume APPLICABILITY:

MODES 1, 2 and of &~99 gallons in either or both condensate

+go, oeO of 8%~9 gallons.

3.

ACTiON:

0 osite Unit in MODES 4 5 or 6 With the CST system inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> restore the CST system to OPERABLE status or be in at least HOT STANDBY in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, g f cD, DNca 0 posite Unit in MODES 1

2 or 3

IACA+dt$y~

1)

With the CST system inoperable d

o less than 87~9 gallons, but greater than or equal to gall 1thin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> re tore the inoperable CST system to OPERABLE status or ace one unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

tucks Cgk.~

g2 I ~, ~~a-r lvk.dt C.s' 2)

With the CST system inoperable with less than &5-999. gallons within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the CST system to OPERABLE status or e in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

This ACTION applies'to both units simultaneously.

TURKEY POINT - UNITS 3 8

4 3/4 7-6 AMENDMENT NOS.

137@NO 132

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued L~e4~~

4.7. 1.3 The condensate storage tank (CST) system shall be demonstrated OPERABLE 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 the ~':a+wed-water volume is.

within its limit when the tank is the supply source for the auxiliary feedwater pumps.

TURKEY POINT - UNITS 3 8

4 3/4 7-7 AMENDMENT NOS.137 AND 132

A S

S TANDB ST AM G E

TOR FEEDM T R

S ST LI ITING CONDITION FOR OPERATION l3g, ohio 3.7.1.

Two Standby Steam Gen rator Feedwater Pumps shall be OPERABLE and at least gallons of water

, shall be in the Demineralized Mater Storage Tank

(>~~4)

PPL

~CT~IO:

MODES 1, 2 and 3

(i~di'~~

vo<

e) a.

Mith one Standby Steam Generator Feedwater Pump inoperable, restore the inoperable pump to available status within 30 days or submit a SPECIAL REPORT per 3.7.1.6d.

b.

With both Standby Steam Generator Feedwater Pumps inoperable, restore at least one pump to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or:

C.

d.

1.

Notify,the NRC within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and provide cause for the inoperability and plans to restore pump(s) to OPERABLE status

and, l

~

2.

Submit a

S AL REPORT per 3.7.1.6d.

I PS'on With less tha gallons of water n the Demineralized Ma r Storage Tank restore the available volume to at least gallons, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or submit a SPECIAL REPORT per 3.7.1.6d.

If a SPECIAL REPORT is required per the above specifications submit a

report describing the cause of the inoperability, action taken and a

schedule for restoration within 30 days in accordance with 6.9.2.

SURVEILLANCE RE UIREM S

4.7.1.6.1 The Demineralized Mater Storage tank water volume shall be determined to be within limits at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.7.1.6.2 At least monthly verify the Standby Steam Generator Feedwater Pumps are OPERABLE by testing in recirculation on a

STAGGERED TEST BASIS.

4.7. 1.6.3 At least once per 18 months, verify operability of the respective standby steam generator feedwater pump by starting each pump and providing feedwater to the steam generators.

• These pumps do not require plant safety related emergency power sources for operability and the flowpath is normally isolated.

~The Demineralized Water Storage Tank is non-safety grade.

TURKEY POINT - UNITS 3 5 4 3/4 7-11 AMENDMENT NOS. 164AND 158

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

])

Verifying that the air cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of greater than or equal to 99K OOP and halogenated hydrocarbon removal at a system flow rate of 1000 cfm ale.

2)

Verifying, within 31 days after removal, that-mlaboratory analysis of a representative carbon sample obtained in. accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and analyzed per ANSI N510-1975, meets the criteria for methyl iodine removal efficiency of greater than or equal to QS)or the charcoal be replaced with charcoal that meets or exceeds the criteria of position C. 6. a. of Regulatory Guide l. 52 (Revision 2),

and d.

3)

Verifying by a visual inspection the absence of foreign materi a 1 s and gas ket deter i orat ion.

At least once per 12 months by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 1000 cfm +10%;

At least once per 18 months by verifying that on a Containment Phase "A" Isolation test signal the system automatically switches into the recirculation mode of operation.

TURKEY POINT - UNITS 3 4 4 3/4 7-17 AMENDMENT NOS.137 AND 132

CO I BF 0

0 (Continued)

Factor Limit Report, the Peaking Factor Limit Report shall be provided to the NRC Document Control desk with copies to the Regional Administrator and the Resident Inspector within 30 days of their implementation, unless otherwise approved by the Commission.

l The analytical methods used to generate the Peaking Factor limits shall be those previously reviewed and approved by the NRC. If changes to these methods are deemed necessary they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for review and approval prior to their use if the change is determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation.

0 0

G S

0 6.9.1.7 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

I.

Axial Flux Difference for Specifications 3.2.1.

2.

Control Rod Insertion Limits for Specification 3.1.3.6.

3 a

Flux Ho annel Factor - F Z for S eci ic t

+.,Al~g(~~ +~~I i'se.

H~4 CM~l F~c.~

The an u e er e

m ts s a

previously reviewed and approved by the NRC in:

1.

MCAP-10216-P-A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL Fo SURVEILLANCE TECHNICAL SPECIFICATION,'une 1983.

2.

MCAP-8385,

'POMER DISTRIBUTION CONTROL AND LOAD FOLLOMING PROCEDURES

- TOPICAL REPORT,'eptember 1974.

Q ~/i Qg Oh.

The analytical methods used to determine the K(1) curve shall be those previously reviewed and approved by the NRC in:

l.

MCAP-9220-P-A, Rev.

1, 'Mestinghouse ECCS Evaluation Nodel - 1981 Version,'ebruary 1982.

2.

MCAP-9561-P-A, ADD. 3, Rev.

1,

'BART A-1: A Computer Code for the Best Estimate Analysis of Reflood Transients - Special Report:

Thimble Nodel n

a o

IWS+g A~v'AC.H~cw~

lg+~

The analytical methods us e

rm n e

o an r ion Limits shall be those previously reviewed and approved by the NRC in:

1.

MCAP-9272-P-A, 'Mestinghouse Reload, Safety Evaluation Nethodology,'uly 1985.

The ability to calculate the COLR nuclear design parameters are demonstrated in:

1.

Florida Power L Light Company Topical Report NF-TR-95-01, 'Nuclear Physics Nethodology for Reload Design of Turkey Point 8 St. Lucie Nuclear Plants'.

TURKEY POIN1' NITS 3 K 4 6-20.

AlKNDNENT NOS. l74 AND l68

INSERT TO TS 6.9.1.7

[NOTE:

References 3 and 4 are included in the Proposed License

. Amendments for the Small Break Loss-of-Coolant (SBLOCA) Re-

analysis, as transmitted to the NRC via L-95-193.]

5.

6.

WCAP-10266-P-A, Rev.

2 (proprietary) and WCAP-11524-NP-A, Rev.

2 (non-proprietary),

"The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code,"

May 1988.

NTD-NRC-94-4143, "Change in Methodology for Execution of BASH Evaluation Model," May 23, 1994.

(Continued Topical Report NF-TR-95-01 was approved by the HRC for use by Florida Power g,

Light Company in:

l.

Safety Evaluation by the Office of Nuclear Reactor Regulations Related to Amendment Ho.

174 to Facility Operating License DPR-31 and Amendment Ho. 168 to Facility Operating License Ho. DPR-41, Florida Power i Light Company Turkey Point Units Nos.

3 and 4, ocket Nos. 50-250 and 50-251.

r~ (+)

The AFD,K(Z), and Rod Bank Insertion, Limits shall be determined such that, all applicable 1imits of the safety analyses are met.

The CORE OPERATIHQ LINITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector, unless otherwise approved by the Coaeission.

6.9.2 Special'eports shall be submitted to the Regional Administrator of the Regional Office of the HRC within the time period specified for each report as stated in the Specifications within Sections 3.0, 4.0, or 5.0.

6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time interval at each power level; b.

Records and logs of principal maintenance activities, inspections,

repair, and replacement of principal items of equipment related to nuclear safety; c.

All REPORTABLE EYENTS; d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; e.

Records of changes made to the procedures required by Specification 6.8.1; f.

Records of radioactive shipments; g.

Records of sealed source and fission detector leak tests and results; and h.

Records of annual physical inventory of all sealed source material of record.

6.10.3 The following records shall be retained for the duration of the unit Operating License:

TURKEY POINT - UNITS 3 K 4 6-20a hNEMWBA'0$. 174 NS 168

2. 1 SAFETY LIMITS BASES

2. 1. 1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

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 been related to DNB.

This relationship has been developed to predict the ONB flux and the location of ONB for axially uniform and nonuniform heat flux distribu-tions.

The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause ONB at a particular core location to the local heat flux and is indicative of the margin to DNB.

The ONB design basis is as follows:

there must be at least a 95 percent probability with 95 percent confidence that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the ONBR limit of the ONB correlation being used.

The correlation ONBR limit is established based on the entire applicable experimental data set such that there is a

95 percent probability with 95 percent confidence that ONB will not occur when the minimum DNBR is at the DNBR limit, The curves of Figure 2. 1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum ONBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor, F

ef-~2 and and a reference cosine with a pe~k of 1.55 for axial power shape.

n a owance is included for an increase in F>H at reduced power based on the expression:

F ~ [1+~~(1-P)]

Where P is the fra i

of RATED THERMAL POWER; A~TACH n Zw~ (i-j<~)

These lim>t ng ea ux cond>>on are

>g r than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion limit assuming the axial power imbalance is within the limits of the f (DI) function of the Overtemperature trip.

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature DT trips will reduce tne setpoints to provide protection consistent with core Safety Limits.

TURKEY PO.'t<T -

UNITS 3 8

4 8 2-1 AMENDMENT NOS.137AND 132

0

INSERT TO TS BASES 2.1.1 RTP FaH F~H limit at RATED THERMAL POWER as specified in the CORE OPERATING LIMITS REPORT.

Power Factor multiplier for FzH as specified in the CORE OPERATING LIMITS REPORT.

LIMITING SAFETY SYSTEM SETTINGS BASES Undervolta e and

- 4. 16 kV Bus A and B Tri s (Continued) power the Undervoltage Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine first stage pressure at approximately 10~ of full power equivalent);

and on increasing power, reinstated automatically by P-7.

Turbine Tri A Turbine trip initiates a Reactor trip.

On decreasing power, the Reactor Trip from the Turbine trip is automatically Clocked by P-7 (a power level of approximately 10~ of RATED THERMAL POWER with a turbine first stage pressure at approximate>y 30 c of full power equivalent);

and on increasing

power, reinstated automatically by P-7.

Safet In'ection In ut from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.

The ES instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

I Reacto~

Coolant Pumo Breaker Position Tri lv The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against ONB.

The open/close position trips assure a reactor tri signal is enerated before the lo lo trip s

oint is reached.

Their u

rona a

).

> y.at the open close os n se ngs is required to enhance the overall reliability of the Reactor Protection System.

Above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine first stage pressure at approximately lOX of full power equivalent) an automatic reactor trip will occur if more than one reactor coolant pump breaker is opened.

Above P-8 (a power level of approximately 45K of RATED THERMAL POWER) an automatic reactor trip will occur if one reactor coolant pump breaker is opened.

On decreasing power between P-8 and P-7, an automatic reactor trip will occur if more than one reactor coolant pump breaker is opened and below P-7 the trip function is automatically blocked.

Underfrequency sensors are also installed on the 4. 16 kV busses to detect underfrequency and initiate breaker trip on underfrequency.

The underfrequency trip setpoints preserve the coast down energy of the reactor coolant

pumps, in case of a grid frequency decrease so DNB does not occur.

TURKEY POINT - UNITS 3

8; 4 B 2-7 AMENDMENT NOS.l 37 AND132

3/4. 2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(1) maintaining the minimum DNBR in the core greater than or equal to the applicable design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 22004F is not exceeded.

The definitions of certain hot channel and peaking factors as-used in these specifications are as follows:

F~(Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and Fx (Z)

Radial Peaking Facto~,

is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

lay t" 9 J~CI'~e.d 3/4. 2. 1 AXIAL FLUX DIFFERENCE a pcs,AW,~CL \\ ~l'7 5

'R<P~

gY'he limits on AXIAL FLUX DIFFERENCE (AFD) assure that the FD(Z)~

Scend-e~Ae times the normalized axial peaking factor is not exceeded ruing either norma operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions.

Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect core bur nup considerations.

TURKEY POINT - UNITS 3 8L 4 B 3/4 2-1 AMENDMENT NOS.137AND 132

1

POWER DISTRIBUTION LIMITS BASES 3/4, 2. 2 and 3/4. 2. 3 MEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that:

(1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 22004F ECCS acceptance criteria limit.

The LOCA peak fuel clad tempera imi be sensitive umb r earn enerat r tubes plugged.

F (Z), Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux.

F~ Nuclear Enthal Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

jhis periodic surveillance is sufficient to ensure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than a 12 steps, indicated, from the group demand position; b.

Control rod groups are sequenced with overlapping groups as described in Specification

3. 1.3.6; c.

The control rod insertion limits of Specifications

3. 1.3.5 and

3. 1.3 '

are maintained; and d.

The axial power distribution, exp~essed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

When an F~ measurement'is

taken, both experimental error and manufacturing tolerance must be allowed for.

Five percent is the appropriate allowance for a full core map taken with the movable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.

These uncertainties only apply if the map is taken for purposes other than the determination of PBL an H'~

will be maintained within its limits provided Conditions

a. through d.

above are maintained.

In the specified limit of F

, there is an 8 percent allowance for uncertainties which means that normal operation of the core is expected to result in F~

1. 08.

The logic behind the larger uncertainty in this

) ~g<p ~

P,~~AC.L-l m<m~

TURKEY POINT - UNITS 3 4 4 B 3/4 2-4 AMENDMENT NOS237 AND 132

INSERT TO TS BASES 3 4.2.2 AND 3 4.2.3 where F

is the F

limit at RATED THERMAL POWER (RTP)

RTP N

aH aH specified in the CORE OPERATING LIMITS REPORT.

POWER DISTRIBUTION LIMITS BASES 3/4.2.4 UAORANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power dis-tribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

The limit of 1.02, at which corrective action is required, provides ONB and linear heat generation rate protection with x-y plane power tilts.

A limit of 1.02 was selected to provide an allowance for the uncertainty asso-ciated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod.

In the event such action action does not correct the tilt, the margin for uncertainty on F

(Z) is reinstated by reducing the maximum allowed power by 3X for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors or incore thermocouple map are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.

The two sets of four symmetric thimbles is a unique set of eight detector locations.

These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-ll, N-8.

3/4. 2. 5 DNB PARAMETERS 56(,k

~F'he limits on the ONB-related arameters assure that each f the param-eters are maintained within the no al steady-state envelope o

operation assumed in the transient and acci ent analyses.

The limits a e consistent with the initial FSAR assumptions and have been analytically emonstrated ade-quate to maintain a minimum DNBR above the applicable esi limits throughout each analyzed ent.

The i dicated T

value of M<~ and the indicated PI 7 ressurizer r

value of psig correspond to analytical limits of F and psi res ectively, with al r measuremen't uncertainty.

mqag~<Q 355 ohio (g +, OOIb Q.a la&i~4;i c The RCS flow value'of pm corr ponds o an ana y real limit of pm which is assumed to have a 3.5X measurement uncertainty.

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 expecte'd transient operation.

trd5egv A~1 Ac+mEw~

( ld C<E )

TURKEY POINT - UNITS 3 8

4 B 3/4 2-8 AMENDMENT NOSl37 ANO132

v

XNSERT TO TS BASES 3 4.2.5 The 18-month periodic measurement of the RCS total flow rate is adequate to ensure that. the DNB-related flow assumption is met and to ensure correlation of the flow indication channels with measured flow.

Six month drift effects have been included for feedwatez temperature, feedwater flow, steam pressure, and the pressurizer pressure inputs.

The flow measurement is performed within ninety days of completing the cross-calibration of the hot leg and cold leg narrow range RTDs.

The indicated percent flow surveillance on a 12-hour basis will provide sufficient verification that flow de radation has not occurred.

t (loJQK%'T)-

An indicated percent flow which is greater than the thermal design flow plus instrument channel inaccuracies and parallax errors is acceptable for the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance on RCS flow.

To minimize measurement, uncertainties it is assumed that the RCS flow channel outputs are averaged.

• Note to reviewer:

The above insert was originally submitted to the NRC for review of the Revised Thermal Design Procedure Technical Specification change (L-95-131).. The marked-up changes represent the subsequent proposed changes in accozdance with the Uprate submittal.

REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735'psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e.,

no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves.

In Mode 5 only one pressurizer code safety is required for overpressure protection.

In lieu of an actual operable code safety valve, an unisolated and unsealed vent pathway (i.e.,

a direct, unimpaired opening, a vent pathway with valves locked open and/or power removed and locked on an open valve) of equivalent size can be taken credit for as, synonymous with an OPERABLE code safety.

Demonstration of the safety valves'ift settings will occur only during shutdown and will be performed in accordance with thh provisions of Section XI of the ASME Boiler and Pressure Code.

( i<sQP ~)

3/4.4.3 PRESSURIZER The 12-hour periodic surveillance is sufficient to ensure that the maximum water volume parameter is restored to within its limit following expected transient operation.

The maximum water volume (1133 cubic feet) ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.

The requirement that both backup pressurizer heater groups be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

li 44 p mssuv-. t.~

~ J<

so-~bp

'J~.

t a.'il~ /

+

~P

'mfa

~y s

/

+~ f~a~Q.Q

'f~

~pCp Al IL( CQ'~MK~~

+4-q~hvez a,v-~

res~+

+

~, 44'~

+

( o(~

Su Pv Qi I I + ~~

TURKEY POINT - UNITS 3 8( 4 B 3/4 4-2 AMENDMENT NOS.137 AND 132 The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 293,330 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 RCS vent opening of at least 2,50 square inches will provide overpr essure relief capability and will prevent RCS overpressurization.

In addition, the Overpressure Mitigating System provides a diverse means of protection against RCS overpressurization at low temperatures.

0

REACTOR COOLANT SYSTEM BASES PRESSURE TEMPERATURE LIMITS (Cont1nued) 1.

The reactor coolant temperature and pressure.and system heatup and cooldown rates (w1th the exception of the pressur1zer) shall be limited in accordance with Figures 3.4-2 to 3.4-4 for the service period specified thereon:

(~

(.)

a.

Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.

Limit 11nes for cooldown rates between those presented may be obtained by interpolation; and 2;

b.

Figures 3.4-2 to 3.4-4 def1ne 11mits to assure prevent1on of non-ductile failure only.

For normal operation, other inherent plant characteristics, e.g.,

pump heat add1tion and pressurizer heater

capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

fJ C3 These limit l1nes shall be calculated per1od1cally using methods provided

below, 3.

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F, 4.

The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200'F/h, respectively.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid 1s greater than 320'F, and 5.

System preservice hydrotests and 1nservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel-are determined in accordance with the NRC Standard Review Plan, the version of the ASTM E185 standard required by 10 CFR 50 Appendix H, and in accordance with add1tional reactor vessel requirements.

0 The propert1es are then evaluated in accordance with Appendix 6 of the 1983 Edit1on of Section III of the ASME Boiler and Pressure Vessel Code and the additional requirements of 10 CFR 50, Appendix 6 and the. calculation

-methods descr1bed in Mestinghouse Report GTSD-A-le12, 'Procedure for Developing Heatup and Cooldown Curves.'eatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT>>T, at the end of

@/effective fu11 power years (EFPY) of service life.

Th EFPY service 11fe period is chosen such that the limiting RT>0T at the 4T location in TURKEY POINT - UNITS 3 K 4 B 3/4 4-8 AMENDMENT NOS.170AND 164

0 G3 i

,o'

RE C

OR COOLANT SYST BAS RESSUR RE I S (Continued) the core region is greater than the RT<pT of the 11miting un1rradiated material.

The selection of such a limiting RT>pT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The heatup and cooldown limit curves, Figures 3.4-2, 3.4-3 and 3.~ are composite curves prepared by determining the est conservative case with either the inside or outside wall controlling, fol any heatup rate up to 100 degrees F per hour and cooldown rates of up to 100 degrees F per hour.

The heatup and cooldown curves were prepared based upon the most 11miting value of predicte adjusted reference temperature at the end of the applicable service period EFPY).

The reactor vessel materials have been tested to determine their initial RTMpT'he results of these tests are shown in Tables B 3/4.4-1 and B 3/4.4-2.

Reactor operation and resultant fast neutron (E greater than 1 NeV) irradiation can cause an increase in the RT>pT.

Therefore, an adjusted reference temperature, based upon the fluence and chemistry factors of the material has been predicted using Regulatory Guide 1.99, Revision 2, dated Nay 1988, 'Radiation Embrittlement of Reactor Vessel Naterials.'he heatup and cooldown limit curves of Figures 3.4-2, 3.4-3, and 3.4-4 include predicted adjustments for this sh1ft 1n RT<pT at the end of the applicable service period.

The actual shifts in RT>pT of the vessel materials will be established periodically during operation by removing and evaluating, in accordance with the version of the ASTN E185 standard required by 10 CFR 50 Appendix H, reactor vessel material 1rradiat1on surveillance specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

Since the limiting beltline materials (Intermediate to Lower Shell Circumferential Meld) in Units 3 and 4 are identical, the RV surveillance program was integrated and the results from capsule testing is applied to both Units.

The surveillance capsule 'T'esults from Unit 3 (MCAP 8631) and Unit 4 (SMRI 02-4221) and the capsule 'V'esults from Unit 3 (SMRI 06-8576) were used with the methodology 1n Regulatory Guide 1.99, Revision 2, to provide TURKEY POINT - UNITS 3 K 4 B 3/4 4-9 ANENDNENT NOS.l70 AND 164

(

f 0

CONTAINMENT SYSTEMS BASES CONTAINMENT VENTILATION SYSTEM (Continued) resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop.

The 0.60 L

leakage limit of Specifica-tion 3.6. 1.2b.

shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3/4. 6. 2 OEPRESSURIZATION ANO COOLING SYSTEMS I

3/4. 6. 2. 1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization capability will be available in the event of a LOCA.

The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses.

The allowable out-of-service time requirements for the Containment Spray System have been maintained consistent with that assigned other inoperable ESF equipment and do not reflect the additional redundancy in cooling capability provided by the Emergency Containment Cooling System.

Pump performance requirements are obtained from the accidents analysis assumptions.

3/4.6.2. 2 EMERGENCY CONTAINMENT COOLING SYSTEM lQS<wz avv'Ac,H~E~v ( He~)

ITY of the Emergency Containment Cooling ystem ensures t at adequate heat remova c

lable during post-LOCA conditions.

The emergency containment coolers are a ful.

ca nd are. redundant.to the spray system in terms of heat removal function for desi n

as The allowable out-of-service time requirements for the Containment Cooling System have been maintained consistent with that assigned other inoperable ESF equipment and do not reflect the additional redundancy in cooling capability provided by the Containment Spray System.

The surveillance requirement for ECC flow is verified by correlating the test configuration value with the design basis assumptions for system configura-tion and flow.

An 18-month surveilance interval is acceptable

-based on the use of water from the CCM system, which results in a low risk of heat exchanger tube fouling.

3/4. 6. 3 EMERGENCY CONTAINMENT FILTERING SYSTEM The OPERABILITY of the Emergency Containment Filtering System ensures that sufficient iodine removal capability will be available in the event of a LOCA.

The reduction in containment iodine inventory reduces the resulting SITE BOUNOARY radiation doses associated with containment leakage.

The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.

System components are not subject to rapid deterioration.

Visual inspection and operating/performance tests after maintenance, prolonged operation, and at the required frequencies provide assurances of system reliability and will prevent system failure.

Filter performance tests are conducted'in accordance with the methodology and intent of ANSI N510- 1975.

TURKEY POINT - UNITS 3 4 4 B 3/4 6-3 AMENOMENT NOS.137 AND 132

INSERT TO TS BASES 3 4. 6.2.2 The OPERABILITY of the Emergency Containment Cooling (ECC)

System ensures that the heat removal capacity is maintained with acceptable ranges following postulated design basis accidents.

To support both containment integrity safety analyses and component cooling water system thermal analyses, a maximum of two ECCs can receive an automatic start signal following generation of a safety injection (SI) signal (one ECC receives an "A" train SI signal and another ECC receives a "B" train SI signal).

To support post-LOCA long-term containment pressure/temperature

analyses, a maximum of two ECCs aze required to operate.

The third (swing)

ECC is required to be OPERABLE to support manual starting following a postulated LOCA event for containment pressure/temperature suppression.

3 4.7 PLANT SYSTEMS BASES 3 4.7.1.

The OPERABILITf 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.

Steam can be supplied to the pump turbines frex either or both units through redundant steam headers.

Two D.C. motor operated valves and one A.C. motor operated valve on each unit isolate the three main steam lines from these headers.

Both the D.C.

and A.C. motor operated valves are powered from safety-related sources.

Auxiliary feedwater can be supplied through redundant lines to the safety-related portions of the main feedwater lines to each of the steam generators.

Air operated fail closed flow control valves are provided to modulate the flow to each steam generator.

Each steam driven auxiliary feed-water pump has sufficient capacity for single and two unit operation 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 Residual Heat Removal System may be placed into operation.

ACTION statement 2 describes the actions to be taken when both auxiliary feedwater trains are inoperable.

The requirement to verify the availability of both standby feedwater pumps is to be accomplished by verifying that both pumps have successfully passed their monthly surveillance tests within the last surveillance interval.

The requirement to complete this action before beginning a unit shutdown is to ensure that an alternate feedwater train is available before putting the affected unit through a transient.

If no alter-nate feedwater trains are available, the affected unit is to stay at the same condition until an auxiliary feedwater train is returned to service, and then invoke ACTION statement I for the other train. If both standby feedwater pumps are made available before one auxiliary feedwater train is returned to an OPERABLE status, then the affected unit(s) shall be placed 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 HOT SHUTDON within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION stateaent 3 describes the actions to be taken when a single auxiliary feedwater pump is inoperable.

The requirement to verify that two independent auxiliary feedwater trains are OPERABLE is to be accomplished by verifying that the requirements for Table 3.7-3 have been successfully met for each train within the last surveillance interval.

The provisions of Specifica-tion 3.0.4 are not applicable to the third auxiliary feedwater pump provided it has not been inoperable for longer than 30 days.

This means that a unit(s) can change OPERATIONAL MODES during a unit(s) heatup with a single auxiliary feedwater pump inoperable as long as the requirements ef ACTION statement 3 are satisfied.

~)T) ))L )

)C

)

d)

Operation with less than all four MSSVs OPERABLE for each steam generator is permissible, if THERMAL POWER is proportionally limited to the relief capacity of the remaining MSSVs.

This is accomplished by restricting THERMAL POMER so that the energy transfer to the most limiting steam generator is not greater than the available relief capacity in that steam generator.

UX IARY FEEDMATER SYSTE I

5+Ra ACHES W

HCRQ The monthly testing of the auxiliary feedwater pumps will verify their operability.

Proper functioning of the turbine admission valve and the opera-tion of the pumps will demonstrate the integrity of the system.

Verification TURKEY POINT - UNITS 3 L 4 B 3/4 7-2 AMENDMENT NOS,172 AND 166

INSERT TO TS BASES 3 4.7.1.1 Table 3.7-2 allows a +

3% setpoint tolerance for OPERABILITY;

however, the valves are reset to +

1% during the Surveillance to allow for drift.

3 4.7 PLANT SYSTEMS BAsES U I SYSTEM (Continued) of correct operation will be made both from instrumentation within the control room and direct visual observation of the pumps.

3 4.7.

3 C

SATE STORAG TAN lAQ< Ca.

voRu~

Jo,ooo There are two (2 essm cally designed 250,000 gallons condensate torage tanks.

A mfnfmum~of N5-OM gallons is maintained for each unit in NO KS 1, 2 or 3.

The OPERABILI of t e condensate storage tank with the mfnfmum~w~W volume ensures that sufficient water is available to maintain the Reactor Coolant System at HOT STANDBY conditfons for approximately 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> or main-tain the Reactor Coolant System at HOT STANDBY conditions for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and then cool down the Reactor Coolant System to below 350 F at which point the Residual Heat Removal System may be placed fn o erat on.

~~s~g.~

P ~q A.C,H~~~ IRK~

3

.7..4 S

ACTI The limit on secondary coolant specific actfvity is based on a postulated release of secondary coolant equivalent to the contents of three steam genera-tors to the atmosphere due to a net load rejection.

The limiting dose for this case would result from radioactive iodine in the secondary coolant.

One tenth of the iodine in the secondary coolant is assumed to reach the site boundary making allowance for plate-out and retention in water droplets.

The inhalation thyroid dose at the site boundary fs then; Dose (Rem)

C

• V ~ B X/9 ~ 0.1 Where:

C

~

secondary co~lant dose equivalent J-131 specific activity 0.2 curfes/m (pCf/cc) or O.l Cf/m, each unit V

equivalent secondary coolant volume released 214 m3 B

breathing rate 3.47 x 10 m /sec.

-4 3

X/g atmospheric dispersion parameter

~ 1.54 x 10 sec/m O.l equivalent fraction of.activity released DCF dose conversion factor, Rem/Cf The resultant thyroid dose fs less than 1.5 Rem.

3

.7.

The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam.line rupture.

This restriction fs required to:

(1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the

blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.

The 24-hour action time provides a reasonable amount of time to troubleshoot and repair the backup air and/or nitrogen system.

TURKEY POINT - UNITS 3 K 4 B 3/4 7-3 AffENDgENT NOS.172AND 166

0

ZNSERT TO TS BASES 3 4.7.1.3

'I The minimum indicated volume includes an allowance for instrument indication uncertainties and for water deemed unusable because of vortex formation and the configuration of the discharge line.

PLANT SYSTBlS BASES 3 4.7. 1.6 STANDBY STEAN GENERATOR FEEDWATER SYST The purpose of this specification and the supporting surveillance requirements is to assure operability of the non-safety grade Standby Steam Generator Feedwater System.

The Standby Steam Generator Feedwater System consists of comaercial grade components designed and constructed to industry and FPL standards of this class of equipment located in the outdoor plant environment typical of FPL facilities system wide.

The system is expected to perform with high reliability, i.e., comparable to that typically achieved with this class of equipment.

FPL intends to maintain the system in good operating condition with regard to appearance, structures,

supports, component maintenance, calibrations, etc.

The function of the Standby Steam Generator Feedwater System for OPERABILITY determinations is that it can be used as a backup to the Auxiliary Feedwater (AFW) System in the event the AFM System does not function properly.

The system would be manually started, aligned and controlled by the operator when needed.

The A pump is electric-driven and is powered from the non-safety related C bus.

In the event of a coincident loss of offsite power, the B pump.is diesel driven and can be started and operated independent of the availability of on-site or offsite power.

Q5; ooo A supply of~en-869 gallons from the D

ineralized Mater Storage Tank for the Standby Steam Generator Feedwater Pumps i sufficient water to remove decay heat from the reactor for six (6) hours for a i

le unit or two (2) hours for two units. This was the basis used for requirin 69-,099 gallons of water in the non-safety grade Demineralized Mater Storage Tank an s judged to provide sufficient time fol restoring the AFM System or establishin make-up to the Demineralized Water Storage Tank.

The Standby Steam Generator Feedwater Pumps are not esigned to NRC require-ments applicable to Auxiliary Feedwater Systems and are not required to satisfy design basis events requirements.

These pumps may be out of service for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before initiating formal notification because of the extremely low probability of a demand for their operation.

The guidelines for NRC notification in case of both pumps being out of service for longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are provided in applicable plant procedures, as a

voluntary 4-hour notification.

Adequate demineralized water for the Standby Steam Generator Feedwater system will be verified once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The Demineralized Mater Storage Tank provides a source of water to several systems and therefore, requires daily verification.

The Standby Steam Generator Feedwater Pumps will be verified OPERABLE monthly on a

STAGGERED TEST BASIS by starting and operating them in the recirculation mode.

Also, during each unit's refueling outage, each Standby Steam Generator Feedwater Pump will be started and aligned to provide flow to the nuclear unit's steam generators.

TURKEY POINT UNITS 3 5 4 B 3/4 7-4 ANENDNENT NOS. 164 AND 158

INSERT TO TS BASES 3 4.7.1.6 The minimum indicated volume (135,000 gallons) consists of an allowance for level indication instrument uncertainties (approximately 15,000 gallons); for water deemed unusable because of tank discharge line location and vortex formation (approximately 50,300 gallons);

and the minimum usable volume (65,000 gallons).

The minimum indicated volume corresponds to a water level of 8.5 feet in the Demineralized Water Storage Tank.

CORE OPERATING LIMITS REPORT FOR FLORIDA POWER AND LIGHT COMPANY TURKEY POZNT UNITS 3 AND 4

L-95-245 Attachment 3

CORE OPERATING LIMITS REPORT 1. 0 INTRODUCTION The Core Operating Limits Report (COLR) for Turkey Point Unit 3 Cycle 15 has been prepared in accordance with the requirements of Technical Specifications 6.9.1.7.

The Technical Specifications affected by this report are:

3.2.1 3.1.3.6 3/4.2.2 3/4.2.3 Axial Flux Difference (AFD)

Control Rod Insertion Limits Heat Flux Hot Channel Factor F()(Z)

Nuclear Enthalpy Rise Hot Channel Factor 2.0 OPERATING LIMITS The AFD, F()(Z), FzH, K(Z) and Rod Bank Insertion Limits have been developed using the NRC. approved methodology specified in Technical Specification 6.9.1.7.

These limits are provided in the following subsections:

2.1 Axial Flux Difference (TS 3.2.1)

The Axial Flux Difference (AFD) limits are provided in Figure 1.

2.2 Control Rod Insertion Limits (TS 3.1.3.6)

The control rod banks shall be limited in physical insertion as shown in Figure 2.

2.3 Heat Flux Hot Channel Factor F~(Z)

(TS 3/4.2.2)

L o

[F()]

2.32 o K(Z) is provided in Figure 3.

2.4 Nuclear Enthalpy Rise Hot Channel Factor RTP 1.62 aH o

PFzH

0. 3

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FZGURE 1"

AXZAL FLUX DZFFERENCE AS A FUNCTZON OF RATED THERMAL POWER (TURKEY POZNT UNZT 3 CYCLE 15)

(-10,100)

(+v,100) 80 60 40 0

UNAC EPTAB OP RATION (m,so)

(+as,so) 40

-20

-10 0

10 Anal F}ux08lerence (%)

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FIGURE 2 ROD BANK ZNSERTZON LIMITS VS.

THERMAL POWER ARO = 228 STEPS WITHDRAWN, OVERZAP = 200 STEPS (TURKEY POZNT UNIT 3 CYCLE 15) 216 168

~

120 108 96 M

5 n

CP 60 BANK BAN 100 71 12 0

10

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~FIG RE 3 K(Z) NORMALIZED Fg(R)

AS A FUNCTION OF CORE HEIGHT (TURKEY POINT UNIT 3 CYCLE 15) 1.0

(+1.000

(+1.000)

(1 0.02')

0.9 0.8 0.7 0.6 0$

10 11

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CORE OPERATING LIMITS REPORT i

1.0 INTRODUCTION

The Core Operating Limits Report (COLR) for Turkey Point Unit 4 Cycle 15 has been prepared in accordance with the requirements of Technical Specifications 6.9.1.7.

The Technical Specifications affected by this report are:

3.2.1 3.1.3.6 3/4.2.2 3/4.2.3 Axial Flux Difference Control Rod Insertion Heat Flux Hot Channel Nuclear Enthalpy Rise (AFD)

Limits Factor F() (Z)

Hot Channel Factor

2. 0 OPERATING LIMITS The AFD, F()(Z), F~H, K(Z) and Rod Bank Insertion Limits have been developed using the NRC approved methodology specified in Technical Specification 6.9.1.7.

These limits are provided in the following subsections:

2.1 Axial Flux Difference (TS 3.2.1)

The Axial Flux Difference (AFD) limits are provided in Figure 1.

2.2 Control Rod Insertion Limits (TS 3.1.3.6)

The control rod banks shall be limited in physical insertion as shown in Figure 2.

2.3 Heat. Flux Hot Channel Factor F(Z)

(TS 3/4.2.2)

L o

[F()]

2.32 o K(Z) is provided in Figure 3.

2.4 Nuclear Enthalpy Rise Hot Channel Factor RTP aH o

PF~H

0. 3

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FIGURE 1 AXIALFLUX DIFFERENCE AS A FUNCTION OF RATED THERMAL POWER (TURKEY POINT UNIT 4 CYCLE 15)

(+F,t

(+$

0}

40 M

M 40

-10 0

10 fidel Rux DNIecence Xl 40

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FIGURE 2 ROD BANK INSERTION LIMITS VS.

THERMAL POWER ARO = 228 STEPS WITHDRAWN, OVERLAP = 100 STEPS (TURKEY POZNT UNIT 4 CYCLE 15) 21i (lava) 0 0

0 S

4 S

Q 70

- N 1

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FIGURE 3 K(Z) NORMAIsIZED F (Z) AS A PUNCTION OP CORE HEIGHT (TURKEY OINT UNIT 4 CYCLE 15)

(eve) 104LQ R

~e l

as

ENCLOSURE 1 WCAP-1427 6 FLORXDA POWER 4 LZGHT COMPANY TURKEY POZNT UNXTS 3 AND 4 UPRATXNG LZCENSZNG REPORT

ENCLOSURE 2 PROPRIETARY WCAP 13719, REVZSZON 2 WESTINGHOUSE REVISED THERMAL DESIGN PROCEDURE ZNSTRUMENTS UNCERTAINTY FOR TUREEY POINT UNZTS 3 AND 4 NON-PROPRIETARY WCAP 13718, REV1SZON 2 WESTINGHOUSE REVISED THERMAL DESIGN PROCEDURE INSTRUMENTS UNCERTAINTY FOR TURKEY POINT UNITS 3 AND 4

ENCLOSURE 3 WESTINGHOUSE AUTHORIZATION LETTER CAW-95-890 PROPRIETARY INFORMATION NOTICE COPYRZGHT NOTICE AFFIDAVIT