ML17353A905

Amends 191 & 185 to Licenses DPR-31 & DPR-41,respectively. Amends Increase Authorized Rated Thermal Power from 2200 Mwt to 2300 Mwt & Approves Changes to TS to Implement Uprated Power Operation

ML17353A905
Person / Time
Site:	Turkey Point
Issue date:	09/26/1996
From:	Russell W NRC (Affiliation Not Assigned)
To:
Shared Package
ML17353A906	List: ML17353A905 ML17353A907
References
NUDOCS 9610030107
Download: ML17353A905 (69)

Text

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UNITED STATES NUCLEAR REGULATORYCOMMtSSlON WASHlNQTON, D.C. 20556ESOt F

OR DA POW R AND LIGHT COMPANY DOC ET NO. 50-250 TURK POINT PLANT UNIT NO.

3 AMENOM NT TO F CILITY OP RATING LICENSE Amendment No. 191 License No.

OPR-31 The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Florida Power and Light Company (the licensee) dated

Oecember, 18,

1995, as supplemented by letters dated May 3, June ll, July 1, July 3, and August 22,

1996, complies with the standards and requirements of the Atomic Energy Act of

1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

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

96i0030i07 960926 PDR ADQCK 05000250 P

PDB

2. Accordingly, Facility Operating License No.

DPR-31 items c.

and 3.A are hereby amended to read as follows:

c.

There is reasonable assurance (i) that the facility can be operated at steady state power levels up to 2300 megawatts thermal in accordance with this license without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations of the Commission; 3.A aximum Power Level 3.

The applicant is authorized to operate the facility at reactor core power levels not in excess of 2300 megawatts (thermal).

Accordingly, the license is amended by changes to the Technical

. Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.

DPR-31 is hereby amended to read as follows:

(B)

Tec c

1 S ecif cations and Environme tal Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 191, 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.

This license amendment's effective as of i.ts date of issuance and shall be implemented within 120 days.

FOR THE NUCLEAR REGULATORY COMMISSION William T. Russell, Director Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 26, 1996

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UNITED STATES

- NUCLEAR REGULATORYCOMMISSION WASHINGTON, D.C. 2055&4%1 FLORIDA POWER AND LIGHT COMPANY DOCK T NO. 50-25 TURKEY POINT P ANT UNIT NO.

4 AMENDN NT TO FACILITY OPERATING LICENSE Amendment No.185 License No.

DPR-41 1.

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

A.

The application for amendment by Florida Power and Light Company (the licensee) dated December 18,

1995, as supplemented by letters dated May 3, June ll, July 1, July 3, and August 22,

1996, complies with the standards and requirements of the Atomic Energy Act of

1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

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

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3. Accordingly, Facility Operating License No.

DPR-41 condition 3.A is hereby amended to read as fo'llows:

3.A owe vel 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 operate the facility at reactor core power levels not in excess of.2300 megawatts (thermal).

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No.

DPR-41 is hereby amended to read as follows:

4.

(8) c S ecif'cat ons and nvironmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.185

, are hereby incorporated in the license.

The Environmental Protection Plan contained in Appendix 8 is hereby incorporated into the'icense.

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

This license amendment is effective as of its date of issuance and shall be implemented within 120 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

SePtember 26, 1996

~ ~

William T. Russell, Director Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT M NOME T NO.

M N M

NO.

FACI ITY OPERA ING LICENSE NO.

OPR-31 FACI ITY OP RATING LICENSE NO.

OPR-41 OC T NOS.

50-250 AND 50-251 Revise Appendix A as follows:

Remove a es 1-5 2-2 2-4 2-5 2-7 2-8 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 6-21 8 2-1 8 2-2 8 2-7 8 3/4 2-1 8 3/4 2-4 8 3/4 2-8 8 3/4 4-2 8 3/4 4-8 8 3/4 4-9 8 3/4 6-3 8 3/4 6-4 8 3/4 7-2 8 3/4 7-3 8 3/4 7-4 8 3/4 7-5 8 3/4 7-6 8 3/4 7-7 Insert a es 1-5 2-2 2-4 2-5 2-7 2-8 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 6-21 8 2-1 8 2-2 8 2-7 8 3/4 2-'1 8 3/4 2-4 8 3/4 2-8 8 3/4 4-2 8 3/4 4-8 8 3/4 4-9 8 3/4 6-3 8 3/4 6-4 8 3/4 7-2 8 3/4 7-3 8 3/4. 7-4 8 3/4 7-5 8 3/4 7-6 8 3/4 7-7 8 3/4 7-8

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 calibrated

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.

1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2300 MWt.

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

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.

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.

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

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.

1.30 A STAGGERED TEST BASIS shall consist of:

a

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A test schedule for n systems, subsystems,

trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

TURKEY POINT UNITS 3 8 4

1-5 AMENDMENT NOS-i91 AND 185

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ALLOWABLE YhlJlE 1.

Manual Reactor Trip 2.

Power

Range, Neutron Flux a.

High Setpoint b.

Low Setpoint 3.

Intermediate

Range, Neutron Flux 4.

Source

Range, Neutron Flux N.A

<112.0% of RTP**

s28.0% of RTP**

~31.0% of RTP0' s1.4 X 10 cps N.A.

sl09% of RTP**

s25% of RTP**

s25% of RTP**

~10 cps 5

5.

Overtemperature aT See Note 2

See Note 1

6.

Overpower aT 7.

Pressurizer Pressure-Low 8.

Pressurizer Pressure-High 9.

Pressurizer Water Level-High 10.

Reactor Coolant Flow-Low ll.

Steam Generator Water Level Low-Low

" Loop design flow = 85.000 gpm

• • RTP,= Rated Thermal Power See Note 4

>>&17 psig s2403 psig s92.2% of instrument span

>88.8% of loop design flow*

~8.15% of narrow range instrument span See Note 3

>>835 psig 52385 psig

~92% of instrument span

>90% of loop design flow*

>>0% of narrow range instr'ument span

4 7Cm(

ALLOWABLE

. 54hLUE 12.

Steam/Feedwater Flow Nismatch Coincident With Steam Generator Water Level-Low Feed Flow s23.9X below rated Steam Flow

~8.15X of narrow range instrument span Feed Flow ~20X below rated Steam Flow

~IOX of narrow range instrument span 0

13.

Undervol tage

- 4.16 kV Busses A and B

>69X bus voltage

>70X bus voltage 14.

Underfrequency

- Trip of Reactor Coolant Pump Breaker(s)

Open

~55.9 Hz

>56.1 Hz 15.

Turbine Trip C) tel 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

~42 psig Fully Closed***

N.A.

>6. Ox 10 amp s

>45 psig Fully Closed***

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N.A.

Nominal 1x10 jmp

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

I

NOTE 1:

OVERTEMPERATURE aT nT 1+TIS

(

) caT K1-K2(1+T4S) [T(

)

- T']+K(P-P') - f (gi) 1+ T3 1+ T6 1+ T2S (1 + T5S)

Where:

aT

=

Measured dT by RTO Instrumentation 1+rS 1+ T2S 1+ T3S Lead/Lag compensator on measured aT; T1=0s, T2 =Os Lag compensator on measured aT;- T3 = Os Indicated nT at RATED THERMAL POWER 1.24; K2 1 + T4S 1 + T5S T4, T5 0.017/

F; The function generated by the lead-lag compensator for T dynamic compensation; avg Time constants utilized in the lead-lag compensator for T 3s'vg' T5 =

S; Average temperature, F;

1 + T6S K3 Lag compensator on measured 6 = Os 577.2 F (Nominal T

at R TEO THERMAL POWFR);

0.001/psi g; Pressurizer

pressure, psig;

NOTE 1:

(Continued) pt 2235 psig (Nominal RCS operating pressure);

Laplace transform operator, s

and f1(zl) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

( 1)

For qt qb between 50X and

+ 2X, f (nl) = 0, where q

and q

are percent, RATED THERMAL POWER in the top and bottom halves of the core respectively, and q

+ q is total THERMAL POWER in percent of RATED THERMAL POWER;

( 2)

For each percent that the magnitude of q q

exceeds

- 50X. the aT Trip Setpoint automatically reduced by O.OX of its value at RATED THERMAL POWER; and I

(3)

For each percent that the magnitude of q q

exceeds

+ 2X. the aT Trip Setpoint shall automatically reduced by 2. 19X of its value at RATED THERMAL POWER.

NOTE 2:

The channels maximum trip setpoint shall not exceed its computed setpoint by more than O.84X of instrument span.

NOTE 3:

(Continued)

K6 T

N 0.0016/

F for T > T" 0 for T z T",

As defined in Note 1,

~ 577.2 F (Nominal T

at RATED THERMAL POWER)

As defined in Note 1, and f

(ni )

=

0 for all zI 2

NOTE 4:

The channel's maximum trip setpoint shall not exceed its computed trip setpoint by more than 0.96$

of instrument span.

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR -

FQ (Z) 3.2.2 FQ (Z) shall be limited by the following relationships:

L M

F LX FQ (Z) ( i Qj LK(Z)j'for P ) 0.5 P

LX FQ (Z)

"~~

LK(Z)] for P ( 0.5

0.5 where

[FQ3

=

FQ limit at RATED THERMAL POWER as specified L

in the CORE OPERATING LIMITS REPORT Rated Thermal Power

[FQ3

= The Measured Value and M

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

MODE 1

hCIlQH:

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

M a.

Reduce THERMAL POWER at least 1X for each 1X FQ (Z) exceeds FQ (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 proceed 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 Delta-T Trip Setpoints (value of K4) have been reduced at least 1X for each 1X FQ (Z) exceeds the FQ (Z); and M

L

'I 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 FQ (Z) is demonstrated through incore mapping to be within its M

limit.

'TURKEY POINT - UNITS 3 8

4 3/4 2-4 AMENDMENT NOS.i9iANDi85

3.2.3 F>H shall be limited by the following relationship:

N F~H ( F~H [1.0 + PF~H (1-P) 3, N

RTP Where:

FzH

=

F~H limit at RATED THERMAL POWER as specified RTP in the CORE OPERATING LIMITS REPORT PF~H = Power Factor Multiplier for F~H as specified in the CORE OPERATING LIMITS REPORT P

=

RATED THERMAL POWER EEJlQH:

MODE 1.

With F~H exceeding its limit:

N a.

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

1.

Restore F~H to within the above limit, or N

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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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 N

through incore flux mapping that F~H 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 />.

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>H is demonstrated, through N

incore flux mapping, to be within the limit of acceptable operation prior to exceeding the following THERMAL POWER levels:

l.

A nominal 50K of RATED THERMAL POWER, 2.

A nominal 75K 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 95K of RATED THERMAL POWER.

TURKEY POINT - UNITS 3 8 4

3/4 2-11 AMENDMENT NOS. "9iANDi85

3.2.5 The following DNB-related parameters shall be maintained within the following limits:

a.

Reactor Coolant System Tavg

( 581.2'F b.

Pressurizer Pressure

) 2200 psig*, and c.

Reactor Coolant System Flow

> 264,000 gpm MODE 1.

ECIlQH:

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 />.

4.2.5. 1 Reactor Coolant System T

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 The RCS flow rate indicators shall be subjected to a

CHANNEL CALIBRATION at least once per 18 months.

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

• Limit not applicable during either a

THERMAL POWER ramp in excess of 5X 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.i9i AND i85

N NEER AF TY AT R

A T AT N

Y T INTRMN T

N P

P N

FUNCTIONAL UNIT A

WA E

A T

P TP T

1.

Safety Injection (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 Flow--High N.A.

N.A

<4.5 psig

>1712 psig

< 114 psig

<A function defined as follows:

A aP corresponding to 44K steam flow at OX load increasing linearly from 20K load to a value corresponding to 116.5X steam flow at full load N.A.

N.A

<4.0 psig

>1730 psig

<100 psi

<A function defined as follows: A aP corresponding to 40K steam flow at OX load increasing linearly from 20K load to a value corresponding to 114K steam flow at full load

(Continued) 4 Steam Line Isolation (Continued) b.

Automatic Actuation Logic'nd Actuation Relays NBA.

N.A.

C.

Containment Pressure--High-High Coincident with:

Containment Pressure--High

<22.6 psig

< 4.5 psig

<20.0 psig

<4.0 psig d.

Steam Line Flow--High

<A function defined as follows:

A aP corresponding to 44K steam flow at OX load increasing linearly from 20K load to a value corresponding to 116.5X steam flow at full load.

function defined as follows:

A zp corresponding to 40'team flow at load increasing linearly from 20X load to a value corresponding to 114K steam flow at full load.

m CI m 5.

Coincident with:

Steam i.ine Pressure--Low or Tavg Low Feedwater Isolation

>588 psig

>542.5 F

>614 psig

>543'F C)

'EO 2

ED CO Ul a.

Automatic Actuation Logic and Actuation Relays b.

Safety Injection N.A.

See item 1. for all Safety Injection Allowable Values.

N.A.

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

5.

Feedwater C.

Isolation (Continued)

Steam Generator Water Level High-High

<81.9X of nar'row range instrument span

<80K of narrow range instrument span 6 ~

Auxiliary a

~

Feedwater (3)

Automatic Actuation Logic and Actuation Relays N.A.

N.A.

b.

Steam Generator Water Level--Low-Low

>8.15K of narrow range instrument span.

>10K of narrow range instrument span.

c.

Safety Injection d.

Bus Stripping see Item.l. for all Safety Injection

'llowable Values.

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

See Item 1.

above for all Safety I'njection Trip Setpoints.

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

e.

Trip of All Main Feedwater Pump Breakers N.A.

N.A.

7.

Loss of Power a.

4. 16 kV Busses A and B

(Loss of Voltage)

N.A.

N.A.

l, n

3.4.2.

1 A minimum of one pressurizer Code safety valve shall be OPERABLE* with a lift setting of 2485 psig

+ 2X, -3X.** ***

MODES 4 and 5.

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

4.4.2. 1 No additional requirements other than those required by Specification 4.0.5.

• While in MODE 5, an equivalent size vent pathway may be used provided that the vent pathway is not isolated or sealed.

• • The lift setti,ng pressure shall. correspond to,ambient conditions:of the valve at nominal operating temperature and pressure.

• • • Allvalves tested must have "as left" lift setpoints that are within

+

1X of the lift setting value.

TURKEY POINT - UNITS 3

& 4 3/4 4-7 AMENDMENT NOS. 191AND185

3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig

+ 2X, -3X.* **

MODES 1, 2 and 3.

KIlQH:

With one pressurizer Code safety valve inoperable, either restore the inoperable 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 />.

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

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

• • All valves tested must have "as left" lift setpoints that are within

+

1X of the lift setting value.

TURKEY POINT - UNITS 3 8 4

3/4 4-8 AMENDMENT NOS. 191AND185

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4.5.2 Each ECCS component and flow path shall be demonstrated OPERABLE:

a.

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

864A and B

862A and 8

863A and B

866A and B

HCV-758*

Supply from RWST to ECCS RWST Supply to RHR pumps RHR Recirculation H.H.S.I. to Hot Legs RHR HX Outlet Open Open Closed Closed Open To permit temporary operation of these valves for surveillance or maintenance

purposes, power may 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 pump casings and accessible discharge

piping, 2)

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, and 3)

Verifying that each RHR Pump 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 least once per 92 days by:

1)

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

SI pump

> 1083 psid at a metered flowrate ) 300 gpm (normal alignment or Unit 4 SI pumps aligned to Unit 3 RWST), or

) 1113 psid at a metered flowrate ) 280 gpm (Unit 3 SI pumps aligned to. Unit 4 RWST).

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

TURKEY POINT UNITS 3

& 4 3/4 5-5 AMENDMENT NOS.191 AND 185

I'

3.6.2.2 Three emergency containment cooling units shall be OPERABLE.

MODES 1, 2, 3, and 4.

SGIlQH:

a.

With one of the above required emergency containment cooling units inoperable restore the inoperable cooling unit to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in 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 />.

b.

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 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 />.

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 COLO SHUTDOWN within the fol)owing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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 two emergency containment cooling units start automatically on a safety injection (Sl) test signal, and 2)

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

TURKEY POINT

- UNITS 3 8

4 3/4 6-14 AMENDMENT NOS.i 91 AND 185

MAXIMUM NUMBER OF INOPERABLE SAFETY VALVES ON ANY MAXIMUM ALLOWABLE POWER LEVEL 53 33 14 ORIFICE SIZE 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

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

• • All valves tested must have "as left" lift setpoints that are within +IX I

of the lift setting value -listed in Table 3.7-2.

I TURKEY POINT

- UNITS 3 8

4 3/4 7-2 AMENOMENT NOS.191 AND185

0

3.7. 1.3 The condensate storage tanks (CST) system shall be OPERABLE with:

A minimum indicated water volume of 210,000 gallons in either or both condensate storage tanks.

A minimum indicated water volume of 420,000 gallons.

MODES 1, 2 and 3.

ECIlQH:

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 />.

With the CST system inoperable due to indicating less than 420,000 gallons, but greater than or equal to 210,000 gallons indicated, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> restore the inoperable CST system to OPERABLE status or place 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 />.

2)

With the CST system inoperable with less than 210,000 gallons indicated, within I

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 be 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'he 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.i 91 AND185

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

TURKEY POINT UNITS 3 8I 4 3/4 7-7 AMENDMENT NOS.191 AND 185

il

3.7. 1.6 Two Standby Steam Generator Feedwater Pumps shall be OPERABLE* and at least 135,000 gallons of water (indicated volume), shall be in the Demin'eralized Water Storage Tank.**

MODES 1, 2 and 3

hGIlQH'-

a.

With 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, resto're 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:

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, 2.

Submit a

SPECIAL REPORT per 3.7. 1.6d.

c.

With less than 135,000 gallons of water indicated 'in the Demineralized Water Storage Tank restore the available volume to at least 135,000 gallons indicated 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.

d.

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.

4.7. 1.6. 1 The Demineralized Water 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 feedwater pumps are OPERABLE by'esting 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 8 4 3/4 7-11 AMENDMENT NOS f91 ANDi85

1)

Verifying that the air cleanup system satisfies the in-place pene-tration and bypass leakage testing acceptance criteria of greater than or equal to 99X DOP and halogenated hydrocarbon removal at a system f1 ow ra te of 1000 cfm +lOX.

2)

Verifying, within 31 days after removal, that a laboratory 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 99K or the charcoal be replaced with charcoal that meets or exceeds the criteria of position C.6.a. of Regulatory Guide 1.52 (Revision 2),

and 3)

Verifying by a visual inspection the absence of foreign materials and gasket deterioration.

d.

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 +lOX; e.

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.

'I TURKEY POINT

- UNITS 3 8

4 3/4 7-17 AMENDMENT NOS.191 AND i85

4 F

6.9. 1.6 The W(Z) function(s) for Base-Load Operation corresponding to a +2K band about the target flux difference and/or a +3K band about the target flux difference, the Load-Follow function Fz(Z) and the augmented surveillance

'turnon power fraction, P~, shall be provided to the U.S. Nuclear Regulatory Commission, whenever P> is (1.0.

In the event, the option of Baseload Operation (as defined in Section 4.2.2.3) will not be exercised, the submission of the W(Z) function is not required.

Should these values (i.e., W(Z), Fz(Z) and P~)

change requiring a

new submittal or an amended submittal to the Peaking 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.

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.

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:

2.

3.

4.

5.

Axial Flux Difference for Specification 3.2. 1.

Control Rod Insertion Limits for Specification

3. 1.3.6.

Heat Flux Hot Channel Factor

- FO(2) for Specification 3/4.2.2.

All Rods Out position for Spe'cification

3. 1.3.2.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3 The analytical methods used to determine the AFD limits shall be those previously reviewed and approved by the NRC in:

1.

WCAP-10216-P-A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL Fg SURVEILLANCE TECHNICAL SPECIFICATION," June 1983.

2.

WCAP-8385, "POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES TOPICAL REPORT," September 1974.

The analytical methods used to determine FO(Z),

F>H and the K(Z) curve shall be I

those previously reviewed and approved by the NRC in:

1.

WCAP-9220-P-A, Rev.

1, "Westinghouse ECCS Evaluation Model 1981 Version," February 1982.

2.

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

1, "BART A-1:

A Computer Code for the Best Estimate Analysis of Reflood Transients

- Special Report:

Thimble Modeling H ECCS Evaluation Model."

3.

WCAP-10054-P-A, (proprietary),

"Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code",

August 1985.

TURKEY POINT UNITS 3 8 4

6-20 AMENDMENT NOS. i91 AND i85

4.

WCAP-10054-P, Addendum 2, Revision 1 (proprietary),

"Addendum to the

'estinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:

Safety Injection in the Broken Loop and Improved Condensation Model",

October 1995.*

5.

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.

6.

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

The analytical methods used to determine Rod Bank Insertion Limits and the All Rods Out position shall be those previously reviewed and approved by the NRC in:

1.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

July 1985.

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

1.

Florida Power

& Light Company Topical Report NF-TR-95-01, "Nuclear Physics Methodology for Reload Design of Turkey Point

& St.

Lucie Nuclear Plants".

Topical Report NF-TR-95-01 was approved by the NRC for. use by Florida Power Light Company in:

1.

Safety Evaluation by the Office of Nuclear Reactor Regulations Related to Amendment No.

174 to Facility Operating License DPR-31 and Amendment No.

168 to Facility Operating License DPR-41, Florida Power

& Light Company Turkey Point Units 3 and 4, Docket Nos.

50-250 and 50-251.

The AFD, FO(Z),

F<H, K(2), and Rod Bank Insertion Limits shall be determined such that all applicable limits of the safety analyses are met.

The CORE OPERATING LIMITS 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 Resident Inspector, unless otherwise approved by the Commission.

• This reference is only to be used subsequent to NRC approval.

TURKEY POINT UNITS 3

& 4 6-21 AMENDMENT NOS.

191 AND 185

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 boil1ng (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter dur1ng operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB.

This relationship has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distr1butions.

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

The DNB 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 DNBR limit of the DNB correlation being used.

The correlation DNBR limit is established based on the entire appl1cable experimental data set such that there is a

95 percent probability with 95 percent confidence that DNB 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 THFRMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR 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, 'FzH, and a

I N

reference cosine with a peak of 1.55 for axial power shape.

An allowance is included for an increase in F~H at reduced power based on the N

expression:

F~H <

F~H [I+ PF~H (1-P) 3 N

RTP I

l Where P is the fraction of RATED THERMAL POWER.

Fz,H

=

Fz,H limit at RATED THERMAL POWER as specified in the CORE RTP OPERATING LIMITS REPORT.

PFzH = Power Factor multiplier for F>H as specified in the CORE OPERATING LIMITS REPORT.

TURKEY POINT - UNITS 3 8 4

B 2-1 AMENDMENT NOS. 191 AND185

(Continued)

These limiting heat flux conditions are higher 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 (al) function of the Overtemperature trip:

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature aT trips will reduce the setpoints to provide protection consistent with core Safety Limits.

Fuel rod bowing reduces the values of DNB ratio'(DNBR).

The penalties are calculated pursuant to "Fuel Rod Bow Evaluation,"

WCAP-8691-P-A Revision 1 (Proprietary) and WCAP-8692 Revision 1 (Non-Proprietary).

The restrictions of the Core Thermal Hydraulic Safety Limits assure that an

'mount of DNBR margin greater than or equal to the above penalties is retained to offset the rod bow DNBR penalty.

The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release. of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110K (2735 psig) of design pressure.

The RCS piping, valves and fittings are designed to ANSI B31. 1 which. permits.

a maximum transient pressure of 120K of design pressure of 2485 psig.

" The Safety Limit of 2735 psig is therefore more conservative than the ANSI B31. 1 design criteria and consistent with associated ASME Code requirements.

The entire RCS is hydrotested at 125K (3107 psig) of design

pressure, to demonstrate integrity prior to initial operation.

TURKEY POINT

- UNITS 3 8

4 B 2-2 AMENDMENT NOS.1 91 AND185

(Continued) power the Undervoltage Bus trips are automatically blocked by P-7 (a power level of approximately 10K of RATED THERMAL POWER with a turbine first stage pressure at approximately 10K of full power equivalent);

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

A Turbine trip initiates a Reactor trip.

On decreasing

power, the Reactor Trip from the Turbine trip is automatically blocked by P-7 (a power level of approximately 10X of RATED THERMAL POWER with a turbine first stage pressure at approximately 10K of full power equivalent);

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

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

The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against DNB.

The open/close position trips assure a

reactor trip signal is generated before the low flow trip setpoint is reached.

Their I

functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Protection System.

Above P-7 (a power level of approximately 10K of RATED THERMAL POWER or a turbine first stage pressure at approximately 10K 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'n 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.

191 AND185

T T

N 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 2200'F is not exceeded.

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

FQ(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; N

F~H 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 FXY(Z)

Radial.Peaking Factor, is defined as the r atio of peak power density to average power density in the horizontal plane at core elevation 2.

/4 AXIA F

X FF RN The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the FQ(Z) limit defined in the CORE OPERATING LIMITS REPORT times the normalized axial

.peaking factor is not exceeded during either normal 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

'HERMAL 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 burnup considerations.

TURKEY POINT

- UNITS 3

& 4 8 3/4 2-1 AMENDMENT NOS. 191 AN0185

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 2200 F

ECCS acceptance criteria limit.

The LOCA peak fuel clad temperature limit may be sensitive to the number of steam generator tubes plugged.

I FO(Z),

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.

FaH N

, 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.

This 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

+ 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 Specifi*cations 3. 1.3.5 and 3. 1.3.6 are maintained; and d.

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

When an FO 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 and PRB.

F>H will be maintained within its limits provided Conditions a.

through d.

N above are maintained.

In the specified limit of F>H, there is an 8 percent allowance for N

uncertaintie~

which ~cans that normal operation of the core is expected to

'esult in FzH c

F>H /1.08, where FzH is the FzH limit at RATED THERMAL POWER RT RTP N

(RTP) specified in the CORE OPERATING LIMITS REPORT.

The logic behind the larger uncertainty in this TURKEY POINT - UNITS 3 8

4 B 3/4 2-4 AMENDMENT NOS.191 AND 185

The QUADRANT POWER TILT RATIO limit assures that the radial power distribution 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 DNB 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 associated 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 correction of a dropped or misaligned control rod.

In the event such action action does not correct the tilt, the margin for uncertainty on FQ(Z) is reinstated by reducing the maximum allowed power by 3X for each percent of tilt in excess of l.

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-B, E-5, E-11. H-3. H-13, L-5. L-11, N-B.

The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR above the applicable design limits throughout each analyzed transient.

The indicated Tavg value of 581.2'F and the indicated pressurizer pressure value of 2200 I

psig correspond to analytical limits of 583;2'F and 2175 psig respectively, with I

allowance for measurement uncertainty.

The measured RCS flow value of 264,000 gpm corresponds to an analytical limit of I

255,000 gpm which is assumed to have a 3.5X cal'orimetric measurement uncertainty.

I The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18-month periodic measurement of the RCS total flow rate is adequate to 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 feedwater 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 degradation has not occurred.

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.

TURKEY POINT - UNITS 3 8 4

B 3/4 2-8 AMENDMENT NOS. 191 AND185

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 overpressure 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.

During operation, all pressur1zer 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 Setpo1nt 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

'pening, 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 wil'l occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

The pressuri-zer code safety valves'ift settings allows a +2K,

-3X setpoint tolerance for OPERABILITY; however, the valves are reset to within +1K during the surveillance to allow for drift.

I I

I 4

4 p

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

hydraulical,ly 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.

TURKEY POINT - UNITS 3 8

4 8 3/4 4-2 AMENDMENT NOS. 191 AND185

(Continued) 1.

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) 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 lines for cooldown rates between those presented may be obtained by interpolation; and b.

Figures'3.4-2 to 3.4-4 define limits to assure prevention of non-ductile failure only.

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

pump heat addition and pressurizer heater

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

2. 'hese limit lines shall be 'calculated periodically 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 is greater than

'20'F, and 5.

System preservice hydrotests and inservice 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, Appendex H,

and in accordance with additional reactor vessel'requirements.

The properties are then evaluated in accordance with Appendix G of the 1983 Edition of Section III of the ASME Boiler and Pressure Vessel Code and the additional requirements of 10 CFR 50, Appendix G and the calculation methods described in Westinghouse Report GTSD-A-1. 12, "Procedure for Developing Heatup and Cooldown Curves."

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 19 effective full power years (EFPY) of service life.

The 19 EFPY I

service life period is chosen such that the limiting RTNDT, at the 1/4T location in TURKEY POINT

- UNITS 3

& 4 8 3/4 4-8 AMENDMENT NOS.191 AND i85

(Continued) the core region is greater than the RTNOT, of the limiting unirradiated material.

The selection of such a limiting RTNOT 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 3.4-4 are composite curves prepared by determining the most conservative case with either the inside or outside wall controlling, for 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 limiting value of predicted adjusted reference temperature at the end of the 'applicable service period (19 EFPY).

The reactor vessel materials have been tested to determine their initial RTNOT,. the results of these tests are shown in Tables B 3/4.4-1 and 8 3/4.4-2.

Reactor operation and resultant fast neutron (E greater than 1

MeV) irradiation can cause an increase in the RTNOT.

Therefore, an adjusted reference temperature, based upon the fluence and chemistry factors of the material has been predictedgusing Regulatory Guide 1.99, Revision 2, dated May 1988, "Radiation Embr/ttlement of Reactor Vessel Materials."

The heatup and cooldown limit curves of Figures 3.4-2, 3.4-3, and 3.4-4 include predicted adjustments for this shift in RTNOT at the end of the applicable service period..

The actual shifts in RTNOT, of the vessel materials will be established periodically during operation by removing and evaluating, in accordance with the version of the ASTM E185 standard required by 10 CFR Appendix H, reactor vessel material irradiation 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 Weld) 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" results from Unit 3 (WCAP 8631) and Unit 4 (SWRI 02-4221) and the capsule "V" results from Unit 3 (SWRI 06-

.8576 were used with the methodology in Regulatory Guide 1.99, Revision 2, to provide TURKEY POINT - UNITS 3

& 4 B 3/4 4-9 AMENDMENT NOS.1 91 ANP 185

V

~

t4 1'h 4

4

(.Con,ti.need

):.

resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop.

The 0.60 L, leakage lim1t of Specification 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 penetrat1ons subject to Type B and C tests.

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 contai.nment 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.

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 thermal

analysis, 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 an "B" train SI signal).

To support post-LOCA long-term containment pressure/temperature analyses',

a maximum of two ECCs are 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.

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 add1tional 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 configuration and fl.ow.

An 18-month surveilance interval is acceptable based on the use of water from the CCW system, which results in a low risk of heat exchanger tube fouling.

TURKEY POINT - UNITS 3 8I 4 B 3/4 6-3 AMENDMENT NOS.191 ANP 185

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 BOUNDARY 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 r'equ'ired 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.

The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to.the containment atmosphere or pressurization of the containment.

Containment isolation within the time limits specified in the In-Service Testing Program is consistent with the assumed isolation times of those valves with specific isolation times in the LOCA analysis.

The OPERABILITY of the Hydrogen Monitors ensures the detection of hydrogen buildup within containment following a LOCA to allow operator action to reduce the hydrogen concentration below its flammable limit.

The OPERABILITY of the Post Accident Containment Vent System ensures the capability for emergency venting of containment following a LOCA to reduce the hydrogen concentration to below its flammable limit.

PACVS systems components are not subject to rapid deterioration, having lifetimes of many years.

even under continuous flow conditions.

Visual inspection and operating tests provide assurance of system reliability and will ensure early detection of conditions which could cause the system to fail or operate improperly.

The performance tests prove that filters have been properly installed, that no deterioration or damage has

occurred, and that all components an" subsystems operate properly.

The tests are performed in accordance with the methodology and intent of ANSI N510-1975 and provide assurance that filter performance has not deteriorated below required specification values due to aging, contamination or other effects'URKEY POINT

- UNITS 3 8 4 B 3/4 6-4 AMENDMENT NOS. i 9) AN/B5

(Continued)

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

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

Table 3.7-2 allows a + 3X setpoint tolerance for OPERABILITY; however, the valves are reset to +

1X during the Surveillance to allow for drift.

The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss-of-offsite power.

Steam can be supplied to the pump turbines from 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 feedwater 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 alternate 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 1 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 SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION statement 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 Specification 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 of ACTION statement 3 are satisfied.

TURKEY POINT - UNITS 3

& 4 B 3/4 7-2 AMENDMENT NOS.i 91 ANDi85

(Continued)

The monthly testing of the auxiliary feedwater pumps will ver1fy their operability.

Proper functioning of the turbine admission valve and the operation of the pumps will demonstrate the integrity of the system.

Verification of correct operation will be made both from instrumentation within the control room and direct visual observation of the pumps.

There are two (2) seismically designed 250,000 gallons condensate storage tanks.

A minimum indicated volume of 210,000 gallons is maintained for each unit in MODES 1, l

2 or 3.

The OPERABILITY of the condensate storage tank with the minimum indicated volume ensures that sufficient water is available to maintain the Reactor Coolant System at HOT STANDBY conditions for approximately 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> or maintain the Reactor Coolant System at HOT STANDBY cond1tions 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'eat Removal System may be placed in operation.

The minimum indicated volume includes an allowance for instrument indication uncertainties and for water deemed unusable because of-vortex formation and the configurat1on of the d1scharge line.

The limit on secondary coolant specific activ1ty is based on a postulated release of secondary coolant equivalent to the contents of three steam generators to the atmosphere due to a net load rejection.

The limiting dose for this case would result from radioactive iodine 1n 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 is then; Dose (Rem)

=

C

• V

• B

• DCF

• X/Q

• 0. 1 Where:

C X/Q secondary coolant dose equivalent 1-131 specific activity 0.2 curies/

m (VCi/cc) or 0. 1 Ci/m3

, each unit 3

equivalent secondary coolant volume released

= 214 m3

-4 breathing rate

= 3.47 x 10 m3/sec.

4 atmospheric dispersion parameter

= 1.54 x 10 sec/m3

0. 1

=

equivalent fraction of activity released DCF

=

dose conversion factor, Rem/Ci The resultant thyroid dose is less than 1.5 Rem.

TURKEY POINT - UNITS 3 8

4 8 3/4 7-3 AMENDMENT NOS. 191 AND185

" 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 is required to:

( I) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in )he event the steam line rupture occurs within containment.

The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Requirements 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.

The purpose of this specification and the supporting surveillance requirements is to assure operability of the non-safety grade Standby Steam Generator Feedwater Sys-tem.

The Standby Steam Generator Feedwater System consists of commercial 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 ia the event the AFW 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 off-site power.

A supply of 65,000 gallons from the Demineralized Water Storage Tank for the Standby Steam Generator Feedwater Pumps is sufficient water to remove decay heat from the reactor for six (6) hours for a single unit or two (2) hours for two units. This was the basis used for requiring 65,000 gallons of water in the non-safety grade I

Demineralized Water Storage Tank and is judged to provide sufficient time for restoring the AFW System or establishing make-up to the Demineralized Water Storage Tank.

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.

The Standby Steam Generator Feedwater Pumps are not designed to NRC requirements 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, TURKEY POINT

- UNITS 3 8 4

8 3/4 7-4 AMEiVDNENT NOS. i 9i ANDiB5

(Continued)

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 Fe'edwater 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 Water 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 recirculat1on mode.

Also, during each unit's refueling outage, each Standby Steam Generator Feedwater Pump will be started and aligned to prov1de flow to the nuclear unit's steam generators.

This surveillance regimen will thus demonstrate operability of the entire flow

path, backup non-safety grade power supply and pump associated with a unit at least each refueling outage.

The pump, motor driver, and normal power supply availability would typically be demonstrated by operation of the pumps in the recirculation mode monthly on a staggered test basis'he diesel engine driver'for the B Standby Steam Generator Feedwater Pump will be verified operable once every 31 days on a staggered test basis performed on the 8

Standby Steam Generator Feedwater Pump.

In addition, an inspection will be performed on the diesel at least once every 18 months in accordance with procedures prepared in conjunction with its manufacture's recommendations for the diesel's class of service.

This inspection will ensure that the diesel driver is maintained in good operating condit1on consistent with FPL's overall objectives for system reliability.

The OPERABILITY of the Component Cooling Water System.ensures that sufficient cooling capacity is available for continued operation o'f safety-related equipment during normal and acc1dent cond1tions.

The redundant cooling capacity of this

system, assuming a single active failure, is consistent with the assumptions used in the sa'fety analyses.

One pump and two heat exchangers provide the heat removal capability for accidents that have been analyzed.

The OPERABILITY of the Intake Cooling Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.

The design and operation of this system, assuming a single active failure, ensures cooling capacity consistent with the assumptions used in the safety analyses.

TURkEY POINT

- UNITS 3

& 4 B 3/4 7-5 AMENDMENT NOS.191 AND 185

The limit on ultimate heat sink temperature in conjunction with the SURVEILANCE REQUIREMENTS of Technical Specification 3/4.7.2 will ensure that sufficient cooling capacity is available either: (I) to provide normal cooldown of the facility, or (2) to mitigate the effects of accident conditions within acceptable limits.

With the implementation of the CCW heat exchanger performance monitoring program, the limiting UHS temperature can be treated as a variable with an absolute upper limit of 100 F without compromising any margin of safety.

Demonstration of actual heat exchanger performance capability supports system operation with postulated canal temperatures greater than 100 F.

Therefore.

an upper Technical Specification limit of 100 F is conservative.

The OPERABILITY of the Control Room Emergency Ventilation System ensures that:

(I) the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

The OPERABILITY of this system in con-junction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rems or less whole body, or its equivalent.

This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50.

System components are not subject to rapid deterioration, having lifetimes of many years, even under continuous flow conditions.

Visual inspection and operating tests provide assurance of system reliability and will ensure early detection of conditions which could cause the system to fail or operate improperly.

The filters performance tests prove that filters have been properly installed, that no deterioration or damage has occurred, and that all components and subsystems operate properly.

The tests are performed in accordance with the methodology and intent of ANSI N510

( 1975) and provide assurance that filter performance has not deteriorated below returned specification values due to aging, contamination, or other effects.

All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to each safety-related system during an earthquake or severe transients Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspections Inspections performed before that interval has elapsed may be used as a

new reference point to determine the next inspection.

However, the results of such early inspections performed before the original required time TURKEY POINT - UNITS 3 8 4 B 3/4 7-6 AMENDMENT NOS.i91 ANDI85

interval has elapsed (nominal time less 25K) may not be used to lengthen the required inspection interval.

Any inspection whose results require a shorter inspection interval will override the previous schedule.

When the cause of the rejection of a snubber is visual inspection is clearly established and remedied for the snubber and for any other snubbers that may be generically susceptible.

and verified operable by inservice functional testing, that snubber may be exempted from being counted as inoperable for the purposes of establishing the next visual inspection interval. Generically susceptible snubbers are those whi'ch are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration.

When a snubber is found inoperable, an evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any Safety Related System or component has been adversely affected by the inoperability of the snubber.

The evaluation shall determine whether or not the snubber mode of failure h'as imparted a significant effect or degradation on the supported component or system.

To provide assurance of snubber functional reliability, a representative sample of the installed snubbers will be functionally tested during plant refueling SHUTDOWNS.

Observed failure of these sample snubbers shall require functional testing of additional units.

In cases where the cause of the functional failure has been identified additional testing shall be based on manufacturer's or engineering recommendations.

As applicable, this additional testing increases the probability of locating possible inoperable snubbers without testing 100K of the safety-related snubbers.

The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed

snubbers, seal

replaced, spring

replaced, in high radiation area, in high temperature area, etc.).

~ The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation

.in view of their age and operating conditions.

These records will provide statistical bases for future consideration of snubber service life.

The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation TURKEY POINT - UNITS 3 5 4 B 3/4 ?-7 AMENDMENT NOS 191 AN01

The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium.

This limitation will ensure that leakage from Byproduct,

Source, and Special Nuclear Material sources w1ll not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group.

Those sources which are frequently handled are required to be tested more often than those which are not.

Sealed sources wh1ch are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measur1ng devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

This specification is provided to ensure that.the concentration of potentially explosive gas mixtures contained in the GAS DECAY TANK SYSTEM (as measured in the inservice gas decay tank) is maintained below the flammability limits of hydrogen and oxygen.

Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive mater1als will be controlled in conformance with the requirements of General Design Cr1terion 60 of Appendix A to 10 CFR Part 50; The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification.

Restricting the quantity of radioactivity 'contained in each.Gas Decay Tank provides assurance that in the event of an uncontrolled release of the tank's contents, the result1ng whole body exposure to a

MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem.

TURKEY POINT UNITS 3 8 4

8 3/4 7-8 AMENDMENT NOS.191 AND185