ML17223A638

Amends 102 & 45 to Licenses DPR-67 & NPF-16,respectively, Revising Tech Specs in Order to Achieve Consistency,Removing Outdated &/Or Satisfied Matl,Making Minor Text Changes & Correcting Errors

ML17223A638
Person / Time
Site:	Saint Lucie
Issue date:	05/08/1990
From:	Berkow H Office of Nuclear Reactor Regulation
To:
Shared Package
ML17223A639	List: ML17223A638
References
NUDOCS 9005160085
Download: ML17223A638 (68)

Text

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UNITED STATES 1

NUCLEAR REGUI ATORY COMMISSION WASHlNGTON, D. C. 20555 FLORIDA POWER

& LIGHT COMPANY DOCKET NO; 50-335 ST.

LUCIE PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

102 License No. DPR-67 1.

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

A.

B.

The applications for amendments by Florida Power

& Light Company, et al. (the licensee),

dated September 7,

1988 and April 4, 1989, as, modified by letters dated February 1,

1990 and April 24, 1990 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; 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.

9005l600S5 900 08 PDR ADOCK 05000335 PDC

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Accordingly, Facility Operating License No.

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

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B,

as revised through Aoandment No. 102, are hereby incorporated in the license.

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

3.

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

FOR THE NUCLEAR REGULATORY COOlISSION

Attachment:

Changes to the Technical S ecifi cations P

Date of Issuance:

May 8, 1990 rbert N. Berkow, Director Proj ect Di re cto rate I I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

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ATTACHMENT TO LICENSE AMENDMENT NO.

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

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

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

Remove Pa es 82-4 82-5 3/4 3-4 3/4 3-11 3/4 3-15 3/4 3-'19 3/4 7-44 3/4 9-1 3/4 9-3 3/4 9-5 3/4 12-1 3/4 12-11 3/4 12-12 6-9 6-10 6-12 Insert Pa es 82-4 82-5 3/4 3-4 3/4 3-11 3/4 3-15 3/4 3-19 3/4 7-44 3/4 9-1 3/4 9-3 3/4 9-5 3/4 12-1 3/4 12-11 3/4 12 6-9 6-10 6-12

4l I

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

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

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

Manual Reactor Tri The Manual Reactor Trip is a redundant channel to, the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Level-Hi h

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

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

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

The trip setpoint is automatically decreased as THERMAL POWER decreases.

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

Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual t'HERMAL POWER level at which a trip would be actuated is,ll2~ of RATED THERMAL POWER, which is consistent with the value used'in the safety analysis.

Reactor Coolant Flow-Low The Reactor Coolant Flow - Low trip provides core protection against DNB in the event of a sudden significant decrease in RCS flow.

The reactor trip setpoint on low RCS 71ow is calculated by a

ST.

LUCIE - UNIT 1

B 2-4 Amendment No.

$7 g p, '..

Il

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2.2.

LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow-Low (Continued) relationship between steam genera".ur differential pressure, core inlet temperature, instrument errors and response times.

When the calculated RCS flow falls below the trip setpoint an automatic reaCtor trip signal is initiated.

The trip setpoint and allowable values ensure that for a degradation of RCS flow resulting from expected transients, a reactor trip occurs to prevent violation of local power density or DNBR safety limits.

Pressurizer Pressure-Hi h

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

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

Containment Pressure-Hi h

The Containment Pressure High trip provides assurance that a reactor trip is initiated concurrently with a safety injection.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and sub-sequent cooldown of the reactor coolant.

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

LUCIE - L'NIT 1

B 2-5 Amendment No. gg,

$5,

$8, SP.

102

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

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

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

Steam Generator Water Level - Low The Steam Generator Water Level-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded due to loss of steam generator heat sink.

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

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

A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2.

The AXIAL SHAPE INDEX is calculated from the upper and lower ex-core neutron detector channels.

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

he trip is automatically bypassed below 15 percent power.

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

In addition, CEA group sequencing in accordance with the pecifications 3.1.3.5 and 3. 1.3. 6 is assumed.

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

ST.

LUCIE - UNIT 1

B 2-6 Amendment No. 27

TABLE 3. 3-1 Continued REACTOR PROTECTIVE INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO ~

OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS APPLICABLE OPERABLE MOOES ACTION 4J I4J ll.

Wide Range Logarithmic Neutron Flux Monitor a.

Startup and Operating--

Rate of Change of Power-High b.

Shutdown 4

12.

Reactor Protection System 4

Logic 13.

Reactor Trip Breakers 2(d) 0 1,

2 and

• 3,4,5 2*

1 2*

2N 3

TABLE 3.3-1 Continued TABLE NOTATION With the protect>ve system tr>p breakers in the closed positron and the CEA drive system capable of CEA withdrawal.

¹The provisions of Specification 3.0.4 are not applicable.

(a}

Trip may be bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 1

of RATED THERMAL POWER.

(b)

Trip may be manually bypassed below 685 psig; bypass shall be automatically removed at or above 685 psig.

(c)

Trip. may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 15% of RATED THERMAL POWER.

(d)

Trip may be bypassed below 10

% and above 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL power is

> 10

% or (

15% of RATED THERMAL POWER.

(e)

Deleted (f)

There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Monitoring Channels.

ACTION 1

ACTION 2 ACTION STATEMENTS With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in 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/or open the protective system trip breakers.

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

ST.

LUCI E - UNIT 1

3/4 3-4 Amendment No. J5, gj, g5 102

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION I

~ t m

O FUNCTIONAL UNIT S.

CONTAINMENT SUMP REC.I RCULATION (RAS) a.

Manual RAS (Trip Buttons)

b. Refueling Water Tank - Low 6.

LOSS OF POWER

a. 4.16 kv Emergency Bus Under-voltage (Loss of Voltage)

b. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage)

(1) Undervoltage Device Nl (2) Undervoltage Device II2 c.

480 V Emergency Bus Under-voltage (Degraded Voltage) 7.

AUXILIARY FEEDWATER (AFAS) a.

Manual (Trip Buttons)

b. Automatic. Actuation Logic c.

SG Level (lA/lB) - Low 8.

AUXILIARYFEEDWATER I SOLATION a.

SG 1A -

SG lB Differential Pressure

b. Feedwater Header SG lA -

SG 1B Differential Pressure

'OTAL NO.

OF CHANNELS 2/Bus 2/Bus 2/Bus 2/Bus 4/SG 4/SG 4/SG 4/SG 4/SG CHANNELS TO TRIP 2/Bus 2/Bus 2/Bus 2/Bus 2/SG 2/SG 2/SG 2/SG 2/SG MINIMUM CHANNELS OPERABLE 1/Bus

'l/Bus 1/Bus 1/Bus 4/SG 3/SG 3/SG 3/SG 3/SG APPLICABLE MODES 1, 2, 3, 4

1, 2, 3

1,2,3 1,2,3 1, 2, 3

1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 ACTION 12 12 12 12 ll 8

13II', 14 13II, 14 13',

14

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TABLE 3.3-3 Continued)

TABLE NOTATION (a)

Trip function may be bypassed in this MODE when pressurizer pressure is ( 1725 psia; bypass shall be automatically removed when pressurizer pressure is

> 1725 psia.

(b)

An SIAS signal is first necessary to enable CSAS logic.

(c)

Trip function may be bypassed in this MODE below 685 psig; bypass shall be automatically removed at or above 685 psig.

The provisions of Specification 3.0.4 are not applicable.

ACTION STATEMENTS ACTION 8 With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-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 />.

ACTION 9 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

b.

Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a.

above for the inoperable channel.

c.

The Minimum Channels OPERABLE requirement is met;

however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.

ST.

LUCIE - UNIT 1

3/4 3-12 Amendment No. /Ih, 45

TABLE 3.3-4 Continued ENGINEERED SAFETY FEATURE AC UATION SYSTEM INSTRUMENTATION TRIP VALUES FUNCTIONAL UN I T 6.

LOSS OF POWER TRIP VALUE ALLOWABLE VALUES a.

(1) 4.16 kv emergency Bus Undervoltage (Loss of Voltage) b.

- 4.16 kv Emergency Bus Undervoltage (Degraded Voltage) 2900 4 29 volts with a 2900

+ 29 volts with a 1

+.5 second time delay 1

+.5 second time delay (1)

Undervol tage Device Nl (2)

Undervol tage Device 82 c.

480 volts Emergency Bus Undervoltage (Degraded Vol tage) 7.

AUXILIARY FEEDWATER (AFAS) a.

Manual (Trip Buttons) b.

Automatic Actuation Logic c.

SG lA 8 1B Level Low 8.

AUXILIARY FEEDWATER ISOLATION a.

Steam Generator aP-High b.

Feedwater Header High aP 3675

+ 36 vol ts wi th a 7

+

1 minute time delay 3592

+ 36 vol ts wi th a 18

+ 2 second time delay 429

+ 5-0 vol ts with a 7

+

1 second time delay Not Applicable Not Applicable

>29.0X

<275 psid

<150.0 psid 3675

+ 36 volts with a 7

+

1 minute time delay 3592

+ 36 volts with a 18 + 2 second time delay t

429

+

5 -0 vol ts with a 7

+

1 second time delay Not Applicable Not Applicable

>28.5X

<281 psid

<157.5 psid

TABLE 3.3-5 ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION 1.

manual a.

SIAS Safety Injection (ECCS)

Containment Fan Coolers Feedwater Isolation Containment Isolation b.

CSAS Containment Spray c.

CIS Containment Isolation Shield Building Yenti lation System d.

RAS Containment Sump Recirculation e.

MS IS Main Steam Isolation Feedwater Isolation AFAS Auxiliary Feedwater Actuation 2.

Pressurizer Pressure-Low a.

Safety Injection (ECCS) b.

Containment Isolation ***

c.

Containment Fan Coolers d.

Feedwater Isolation

RESPONSE

TIME IN SECONDS Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not,Applicable Not Appl icabl e

< 30.0*/19.5**

3Q 5*/20 5**

3Q Q*/1 7 Q**

< 60.0 ST.

LUCIE - UNIT 1

3/4 3-16 Amendment No. J7,87,89~ "

TABLE 4.3-2 Continued ENGINEERED SAFETY. FEATURE ACTUATION SYSTEM INS RUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 6.

LOSS OF POWER a.

4.16 kv Emergency Bus Under-voltage (Loss oi Voltage) b.

4.16 kv Emergency Bus Under-voltage (Degraded Voltage)

(1) Undervoltage Device Nl (2) Undervoltage Device N2 480 V Emergency Bus Under-voltage (Degraded Voltage) 7.

AUXILIARY FEEDWATER (AFAS) a.

Manual (Trip Buttons) b.

SG Level (A/B) - Low c.

Automatic Actuation Logic 8.

AUXILIARYFEEDWATER ISOLATION a.

SG Level (A/B) - Low and SG Differential Pressure (BtoA/AtoB) - High b.

SG Level (A/B) - Low and Feedwater Header Differential

- Pressure (BtoA/AtoB) - High CHANNEL CHECK N.A.

N.A.

N.A.

N.A.

CHANNEL CALIBRATION N.A.

N.A.

FUNCTIONAL TEST R

MODES IN WHICH SURVEI LLANO RE VIREO 1,2,3 1, 2, 3

1, 2.

3 1, 2, 3

1, 2, 3

1, 2, 3

1, 2, 3

1,2,3 1,2,3

TABLE 4.3-2 Continued TABLE NOTATION (1)

The logic circuits shall be tested manually at least once per 31 days.

ST.

LUCIE - UNIT 1

3/4 3-'20

PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.11.2 The fire hose stations shown in Table 3.7-3 shall be OPERABLE.

APPLICABILITY:

Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.

ACTION:

a.

With one or more of the fire hose stations shown in Table 3.7-3 inoperable, route an additional equivalent capacity fire hose to the unprotected area(s) from an OPERABLE hose station within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Restore the fire hose station to OPERABLE status within 14 days or, in lieu of any other 'report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the station to OPERABLE status.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.2 Each of the fire hose stations shown in Table 3.7-3 shall be demonstrated OPERABLE:

a.

At least once per 31 days by visual inspection of the station to assure all required equipment is at the station.

b.

At least once per 18 months by:

c ~

1.

Removing the hose for inspection and re-racking, and 2.

Replacement of all degraded gaskets in couplings.

At least once per 3 years by:

1.

Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.

2.

Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater.

(Hoses on exterior hose stations shall be hydrostatic tested once per year.)

ST.

LUCIE - UNIT 1

3/4 7-43 Amendment No.

8$ ~ 66

TABLE 3. 7-3 FIRE HOSE STATIONS A.

Hose Stations (Turbine Building) 1.

Operating Floor (northeast corner) 2.

Operating Floor (southeast cor ner) 3.

Operating Floor (middle east side)

B.

Hose Stations (Reactor Auxiliary Building) l.

43 ft. level south wall of HVE room 2.

43 ft. level "8" switc."gear room by roll-up door 3.

43 ft. level southwest corner of "B" switchgear room near door.

4.

43 ft. level cable spreading room west wall 5.

19.5 ft. level east end of east-west hall 6.

19.5 ft. level middle of east-west hall 7.

19.5 ft. level south end of north-south hall 8.

19.5 ft. level entrance hall on south wall 9.

-5 ft. level east end of hall 10...-5 ft. level south wall of hall near HCC 182 11.

-5 ft. level west end of hall ST.

LUCIE - UNIT 1

3/4 7-44 Amendment No 2$ ~ 8 102

3/4. 9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3,9.1 With the reactor vessel head unbolted or removed, the boron concentration of all-filled portions of the Reactor Coolant System and the refueling cavity shall be maintained uniform and sufficient to ensure that the more restrictive of following reactivity conditions is met:

a.

Either a Keff of 0.95 or less, which includes a 1000 pcm conservative allowance for uncertainties, or b.

A boron concentration of > 1720 ppm, which includes a

50 ppm conservative allowance for uncertainties.

APPLICABILITY'ODE6*.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all o'perations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at 40 gpm of 1720 ppm boron or its equivalent until K qq is reduced to

< 0.95 or the boron concentration is restored to

> )720 ppm, whichever is the more restrictive.

The provisions of Specification 3,0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.1,1.

The more restrictive of the above two reactivity conditions shall be determined prior to:

a.

Removing or unbolting the reactor vessel

head, and b.

Withdrawal ot any full length CEA in excess of 3 feet from its fully inserted position.

4; 9.1.2 The boron concentration of the refueling cavity shall be determined by chemical analysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

<<The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

ST.

LUGI E - UNIT 1

3/4 9-1 Amendment No.

REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two wide range logarithmic neutron flux monitors shall be operating, each with continuous visual indication in the con-trol room and one with audible indication in the containment.

APPLICABILITY:

MODE 6.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

The provisions of Specification 3.0.3 are not applicable.

SURYEILLANCE RE UIREMENTS 4,9.2 Each wide range logarithmic neutron flux monitor shall be demonstrated OPERABLE by performance of:

a.

A CHANNEL FUNCTIONAL TEST at least once per 7 days.

b.

A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the start of CORE ALTERATIONS, and c.

A CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.

ST.

LUCIE - UNIT 1

3/4 9-2

REFUEL ING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel.

ACTION:

With the reactor subcritical for less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, suspend all operations involving movement of irradiated fuel in the reactor pres-sure vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by verification, of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

ST.

LUCI E - UNIT 1

3/4 9-3 Amendm nt No. 102

REFUELING OPERAT IONS CONTAINMENT PENETRAT IONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment penetrations shall be in the following status:

a.

The equipment door closed and held in place by a minimum of four bolts, b.

A minimum of one door in each airlock is closed, and c,

Each penetration, except as provided in Table 3.6-2 of Specification 3.6.3.1, providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1.

Closed by an isolation valve, blind flange, or manual valve, or 2.

Be capable of being closed by an OPERABLE automatic containment isolation valve, or 3.

Be capable of being closed by an OPERABLE containment vacuum relief valve.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment'CT ION:

With the requirements of the above specification not satisfied, immedi-ately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment'he provisions of Specification 3.0.3 are not applicable.

SURVE ILLANCE REQUIREMENTS 4.9 '

Each of the above required containment penetrations shall be determined to be either in its closed/isolated condition or capable of being closed by an OPERABLE automatic containment isolation valve within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment by:

a

~

Verifying the penetrations are in their closed/isolated condition, or b.

Testing the containment isolation valves per the applicable portions of Specifications 4 '.3,1.1 and 4.6.3.1.2.

ST

~ LUCIE - UNIT 1

3/4 9-4 Amendment No.

17

REFUELING OPERATIONS COMMUN!CATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.

APPLICABIlITY: During CORE ALTERATIONS.

ACTION:

When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.

ST.

LUCIE - UNIT 1

3/4 9-5 Amendment No. 102

REFUELING OPERATIONS MANIPULATOR CRANE OPERABILITY LIMITING CONDITION FOR OPERATION 3.9.6 The manipulator crane shall be used for movement -of CEAs or fuel assemblies and shall be OPERABLE with:

a.

A minimum capacity of 2000 pounds, and b,

An overload cut off limit of

< 3000 pounds.

APPLICABILITY: During movement of CEAs or fuel assemblies within the reactor pr essure vessel.

ACTION:

With the requirements for crane OPERABILITY not satisfied, suspend use of any inoperable manipulator crane from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.6 The manipulator crane used for movement of CEAs or fuel assem-blies within the reactor pressure vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior. to the start of such operations by performing a

load test of at least 2500 pounds and demonstrating an automatic load cut off when the crane load exceeds 3000 pounds.

ST.

LUCIE - UNIT 1

3/4 9-6

3 4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3 4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.

APPLICABILITY:

At all times.

ACTION:

a.

With the radiological environmental monitoring program not being con-ducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1,8, a description of the reasons for

)

not conducting the program as required and the plans for preventing a recurrence.

b.

With the confirmed* level of radioactivity as the result of plant eff1uents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Cowlission within 30 days, pursuant to Specification 6,9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a

MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3.

When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

C.

concentration 1

+

concentration 2

+...

> 1.0 reporting level 1

reporting level 2

When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a

MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3.

This report shall include the methodology for calculating the, cumulative potential dose contributions for the calendar year from radionuclides detected in environmental samples and can be determined in accordance with the methodology and parameters in the ODCM.

This r eport is not required if the measured level of radioactivity was not the result of plant effluents;

however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

With milk or broadleaf vegetation samples unavailable from one or more of the sample locations required by Table 3.12-1, identify locations

• A confirmatory reanalysis of the original, a duplicate, or a new sample may be desirable, as appropriate.

The results of the confirmatory analysis shall be completed at the earliest time consistent with the analysis but in any case

.within 30 days.

ST.

LUCIE - UNIT 1

3/4 12-1 Amendment No. /g)~)) ~1O2

RADIOLOGICAL ENVIRONMENTAL MONITOR ING LIMITING CONDITION FOR OPERATION Continued ACTION:

(Continued) for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days.

The specific loca-tions from which samples were unavailable may then be deleted from the monitoring program.

Pursuant to Specification 6.9.1.106 identify the cause of the, unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a

"evised figure(s) and table for the ODCM reflecting the new location(s).

d.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the table and figure(s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1.

ST.

LUCIE - UNIT 1

3/4 12-2 Amendment No.

$/9',6g

~

~

RADIOLOGiCAL ENVIRONMENTAL MONITORING 3 4.12.2 'AND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sector s of the nearest milk animal, the nearest residence and the nearest garden* of greater than 50 m2(500 ft ) producing broad leaf vegetation.

APPLICABILITY: At all times.-

ACTION:

a

~

With a land use census identifying a location(s) that yields a cal-culated dose or dose commitment greater than the values currently being calculatea in Specification 4.11.2.3i identify the new loca-tion(s) in the next Semiannual Radioactive Effluent Release

Report, pursuant to Specification 6.9.1. 7.

b.

With a land use census identifying a location(s) that, yields a cal-culated dose or dose commitment (via the same exposure pathway) 20%

gre'ater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new loca-tion(s) to the radiological environmental monitoring program within 30 days.

The sampling location(s),

excluding the control station

location, having the lowest calcuIated dose or dose commitment(s),

via the same exposure

pathway, may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted, Pursuant to Specification 6.9.1.7 identify the new location(s) in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure(s) 'and table for the ODCM reflecting the new location(s).

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE R'E U I REMEN TS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best

results, such as by a door-to-door survey, aerial

survey, or by consulting local agriculture authorities.

The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.8.

• Broad leaf vegetation sampling may be performed at the site boundary in each of two different direction sectors with the highest predicted D/gs in lieu of the garden census.

Specifications for broad leaf vegetation sampling in Table 3.12-1.4b shall be followed, including analysis of control samples.

ST.

LUG I E - UNIT 1

3/4 12-11 Amendment No. O'9,)3',

102

RADIOLOGICAL ENVIRONMENTAL MONITORING 3 4.12.3 IflTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission.*

APPLICABILITY: At all times.

ACTION:

a.

With analyses not being performed as required above, report the corrective actions to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.8.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

R)

SURVEILLANCE RE UIREMENTS 4.12.3 A summary of the results obtained as. part of the above required Inter-laboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.8.

• This condition is satisfied by participation in the Environmental Radioactivity Laboratory Intercomparison Studies Program conducted by the Environmental Protection Agency (EPA).

ST.

LUCIE - UNIT 1

3/4 12-12 Amendment No.

$ g,

ADMINISTRATIVE CONTROLS 6.5.2 COMPANY NUCLEAR REVIEW BOARD CNRB)

FUNCTION 6.5.2.1 The Company Nuclear Review Board shall function to provide indepen-dent review and audit of designated activities in the areas of:

a.

nuclear power plant operations b.

nuclear engineering c.

chemistry and radiochemistry d.

metallurgy e.

instrumentation and control f.

radiological safety g.

mechnical and electrical engineering h.

quality assurance practices COMPOS IT ION 6.5.2.2 The Executive Vice President shall appoint, in writing, a minimum of five members to the CNRB and shall designate from this membership, in

writing, a Chairman.

The membership shall function to provide independent review and audit in the areas listed in Specification 6.5.2.1.

The Chairman shall meet the requirements of ANSI/ANS-3.1-1987, Section 4.7.1.

The members of the CNRB shall meet the educational requirements of the ANSI/ANS-3.1-1987, Section 4.7.2, and have at least 5 years of professional level experience in one or more of the fields listed in Specification 6.5.2.1.

CNRB members who do not possess the educational requirements of ANSI/ANS-3.1-1987, Section 4.7.2 (up to a maximum of 2 members) shall be evaluated, and have their membership approved and documented, in writing, on a case-by-case basis by the Executive Vice President, considering the alternatives to educational requirements of ANSI/ANS-3.1-1987, Sections 4.1.1 and 4,1.2.

ALTERNATES 6.5.2.3

".11 alternate members shall be appointed in writing by the CNRB Chair-man to serve on temporary basis;

however, no mor e than two alternates shall participate as voting members in CNRB activities at any one time.

ST, LUCIE - UNIT 1

6-9 A endm lo>

ADMINISTRATIVE CONTROLS CONSULTANTS 6.'5.2.4 Consultants shall be utilized as determined by the CNRB Chairman to provide expert advice to the CNRB.

MEETING FRE UENCY 6.5.2.5 The CNRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter and as convened by the CNRB Chairman or his designated alternate.

UORUM 6.5.2.6 The quorum of the CNRB necessary for the performance of the CNRB review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least a majority of Cling members including alternates.

No more than a minority of the quorum shall have line responsibility for operation of the facility.

REVIEW 6.5.2,?

The CNRB shall review:

a.

C.

d.

The safety, evaluations for (1) changes to procedures, equipment, or systems and (2) tests or experiments completed under the provisions of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.

Proposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

Proposed changes to Technical Specifications or this Operating License.

e.

Violations of codes, regulations,

orders, Technical Specifications, license requirements, or of internal procedures or instructions

. having nuclear safety significance.

Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety.

ST.

LUCIE - UNIT 1

6-10 Amendment No.

ESs$ $ e 0 a

102

~ ~ ~

il ADMINISTRATIVE CONTROLS g ~

h.

All REPORTABLE EVENTS.

All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures,

systems, or components that could affect nuclear safety.

Reports and meeting minutes of the Facility Review Group.

AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the CNRB.. These audits shall encompass:

a

~

The conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b. 'he performance, training and qualifications of the entire unit C.

d.

e.

staff at least once per 12 months.

The results of actions taken to correct deficiencies occurring in unit equipment, structures,

systems, or method of operation that affect nuclear s'afety at least once per 6 months.

The performance of activities required by the guality Assurance Program to meet the-criteria of Appendix 8, 10 CFR Part -50, at least once per 24 months.

Any other area of unit operation considered appropriate by the CNRB or the Executive Vice President.

g.

,h.

The fire protection prograamatic controls including the implementing procedures. at least once per 24 months by qualified licensee gA personnel.

The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant.

An outside independent fire protection consultant shall be used at least every third year.

The radiological environmental monitoring program and the results thereof at least once per 12 months'he OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

The PROCESS CONTROL PROGRAM and implementing procedures for dewater-ing of radioactive bead resin at least once per 24-months.

ST.

LUCIE - UNIT 1

6-11 Amendment No. 25~$ 9,69

ADMINISTRATIVE CONTROLS AUTHORITY 6.5.2.9 The CNRB shall report to and advise the Executive Vice President on those areas of responsibility specifed in Specifications 6.5.2.7 and 6.5.Z.B.

RECORDS 6.5.2.10 Records of CNRB activities shall be prepared, approved and distrib-uted as indicated below:

a.

Minutes of each CNRB meeting shall be prepared, approved and forwarded to the Executive Vice President within 14 days following each meeting.

b.

Reports of reviews encompassed by Specification 6.5.2.7

above, shall be prepared, approved and forwarded to the Executive Vice President within 14 days following completion of the review.

C.

Audit reports encompassed by Specification 6.5.2.8 above, shall be forwarded to the Executive Vice President and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS; a.

The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the

FRG, and the results of the review shall be submitted to the CNRB, and the Senior Vice President - Nuclear.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

'a

~

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The Senior Vice President

- Nuclear and the CNRB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A Safety Limit Violation Report shall be prepared.

The report shall be reviewed by the FRG.

This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

ST.

LUCIE - UNIT 1

6-12 Amendment No.

~J~ 7),

g6"8 RE~

~

P()

"o 5

/p

~Oy*y4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FLORIDA POWER

& LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST.

LUCIE PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 45 License No. NPF-16 1.

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

A.

The applications for amendments by Florida Power 8 Light Company, et al. (the licensee),

dated September 7,

1988 and April 4, 1989, as modified by letters dated February 1,

1990 and April 24, 1990, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the 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.

~

~

I

W 2.

Accordingly, Facility Operating License No.

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

2.

Technical S ecifications The Technical Specifications contained in Appendices A and 8

as revised through Amendment No. <5, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Techni cal Specifications 3.

This license amendment is effective as of the date of its issuance Attac haunt:

Changes to the Technical Specifications Date of Issuance:

Nay 8 1990 FOR THE NUCLEAR REGULATORY COt1NISSION Y)

Herbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE NENDHENT NO.45 TO FACIL'TY OPERATING LICENSE NO.

NPF-16 DOCKET NO. 50-389 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

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

The corresponding overleaf pages are also provided to maintain document comp lete ness.

Remove Pa es B2-1 B2-6 3/4 3-41 3/4 3-45 3/4 3-46 3/4 6-1 3/4 7-16 3/4 7-22 3/4 7-32 3/4 7-36 3/4 7-37 3/4 12-1 3/4 '12-11 3/4 12-12 6-1 6-6 6-10 6-23 Insert Pa es B2-1 82-6 3/4 3-41 3/4 3-45 3/4 3-46 3/4 6-1 3/4 7-16 3/4 7-22 3/4 7-32 3/4 7-36 3/4 7-37 3/4 12-1 3/4 12-11 3/4 12-12 6-1 6-6 6-10 6-23

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel clad-ding and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centerline fuel melting will occur.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coeffi-cient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the CE-1 correlation.

The CE-1 DNB correlation has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

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

The minimum value of the DNBR during steady state operation, normal opera-tional transients, and anticipated transients is limited to 1.28.

This value is derived through a statistical combination of the system parameter probability distribution functions with the CE-1 DNB correlation uncertainty.

This value corresponds to a 95 probability at a

95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all opeeating conditions.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Cool-ant Pumps operating for which the minimum DNBR is no less than 1.28 for the family of axial shapes and corresponding radial peaks shown in Figure B 2.1-1.

The limits in Figure 2.1-1 were calculated for reactor coolant inlet temperatures less than or equal to 580'F.

The dashed line at 580'F coolant inlet temperature is not a safety limit; however, operation above 580'F is not possible because of the actuation of the main steam line safety valves which-limit the maximum value of reactor inlet temperature.

Reactor operation at THERMAL POWER levels higher than 112 of RATED THERMAL POWER is prohibited by the high power level trip set-point specified in Table 2.2-1.

The area of safe operation is below and to the

(

left of these lines.

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

The Thermal Margin/Low Pressure and Local Power Density Trip Systems, in conjunction with Limiting Conditions for Operation, the Variable Overpower Trip and the Power 'Dependent Insertion Limits, assure that the Specified Acceptable Fuel Design Limits on DNB and Fuel Centerline Melt are not exceeded during normal operation and design basis Anticipated Operational Occurrences.

ST.

LUCIE-UNIT 2 8 2-1 Amendment No. P, 45

C/l In Dl I

2.0 1.6 cx 1.4 1.2 lalo 1.0 0.8

>C o

0.6 LLJ l4 0.4 C) 0.2 0.0 FT = 1.67 T

1.81 25 FR

= 1.79 FTR = 1.77 FT 1

u R

50 75 100 PERCENT OF ACTlVE CORE LENGTH FROH BOTTOM figure 0 2.1-1 Axial ~)ower distribution for tbeonal oiargin safety liniits

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Containment Pressure-Hi h

The Containment pressure-High trip provides assurance that a reactor trip is initiated prior to or concurrently with a safety injection (SIAS).

This also provides assurance that a reactor trip is initiated prior to or concurrently with an MSIS.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant.

7he setpoint of 620 psia is sufficiently below the full load operating point of approximately 885 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow.

This setting was used with an uncertainty factor of 30 psi in the safety analyses, Steam Generator Level-Low The Steam Generator Level-Low trip provides protection against a loss of feedwater flow incident and,assures that the design pressure of the Reactor Coolant System will not be exceeded due to loss of the steam generator heat sink.

This specified setpoint provides allowance that there will be sufficient water inventory in the steam generator at the time of the trip to provide a

margin of at least 10 minutes before auxiliary feedwater is required.

This trip also protects against violation of the specified acceptable fuel design limits (SAFDL) for DNBR, offsite dose and the loss of shutdown margin for asymmetric steam generator transients such as the opening of a main steam safety valve or atmospheric dump valve.

Local Power Densit -Hi h

'The Local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a

consequence of axial power maldistributions.

A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2.

The AXIAL SHAPE INDEX is calculated'rom the upper and lower excore neutron detector channels.

The calculated setpoints are generated as a function of THERMAL POWER level with the allowed CEA group positiion being inferred, from the THERMAL POWER level.

The trip is automatically bypassed below 15% power.

The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints.

In

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

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

Power Level-High trip is assumed.

ST

~

LUCIE - UNIT 2 8 2-5 Amendment No. 23

SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS BASES RCP Loss of Com onent Cool in Mater A loss of component cooling water to the reactor coolant pumps causes a

delayed reactor trip.

This trip provides protection to the reactor coolant pumps by ensuring that plant operation is not continued without cooling water avalable.

The trip is delayed l0 minutes following a reduction in flow to below'he trip setpoint and the trip does not occur if flow is restored before 10 minutes elapses.

No credit was-taken for this trip in the safety analysis.

Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reacto~ Protective System.

Rate of Chan e of Power-Hi h

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

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

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

Reactor Coolant Flow -

Low The Reactor Coolant Flow -

Low trip provides core protection against DNB in the event of a sudden significant decrease in RCS flow.

The reactor trip setpoint on low RCS flow is calculated by a relationship between steam generator differential pressure, core inlet temperature, instrument errors and response times.

'.,'hen the calculated RCS flow fails below the trip setpoint an automatic reactor trip signal is initiated.

The trip setpoint and allowable values ensure that for a degradation of RCS flow resulting from expected transients, a reactor trip occurs to prevent violation of local power density or DNBR safety limits.'T.

LUCIE - UNIT 2 B 2-6 Amendment No.

~e>

~a

~

1

~

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3 3 3 6

The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

a.*

With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3:3-10, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.*

With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c.** With the number of OPERABLE Channels one less than the Total Number of Channels shown in Table 3,3-10, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant,to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

d:** With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are feasible without shutting down or:

l.

Initiate an alternate method of monitoring the reactor vessel i'nventory; and 2.

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event out-lining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status, and 3.

Restore the Channel to OPERABLE status at the next scheduled refuel ing.

e.

The provisions of Specification 3.0.4 are not applicable.

• Action statements do not apply to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level (wide range) instruments..

• "Action statements apply only to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level

{wide range) instruments.

ST.

LUC IE - UNIT 2 3/4 3-41 Amendment No.

19 s 45

INSTRUMENTATION ACCIOENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS l

4.3.3.6 Each accident monitoring instrumentation channel will be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

ST.

LUCIE - UNIT 2 3/4 3-4la Amendment No. lg

INSTRUME'NT LOCATION REACTOR AUXILIARY.BUILDING HEAT (x/y)

SMOKE (x/y)

TABLE 3. 3-11 FIRE DETECTION INSTRUMENTS TOTAL NUMBER OF INSTRUMENTS" ZONE-lA REACTOR AUX.

BLDG.

EL. 0.50 ZONE-2A REACTOR AUX.

BLDG.

EL. 0.50 ZONE-3A REACTOR AUX.

BLDG.

EL. 19.50 ZONE-4A REACTOR AUX. E'LOG.

EL'. 19.50 ZONE-5A REACTOR AUX.

BLDG.

EL. 19.50 ZONE-6A REACTOR AUX.

BLDG. EL.'3.00 ZONE"7A REACTOR AUX.

BLDG.

EL. 43.00 ZONE-BA REACTOR AUX.'LDG. EL. 62.00 ZONE-9A REACTOR AUX. BLDG.

EL. 43.00 ZONE-10A REACTOR AUX.

BLDG.

EL. 43.00 ZONE-12A ELECT.

PEN.

ROOM EL. 19.50 ZONE-1B REACTOR AUX.

BLDG.

EL.

0; 50 ZONE-2B REACTOR AUX.

BLDG.

EL. 0.50 ZONE-3B REACTOR AUX.

BLDG.

EL.

19.50 ZONE"48 HSCP-1 REC.

AUX.

BLDG.

EL. 43.00 ZONE-5B REACTOR AUX.

BLDG.

EL. 19.50 ZONE-68 REACTOR AUX. BLDG.

EL. 43.00 ZONE-7B REACTOR AUX. BLDG.

EL. 43.00 ZONE-BB REACTOR AUX.

BI DG.

EL. 62.00 ZONE-98 REACTOR AUX. BLDG.

EL. 43.00 ZONE-10B REACTOR AUX.

BLDG.

EL. 43.00 ZONE-128.ELECT.

PEN.

ROOM EL. 19.50 ZONE-1F FAN ROOM EL. 43. 00 ZONE-2F CABLE LOFT EL.

19.50 ZONE-3F IODINE REMOVAL/WASTE GAS/

HALLWAYS EL. 0.50 ZONE-4F B ELECTRICAL PENETRATION ROOM EL.

19. 50 ZONE-5F A ELECTRICAL PENETRATION ROOM EL.

19. 50 ZONE-6F CABLE SPREADING ROOM EL. 43.00 FUEL HANDLING BUILDING ZONE-20A FUEL HANDLING BLDG.

EL.

19. 50 ZONE-21A FUEL HANDLING BLDG.

EL. 48.00 ZONE-20B FUEL HANDLING BLDG.

E L. 19. 50 ZONE-21B FUEL HANDLING BLDG.

EL. 48. 00 DIESEL GENERATOR BUILDING 2/0 1/0 0/2 0/26 0/15 0/2 0/1 0/9 6/0 4/0 6/0

, 5/0 8/0 5/0 7/0 6/0 2/0 2/0 3/0 6/0 5/0 6/0 1/0 6/0 4/0 6/0 5/0 2/0 2/0 4/0 1/0 3/0 1/0 2/0 ZONE-22A DIESEL GEN. BLDG./9.0.

STORAGE TANK 2/2 ZONE-22B DIESEL GEN.

BLDG./D.O.

STORAGE TANK 2/2 2/0 2/0 ST.

LUGIE - UNIT 2 3/4 3-45 Amendment No. 45

INSTRUMENT LOCATION TABLE 3.3-11 (Continued)

FIRE DETECTION INSTRUMENTS TOTAL NUMBER OF INSTRUMENTS" SAFETY RELATED PUMPS ZONE-17A COMPONENT COOLING AREA ZONE-18A INTAKE COOLING WATER PUMP AREA ZONE-19A STEAM TRESTLE AREA-AUX.

FEEDWATER PUMP ZONE-17B COMPONENT COOLING AREA ZONE-18B INTAKE COOLING WATER PUMP AREA ZONE-198 STEAM TRESTLE AREA-AUX.

FEEDWATER PUMP TURBINE BUILDING/SWITCHGEAR ROOM ZONE-16A TURBINE BLDG.

SWITCHGEAR ROOM ZONE-168 TURBINE BLDG.

SWITCHGEAR ROOM CONTAINMENT HEAT

~x/y) 1/0 1/0 SMOKE

~(x y) 4/0 2/0 2/0 2/0 3/0 3/0 ZONE-11A ZONE-13A ZONE-14A ZONE-15A ZONE-118 20!!E-13B ZO!<E-14B ZQllE-158 ANNULUS REACTOR TUNNEL BELOW EL. 18.00 REACTOR =L. 18.00 REACTOR EL. 45.00 A!!NULJS REACTOR, TUNNEL BELOW EL. 18.00 REACTOR EL. 18.00 REACTOR EL. 45.00 2/0 2/0 1/0, 2/0 5/0 4/0 1/0 1/0 5/0 5/0 (x/y):

x is number of ear ly warning fire detection and notification only i,nstruments.

y is number of actuation of fire suporession systems and early warning. notification instruments.

¹The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate Tests.

ST.

LUGIE - UNIT 2 3/4 3-46 Amendment No.45

3/4.6 CONTAINMENT SYSTEMS 3/4. 6. 1 PRIMARY CONTAINMENT CONTAINHENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

MODES 1", 2", 3, and 4.

ACTION:

Without primary CONTAINHENT INTEGRITY, restore CONTAINMENT INTEGRITY 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 />.

SURVEILLANCE RE UIREMENTS

4. 6. 1.

1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a.

C.

At least once per 31 days by verifying that all penetrations""

not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-2 of Specification 3.6.3

~

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6. 1.3.

After each closing of each penetration subject to Type B testing, except containment air locks, if opened following a Tyoe A or B

test, by leak rate testing the seal with gas at P

, 41.8 psig and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specifica-tion 4. 6. l. 2d. for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L

In MODES 1 and 2, the RCB polar crane shall be rendered inoperable by locking the power supply breaker open.

Except valves, blind flanges, and deactivated automatic valves which are located insi'de the containment and are

locked, sealed or otherwise secured in ttie closed position.

These penetrations shall be verified closed during each COLO SHUTDOWN except that such verification need not be performed more often than once per 92 days.

ST.

LUCIE - UNIT 2 3/4 6-1 Amendment

."Io

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6. 1.2 Containment leakage rates shall be limited to:

a.

An overall integrated leakage rate of:

1.

Less than or equal to L

, 0.50 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P

, 41.8 psig, or Less than or equal to L

, 0.35 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a reduced pressure of P

20.9 psig.

A combined leakage rate of less than or equal to 0.60 L

for all a

penetrations and valves subject to Type B and C tests, when pressurized to P

a'.

A combined bypass leakage rate of less of,than or equal

0. 12 L

for a

all penetrations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P

a'PPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding

0. 75 L

or 0.75 Lt, as applicable, or (b) with the measured combined a

leakage rate for all penetrations and valves subject to Types 8 and C tests exceeding 0.60 L

or (c) with the combined bypass leakage rate exceeding a'.

12 L

, restore the overall integrated leakage rate to less than or equal a'o 0.75 L

or less than or equal to 0.75 L, as applicable, and the combined leakage rate for all. penetrations and valves subject to Type B and C tests to less than or equal to 0.60 L

and the bypass leakage rate to less than or a'qual to 0. 12 L

prior to increasing the Reactor Coolant System temperature a

above 200 F.

SURVEILLANCE RE UIREMENTS 4.6. 1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50:

a.

Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40

+

10 month intervals during ST.

LUCIE - UNIT 2 3/4 6"2 Amendment No. )P, 37

PLANT. SYSTEMS 3/4. 7. 5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5. 1 The ultimate heat sink shall be OPERABLE with:

a.

Cooling water from the Atlantic Ocean providing a water level above

-10.5 feet elevation, Mean Low Water, at the plant intake structure, and b.

Two OPERABLE valves in the barrier dam between Big Mud Creek and the intake structure.

APPLICABILITY: At al 1 times.

ACTION:

a.

With the water level requ'irement of the above specification not satisfied, 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 provide cooling water from Big Mud Creek within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With one isolation valve in the barrier dam between Big Mud Creek and the intake structure inoperable, restore the inoperable valve 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 within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, install a temporary flow barrier and open the barrier dam isola-tion valve.

The availability of the onsite equipment capable of removing the barrier shall be verified at least once per 7 days thereafter.

c.

With both of the isolation valves in the barrier dam between the in-take structure and Big Mud Creek inoperable, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, either:

1.

Install both temporary flow barriers and manually open both barrier dam isolation valves.

The availability of the onsite equipment capable of removing the barriers shall be verified at least once per 7 days thereafter, or 2.

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 />.

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.5. l. 1 The ultimate heat sink shall be determined OPERABLE at least once. per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the average water level to be within limits.

4.7.5. 1.2 The isolation valves in the barrier dam between the intake structure and Big Mud Creek shall be demonstrated OPERABLE at least once per 6 months by cycling each valve through at least one complete cycle of full travel.

ST.

LUC!E - UNIT 2, 3/4 7-15

PLANTS SYSTEMS 3/4.7. 6 FLOOO PROTECTION LIMITING CONDITION FOR OPERATION 3.7.6.1 Flood protection shall be provided for the facility site via stoplogs which shall be installed on the southside of the RAB and the southernmost door on east wall whenever a hurricane warning for the plant is posted.

'PPLICABILITY:

At al 1 times ACTION:

With either a Hurricane Watch or a Hurricane Warning issued for the facility

site, perform the St.

Lucie Plant Beach Survey Procedure pursuant to Surveil-lance Requirement

4. 7.6. l. 1 below and ensure the stoplogs are removed from storage and are prepared for installation.

The stoplogs shall be installed anytime a hurricane warning is posted.

SURVEILLANCE RE UIREMENTS 4.7.6. 1. 1 The St.

Lucie Plant Beach Survey Procedure shall be conducted at least once per year between the dates of May 25 and June 7 and within 30 days following the termination of either a Hurricane Watch or a Hurricane Warning for the facility site.

A Special Report containing the results of these surveys shall be prepared and submitted to the Commission pursuant to Speci i-cation 6.9.2 within 30 days following the completion of the survey.

4. 7. 6. 1. 2 The St.

Lucie Mangrove Photographic Survey Procedure shall be conducted at least once per 12 months and shall be a color infrared photo-graph(s),

'or equivalent, of the mangrove area between the facility and the FP8 L east property line.

The results of these surveys shall be included in the Annual Operating Report for the period in which the survey was completed.

This report shall include an evaluation of the facility flood protection if the survey indicates deterioration, either man-made or natural, of this mangrove area.

4.7.6. 1.3 Meteorological forecasts shall be obtained from the National Hurricane Center in Miami, Florida at least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> during either a

Hurricane Watch or a Hurricane Warning.

-ST.

LUCIE - UNIT 2 3/4 7-16 Amendment

.'lo. 46

PLANT SYSTEMS A

3/4.7.9 SNUBBERS LIMITING CONDITION FOR OPERATION 3 ~ 7.9 All safety-related snubbers shall be OPERABLE.

AppLICABILITY:

MODES 1, 2, 3, and 4.

MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.

ACTION:

With one or more safety related snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.9g.

on the supported component or declare the supported system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE RE UIREMENTS 4.7.9 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program.

a.

Ins ection T

es As used in this -specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.

b.

Visual Ins ections Visual inspections shall be performed in accordance with. the following schedule:

No. Inoperable Snubbers of Each T

e er Inspection Period Subsequent Visual Ins ection Period*4 0

1 2

3,4 5, 6, 7

8 or more 18 months

+ 25K 12 months

+ 25%

6 months

+ 25K 124 days

+ 25%

62 days

+ 25K 31 days

+ 25K

• The inspection interval for each type of snubber shall not be lengthened

.more than one step at a time unless a generic problem has been identified and corrected; in that event the inspection interval may be lengthened one step the first time and two steps thereafter if no inoperable snubbers of that type are found.

.jThe. provisions of Specification 4.0.2 are not applicable.

ST.

LUC IE - UNIT 2 3/4 7-21 Amendment No. 22

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS (Continued c.

Refuelin Outa e Ins ections At least once per 18 months an inspection shall be performed of all safety related snubbers attached to sections of safety systems piping that have experienced unexpected, potentially damaging transients as determined from a review of operational data and a

visual inspection of the systems.

In addition to satisfying the visual inspection acceptance

criteria, freedom of motion of mechanical snubbers shall be verified using one of the following:

(1) manually induced snubber movement; (2) evaluation of in-place snubber piston setting; (3) stroking the mechanical snubber through its full range of travel.

d.

Visual Ins ection Acce tance Criteria Visual inspections shall verify (1) that there are no visble indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure.

Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the r ejection is clearly established and remedied for that particular snubber and for other

snubbers, irrespective of type, that may be generically susceptible; and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specification 4.7.9f.

When a

fluid port of a hydraulic snubber is found to be uncovered the snubber shall be declared inoperable and cannot be determined OPERABLE via functional testing unless the test is star ted with the piston in the as found setting, extending the piston rod in the tension mode direction.

All snubbers connected to an inoperable common hydraulic fluid reservoir shall be determined to be OPERABLE by visually verifying the required level of oil for operation for each affected snubber; otherwise declare the snubbers inoperable.

e.

Functional Tests During the first refueling shutdown and at lea'stonce per 18 months thereafter during shutdown, a representative sample of either:

(1) At least 10% of the total of each type of safety related snubber in use in the plant shall be functionally tested either in place or in a bench test.

For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.9f.

an additional 10% of that type of snubber shall be functionally tested until no more failures are found or unti 1 all snubbers of that type have been functionally tested or (2)

A representative sample of each type of snubber shall be functionally tested in accordance with Figure 4.7-1.

"C" is the total number of snubbers of a type found not meeting the acceptance requirements of Specification 4.7.9f.

The cumulative-number of snubbers of a type tested is denoted by "N."

At the end of each day's testing, the new values of "N" and "C" (previous day' total plus current day's increments) shall be plotted on Figure 4.7-1

~

If at any time the point plotted falls in the ".Reject" region, all ST.

LUC IE - UNIT 2 3/4 7-22 Amendment No. 22~ 45

PLANT SYSTEHS SURVEILLANCE RE UIREHENTS

4. 7. 11. 1 a.

The fire suppression water system shall be demonstrated OPERABLE:

At least once per 7 days by verifying the contained water supply volume.

b.

At least once per 31 days on a

STAGGERED TEST BASIS by starting each electric motor driven pump and operating it for at least 15 minutes on.recirculation flow.

C.

At least once per 31 days by verifying that each valve (manual, power-operated or automatic) in the flow path is in its correct position.

d.

At least once per 12 months by performance of a system flush.

At least once per 12 months'y cycling each testable valve in the flow path through at least one complete cycle of full travel.

At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating

sequence, and:

l.

Verifying that each automatic valve in the flow path actuates to its correct position, 2.

Verifying that each pump develops at least 2350 gpm at a system head of 232 feet, 3.

Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full

travel, and 4.

Verifying that each fire suppression pump starts (sequentially) to maintain the fire suppression water system pressure greater than or equal to 85 psig.

At, least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section ll of.the Fire Protection

Handbook, 14th Edition, published by the National Fire Protection Association.

ST.

LUCIE - UNIT 2 3/4 7-31

PLANT SYSTEMS SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION

3. 7.11. 2 The following sprinkler systems shal 1

be OPERABLE:

1.

2.

3

~

4.

5.

6.

7.

8.

9.

Fire Zone 8 - Diesel Generator Building 2A Fire Zone 9 - Diesel Generator Building 2B Fire Zone 19 "

RAB East Hallway and Miscellaneous Equipment Areas Fire Zone 20 -

RAB East-West Common Hallway Fire Zone 22 -

RAB Electrical Penetration Area Fire Zone 23 -

RAB Electrical Penetration Area Fire Zone 39 -

RAB HVAC Equipment Room

'Fire Zone 51 -

RAB Ceiling and Hallways Fire Zone 52 - Cable Spreading Room APPLICABILITY:

Whenever equipment protected by the sprinkler system is required to be OPERABLE.

ACTION:

With one or more of the above required sprinkler systems inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4'. 11.2 Each of the above required sprinkler systems shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow. path is in its correct position.

b.

At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

ST.

LUCIE - UNIT 2 3/4 7-32 Amendment No. 45

I Er r

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS C.

d.

At least once per 18 months by:

1.

Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station.

2.

Removing the hose for inspection and re-racking, and 3.

Inspecting all gaskets and replacing any degraded gaskets in the couplings.

For all fire hose stations not located in the turbine building, at least once per 3 years by:

1.

Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.

2.

Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater, ST

~

LUGIE - UNIT 2 3/4 7-35

TABLE 3.7-4

'FIRE HOSE STATIONS 6.

8.

9.

10.

11.

12.

43'ft level cable spreading room 43 ft level southwest corner of "B" switchgear room near door 19.5 ft level east end of east-west hall 19.5 ft level middle of east-west hall 19.5 ft level south end of north-south hall 19.5 ft level entrance hall on south wall

-0.5 ft level east end of hall 13.

-0.5 ft level south wall of hall 14.

-0.5 ft level west end of hall C.

Hose Stations (Reactor Containment Building) 1.

RCB at 23 ft level (near stairway no.

3) 2.

RCB at 45 ft level (near stairway no.

1) 3.

RCB at 45 ft level (near stairway no.

2) 4.

RCB at 62 ft level (near stairway no.

3)

D.

Hose Station (Fuel Handling Building) 62 ft level northwest corner LOCATION/ELEVATION A.

Hose Stations (Turbine Building) 1.

Operating Floor (northeast corner) 2.

Operating Floor (southeast corner) 3.

Operating Floor (middle east side)

B.

Hose Stations (Reactor Auxiliary Building) 1.

62 ft level east wall entrance 2.

62 ft level west wall entrance 3.

62 ft level west wall entrance to H&V room 4.

43 ft level S.E. corner cable spreading room 5.

43 ft level south wall of H8V room HOSE RACK ¹ HS-15-4 HS-].5-10 HS" 15-7 HS-15-44 HS-15-45 HS-15-46 HS-15" 36 HS-15-37 HS-15-31 HS-15-42 HS-15-38 HS-15-40 HS-15-33 HS-15-34 HS" 15-41 HS-15-28 HS-15-43 HS-15-47 HS-15-48 HS-15-54 HS-15-49 HS-15-55 ST.

LUCIE - UNIT 2 3/4 7"36 Amendment No. 45

k I

h

~ y I

~ PLANT SYSTEMS YARO FIRE HYORANTS AND HYORANT HOSE HOUSES LIMITING CONOITION FOR OPERATION 3.7. 11.4 The yard fire hydrants and associated hydrant hose houses shown in Table 3.7-5 shall be OPERABLE.

APPLICABILITY:

Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE.

ACTION:

'a ~

With one or more of the yard fire hydrants or associated hydrant hose houses shown in Table 3.7-5 inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> have sufficient additional lengths of 2 1/2 inch diameter, hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area(s) if the inoperable fire hydrant or associated hydrant hose house is the primary means of fire suppression; otherwise provide the additional hose within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7. 11.4 Each of the yard fire hydrants and associated hydrant hose houses shown in Table

3. 7-5 shall be demonstrated OPERABLE:

b.

At least once per 31 days by visual inspection of the hydrant hose house to assure all required equipment is at the hose house.

At least once per 6 months by visually insoecting each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged.

At least once per 12 months by:

1.

Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater.

2.

Inspecting all the gaskets and replacing any degraded gaskets in the couplings.

3.

Performing a flow check of each hydrant to verify its OPERABILITY.

ST.

LUCIE " UNIT 2 3/4 7-37 Amendment No. 45

TABLE 3.7-5 YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES LOCATION RCB RCB CST CCQB OGB(2A)

DGB(2B)

RAB NE NW SE HYDRANT NUMBER FH¹6 FH¹7 FH¹9 FH¹20 FH¹21 FH¹22 FH¹23 Intake Structure SE FH¹25 CST SW FH¹26 ST.

LUCIE - UNIT 2 3/4 7-38

I

~I I

3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12.

1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.

APPLICABILITY: At al 1 times.

ACTION:

a.

With the radiological environmental monitor;:tng program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.8, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b.

With the confirmed" level of radioactivity as the result of plant

. effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit'o the Comission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a

MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3,11.1.2, 3,11.2.2, and 3.11.2.3.

When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) reporting level 1

reporting level 2

When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to A MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications

3. 11.1.2, 3.11.2.2 and 3.11.2.3.

This report shall include the methodology for calculating the cumulative potential dose contributions for the calendar year from radionuclides detected in environmental samples and can be determined in accordance with the methodology and parameters in the ODCM.

This report is not required if the measured level of r adio-

. activity was not the result of plant effluents;

however, in such an

event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c.

With milk or broadleaf vegetation samples unavailable from one or more of the sample locations required by Table

3. 12-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program withi'n 30 days.

The specific A confirmatory reanalysis of the original, a duplicate, or a new sample may be desirable, as appropriate.

The results of the confirmatory analysis shall be completed at the earliest time consistent, with the analysis but in any case within 30 days.,

ST.

LUCIE " UNIT 2 3/4 12-3.

Amendment No, fP, 45

RAOIOLOGIGAL ENVIRONMENTAL MONITORING ACTION:

d.

(Continued) locations from which samples were unavailable may then be deleted from the monitoring program.

Pursuant to Specification 6.9.1.10, identify the cause of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also 'include in the report a revised figure(s) and table for the 00CM reflecting the new location(s).

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 12.

1 The radiological environmental monitoring samples shall be collected pursuant to Table 3. 12-1 from the specific locations given in the table and figure(s) in the

OOCM, and shall be analyzed pursuant to the requirements of Table

3. 12-1 and the detection capabilities required by Table 4. 12-1.

ST.

LUCIE - UNIT 2 3/4 12-2 Amendment No 13

I

'ADIOLOGICALENVIRONMENTAL MONITORING Je

~

3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12 '

A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of cthe nearest milk animal, the nearest residence and the nearest garden" of greater than 50 m

(500 ft ) producing broad leaf vegetation.

APPLICABILITY: At al 1 times.

ACTION:

'a ~

With a land use census identifying a location(s) that yields a

calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, identify the new location(s) in the next Semiannual Radioactive Effluent Release

Report, pursuant to Specification 6.9.1.7.

.With a land 'use census identifying a location(s) that yields a

calculated dose or dose commitment (via the same exposure pathway)

'0K greater than at a location from which samples are currently being obtained in accordance with Specification

3. 12. 1, add the new location(s) to the radiological environmental monitoring program within 30 days.

The sampling location(s),

excluding the control station location, having the lowest calculated dose or dose commitment(s),

via the same exposure

pathway, may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

Pursuant to Specification 6.9.1.7,.

identify the new location(s) in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

'URVEILLANCE'E UIREMENTS

4. 12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best

results, such as by a door-to-door survey, aerial

survey, or by consulting local agriculture authorities.

The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification

6. 9. 1.8.

"Broad leaf vegetation sampling may be performed at the site boundary in each of two different direction sectors with the highest predicted 0/gs in lieu of the garden census.

Specifications for broad leaf vegetation sampling in Table

3. 12-1.4b shall be followed, including analysis of control samples.

ST.

LUGIE - UNIT 2 3/4 12-11 Amendment No. I8,45

RAOIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12. 3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be per formed on radioactive materials supplied as part

'of an Interlaboratory Comparison Program that has been approved by the Commission."

APPLICABILITY: At al 1 times.

ACTION:

With analyses not being performed as required

above, report the corrective actions to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9. 1.8.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9. 1.8.

This conditi'on is satisfied by participation in the Environmental Radioactivity Laboratory Intercomparison Studies program conducted by the Environmental Protection Agency (EPA).

ST.

LUCIE - UNIT 2 3/4 12-12 Amendment No. 45

I

'ADN'INISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor, or during his absence. from the control

room, a

designated individual, shall be responsible for the control room command function.

A management directive to this effect, signed by the Senior Vice President

- Nuclear

, shall be r'eissued to all station personnel on an annual basis.

6. 2 ORGANIZATION ONS ITE AND OFFS ITE ORGANIZATION An onsite and an offsite organization shall be established for unit operation and corporate management.

The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

a.

Lines of authority, responsibility and coamunication shall be established and defined from the highest management levels through intermediate levels to and'ncluding all operating organization positions.

Those relationships shall be documented and updated, as appropriate, in the form of organizational charts.

These organizational charts will be documented in the Topical guality Assurance Report and updated in accordance with 10 CFR 50.54(a)(3).

b.

The, Senior Vice President

- Nuclear shall be responsible for overall plant nuclear safety.

This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

c.

The Plant Manager shall. be responsible for overall unit safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant.

r

d.,Although the individuals who train the operating staff and those who carry out the quality assurance functions may report to the appropriate manager onsite, they shall have sufficient organiza-tional freedom to be independent from operating pressures.

e.

Although health physics individuals may report to any appropriate manager onsite, for matters relating to radiological health and safety of employees and the public, the health physics manager shall have direct access to that onsi te individual having responsibility for overall unit management.

Health physics personnel shall have the authority to cease any work activity when worker safety is jeopardized or in the event of unnecessary personnel radiation exposures.

ST.

LUC IE - UNIT 2 6-1 Amendment No.

Jg

AD4lIVISTRATIVE CONTROLS

6. 2 ORGANIZATION (Continued)

UNIT STAFF The unit organization shall be subject to the following:

a..Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor.

In addition, while the reactor is'in MODE 1, 2, 3, or 4, at least one licensed Senior Reactor Operator shall be in the control room.

C ~

d.

A health physics technician shall be on site when fuel is in the reactor.

All CORE ALTERATIONS shall be observed by a license'd operator and supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

The SRO'in charge of fuel handling normally supervises from the control room and has the flexi-bility to directly supervise at either the refueling deck or the spent fuel pool.

e.

A site Fire Brigade ~f at least five members shall be maintained onsite at all times.

The Fire Brigade shall not include the Shift Supervisor, the STA, nor the two other members of the minimum shift crew necessary for 'safe shutdown of the unit and any personnel required for other essential functions during a.fire emergency.

Administrative procedures shall'be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g.,

senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

Adequate shift coverage shall be maintained without routine heavy use of overtime.

The objective shall be to have operating personnel work a normal 8-hour day, 4Q-hour week while, the plant is operating.

However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modification, on a temporary basis the following guidelines shall be followed:

dThe health physics technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected

absence, provided immediate action is taken to fill the required positions.

ST.

LUCIE - UNIT 2 6-2 Amendment No. 2g

I gi

~

Table 6.2-1 MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH TWO SEPARATE CONTROL ROOMS WITH UNIT 1 IN MODE 5 OR 6 OR OEFUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION SS (SRO)

SRO RO AO STA MODE 1, 2, 3, or 4 1a 1

2 2

1 MODE 5 or 6 a

None lb 2

None POSITION WITH UNIT 1 IN MODE 1, 2, 3

OR 4 NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION SS (SRO)

SRO RO AO STA MODE 1, 2, 3, or 4 la 1

2 2

1a MODE 5 or 6 1a None 1

1 None SS Shift Supervisor with a Senior Reactor Operator-'s License on Unit 2 SRO Individual with a Senioi Reactor Operator's License on Unit 2 RO Individual with a Reactor Operator's License on Unit 2 AO Auxiliary Operator STA Shift Technical Advisor Except for the Shift Supervisor, the Shift Crew Composition may be one less than the minimum requirements of Table 6: 2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid SRO license shall be'esignated to assume the Control Room command function.

During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 5 or 6, an individual with a valid SRO or RO license shall be designated to assume the Control Room command function

\\

a/

Individual may fi 1 1 the same pos ition on Unit 1 b/

One of the two required individuals may fill the same position on Unit 1, ST.

LUGIE - UNIT 2 6-5

ADMINISTRATI VE CONTROLS J

4 0

6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP ISEG FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics, NRC issuances, industry advisories, Licensee Event Reports and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improving plant safety.

COMPOS ITION 6.2.3.2 The ISEG shall be composed of five dedicated, full-time members with varied backgrounds and disciplines related to nuclear power-plants.

No more than two members shall be assigned from any one department.

Three or more of the members shall be engineers with a Bachelor's degree in engineering or a

related science, with at least 2 years of professional level experience in the nuclear field.

Any nondegreed ISEG members will either be licensed as a Reactor Operator or Senior Reactor Operator, or will have been previously licensed as a

Reactor Operator or Senior Reactor Operator within the last year at the St. Lucie Plant site; or they will meet the qualifications of a department head as specified in Specification 6.3.1 of the St. Lucie Unit 2 Technical Specifications.

The qualifications of each nondegreed candidate for the ISEG shall be approved by the Site Vice President

- St. Lucie Plant, prior to joining the groap.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of selected plant activities to provide independent verification* that these activities are performed correctly and that human errors are reduced as much as practical.

The ISEG shall make detailed recommendations for revised procedures, equipment modi-fications, maintenance activities, operations activities, or other means of improving plant safety to the Site Vice President

- St. Lucie Plant.

AUTHORITY 6.2.3.4 The ISEG is an onsite independent technical review group that reports to the Site Vice President

- St. Lucie Plant.

The ISEG shalT have the authority necessary to perform the. functions and responsibilities as delineated above.

RECORDS 6.2.3.5 Records of activities performed by the ISEG shall be prepared, main-tained and a report of the activi ti es forwarded each calendar month to the Site Vice President

- St.

L'ucie Plant.

6.2.4 SHIFT TECHNICAL ADVISOR The Shift Technical Advisor function is to provide on shift advisory technical support in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI/ANS-3.1-1978 as endorsed by Regulatory Guide 1.8, September 1975 (reissued tiay 1977),

except for the (1) Health Physics Supervisor who shall meet "Not responsible for sign-off function, ST

~

LUCIE - UNIT 2 6-6 Amendment No. f,45

ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) m.

AUTHORITY Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Senior Vice President - Nuclear and to the Company Nuclear Review Board.

Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL and RADWASTE TREATMENT SYSTEMS.

Review and documentation of judgment concerning prolonged operation in bypass; channel trip, and/or repair of defective protection channels of process variables placed in bypass since the last FRG meeting.

6..5.1.7 The Facility Review Group shall:

a.

Recommend in writing to the Plant Manager approval or disapproval of items considered under Specifications 6.5.1.6a.

through d.

and m.

above.

b.

Render determinations in writing with regard to whether or not each item considered under Specifications 6.5.1.6a.

through e.

above constitutes an unreviewed safety question.

c.

Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Senior Vice President - Nuclear'nd the Company Nuclear Review Board of disagreement between the FRG and the Plant Manager;

however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to SpeCification 6.1.1 above.

RECORDS 6.5.1:8 The Facility Review Group shall maintain written minutes, of each FRG meeting that, at a minimum, document the results of all FRG activities performed under the responsibility-and authority provisions of these technical specifica-tions.

Copies shall be provided to the Senior Vice President - Nuclear and the Chairman of the Company Nuclear Review Board.

6.5.2 COMPANY NUCLEAR REVIEW BOARD CNRB)

FUNCTION 6.5.2.1 The Company Nuclear Review Board shall function to provide independent review and audit of designated activities in the areas of:

a.

b.

C.

d.

nuclear power plant operations nuclear engineering chemistry and radiochemistry metallurgy ST.

LUCIE - UNIT 2 6-9 Amendment No.49, 29

I

~

~ e FUNCTION (Continued) e.

ge h.

COMPOSITION i ns trumenta tion and control radi ol ogi ca 1 sa fety mechanical and electrical engineering quality assurance practices 6.5.2.2 The Executive Ilice President shall appoint, in writing, a minimum of five members to the CNRB and shall designate from this membership, in writing, a Chairman.

The membership shall function to provide independent review and audit in the areas listed in Specification 6.5.2.1.

The Chairman shall meet the requirements of ANSI/ANS-3.1-1987, Section 4.7.1.

The members of the CNRB shall meet the educational requirements of the ANSI/ANS-3.1-1987, Section 4.7.2,, and have at least 5 years of professional level experience in one or more of the fields listed in Specification 6.5.2.1.

CNRB members who do not possess the educational

'requirements of ANSI/ANS-3.1-1987, Section 4.7.2 (up to a maximum of 2'embers) shall be evaluated, and have their membership approved and documented, in writing, on a case-by-case basis by the Executive Vice President; considering the alternatives to educational requirements of ANSI/ANS-3.1-1987, Sections 4.1.1 and 4.1.2.

ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the CNRB Chairman to serve on a temporary basis;

however, no more than two alternates shall participate as voting members in CNRB activities at any one time.

CONSULTANTS 6.5.2.4 Consultants shall'e utilized as determined by the CNRB Chairman to provide expert advice to the CNRB.

ME'ETING FR'E UENCY 6.5.2.5 The CNRB shall meet at, least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter and as convened by the CNRB Chairman or his designated alternate.

QUORUM 6.5.2.5

,he quorum of the CNRB necessary for review and audit functions of these Technical the Chairman or. his designated alternate and

.members including alternates.

No more than a

have line responsibility

.or operation of the the performance of the CNRB Specifications shall.consist of at least a majority of CNRB minority of the quorom shall unit.

ST, LUC,IE - UNIT 2 6-10 Amendment No.

g$, M/>> 88, 45

~

~

~

AOMINIST TIVE CONTROLS

6. 13 PROCESS CONTROL PROGRAM PCP Licensee initiated changes to the PCP:

l.

Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made.

This submittal shall contain:

a.

Sufficiently detailed information to totally suport the rationale for the change without benefit of additional or supplemental information; b.

A determination that the change did not reduce the overall conformance of the dewatered bead resin to existing criteria for radioactive wastes; and c.

Oocumentation of the fact that the change has been reviewed and found acceptable by the FRG.

2.

Shall become effective upon review and acceptance by the FRG.

6. 14 OFFSITE OOSE. CALCULATION MANUAL OOCM Licensee initiated changes to the ODCM:

Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made effective.

This submittal shall contain:

a.

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supple" mental information.

Information submitted should consist of a package of those pages of the OOCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s);

b.

A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c,

Oocumentation of the fact that the change has been reviewed and found acceptable by the FRG.

2.

Shall become effective upon review and acceptance by the FRG.

ST ~

LUCIE - UNIT 2 6-23 Amendment No, ~~ '~

45

ADMINISTRATIVE CONTROLS

6. 15 MAJOR CHANGES TO RADIOACTIVE LI UID GASEOUS AND SOLID WASTE TREATMENT SYSTEMS

6. 15. 1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):

Shall be >epor ted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Facility Review Group.

The discussion of each shall contain:

a.

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59.

b.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; c.

A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; d.

An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto,;

An evaluation of the change which shows the expected maximum exposures to individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; f.

A comparison of the predicted releases of radioactive materials, in liquid'nd gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; g.

An estimate of the exposure to plant oper ating personnel as a

result of the change; and h.

Documentation of the fact that the change was reviewed and found acceptable by the FRY 2.

Shall become effective upon review and acceptance by the FRG.

Licensees may chose to submit the information called for in this Specification as part of the annual FSAR update.

ST.

LUCIE - UNIT 2 6

24 Amendment No.