Text

Amends 152 & 89 to Licenses DPR-67 & NPF-16,respectively, Consisting of Changes to TS in Response to 970529 Application Re Removing Outdated Matl & Incorporating Minor Changes in Text
	ML17229A481
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Site:	Saint Lucie
Issue date:	09/22/1997
From:	Hebdon F; NRC (Affiliation Not Assigned)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20M')001 FLORIDA POWER

& LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.152 License No.

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

A.

The application for amendment by Florida Power

& Light Company, (the licensee),

dated May 29.

1997. complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.

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.

g (f02 OI@ I

'77i0020i59 970922 PDR ADQCK 05000335 P

PDR 2.

3 ~

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 Amendment No. 152, are hereby incorporated in the license.

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

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 22, 1997 Frederick J.

Heb on, Director Project Directorate II-3 Division of Reactor Projects

- I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT-NO. 152 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 VII B 2-6 B 2-8 3/4 1-1 3/4 1-4 3/4 1-8 3/4 1-12 3/4 1-21 3/4 1-22 3/4 2-9 3/4 4-21 3/4 6-3 3/4 6-20 3/4 6-26 3/4 9-4 3/4 9-6 3/4 10-2 3/4 11-14 B 3/4 1-3 B 3/4 4-15 B 3/4 9-1 5-4 Insert Pa es VII.

B 2-6 B 2-8 3/4 1-1 3/4 1-4 3/4 1-8 3/4 1-12 3/4 1-21 3/4 1-22 3/4 2-9 3/4 4-21 3/4 6-3 3/4 6-20 3/4 6-26 3/4 9-4 3/4 9-6 3/4 10-2 3/4 11-14 B 3/4 1-3 B 3/4 4-15 B 3/4 9-1 5-4

(Continued)

St Ul~

E to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam fiow. This setting was used with an unceitainty factor of a 22 psi in the accident analyses.

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 willnot be exceeded due to loss of steam generator heat sink The specNied setpoint provides allowance that there willbe sufficient water inventoiy in the steam generators at the time of trip to provide sufficient time forany operator action to initiate auxiiiaiy feedwater before reactor coolant system subcooling is lost.

I wr ni-The local Power Density-High trip, functioning from AXIALSHAPE INDEXmonitoring, is provided to ensure that the peak local power density ln the fuel which corresponds to fuel centeriine melting will not occur as a consequence of axial power maldistiibutions. A reactor trip is initiated whenever the AXIALSHAPE INDEXexceeds the allowable limits of Figure 2.2-2. The AXIALSHAPE INDEXis calculated from the upper and lower ex~re neutron detector channels.

The calculated setpoints are generated as a function of THERMALPOWER level with the allowed CEA group position being inferred from the THERMALPOWER level, The trip is automatically bypassed below 15 percent powei; The maximum AZIMUTHALPOWER TILTand maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints.

In addition, CEA group sequencing in accordance with the Spec Tiications 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 occurrence prior to a Power Level-High trip is assumed.

ST. LUCIE-UNIT1 82%

Amendment No. W,152

A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMALPOWER. This trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the liftingof the main steam line safety valves during the ensuing transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System, 0

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 administratively enforced startup rate limit. The trip is not credited in any design basis accident evaluated in UFSAR Chapter 15; however, the trip is considered in the safety analysis in that the presence of this trip function precluded the need forspecific analyses of other events initiated from subcritical conditions.

ST. LUCIE-UNIT1 B24 Amendment No.43, ] 5o

/

3.1.1.1 The SHUTDOWN MARGINshall be ~ 3600 pcm.

JOSEEOE:

IIOOESI, ',

44.

hGIIQH:

With the SHUTDOWN MARGIN< 3600 pcm, immediately initiate and continue boration at a 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGINis restored, 4.1.1.1.1 The SHUTDOWN MARGINshall be determined to be x 3600 pcm:

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

Ifthe inoperable CEA is not fullyinserted, and is immovable as a result of excessive friction or mechanical interference or is known to be untrippable, the above required SHUTDOWN MARGINshall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).

b.

When In MODES1 or2, at least once per12 hours byverifying that CEA group withdrawal is within the Power Dependent Insertion Umits of Specification 3.1.3.6.

c.

When in MODE 2 at least once during CEA withdrawal and at least once per hour thereafter until the reactor is critical.

d.

Prior to initialoperation above 5% RATED THERMALPOWER after each fuel loading, by consideration of the factors of e below, with the CEA groups at the Power Dependent Insertion Umits of SpecTiication 3.1.3.6.

See Special Test Exception 3.10.1.

¹ With Keff w 1.0.

¹¹ With Keff< 1.0.

ST. LUCIE-UNIT1 8/4 1-1 Amendment No. H',46,68,88, 152

3.1.1.3 The flowrate of reactor coolant to the reactor pressure vessel shall be a 3000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.

ALLLIDDED.

hQllQhl:

With the flowrate of reactor coolant to the reactor pressure vessel < 3000 gpm, immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System, 4.1.1.3 The flowrate of reactor coolant to the reactor pressure vessel shall be determined to be a 3000 gpm within one hour prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either.

a.

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

Verifyingthat at least one low pressure safety injection pump is in operation and supplying a 3000 gpm to the reactor pressure vessel.

ST. LUCIE-UNIT 1 3/41M Amendment No. 152

3.1.2.1 As a minimum, one of the following boron injection flowpaths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

a.

Aflowpath from the boric acid makeup tank via either a boric acid pump or a gravity feed connection and any charging pump to the Reactor Coolant System If only the boric acid makeup tank in SpecTiication 3.1 2.7a is OPERABLE, or b.

The flowpath from the refueling water tank via either a charging pump or a high pressure safety injection pump* to the Reactor Coolant System ifonly the refueling water tank in SpeciTication 3.1.2,7b is OPERABLE.

MODES5and6.

hQIIQbl:

With none of the above flowpaths OPERABLE, suspend all operations involving CORE ALTERATIONSor positive reactivity changes until at least one Injection path is restored to OPERABLE status.

4.1.2.1 At least one of the above required flowpaths shall be demonstrated OPERABLE:

a.

A least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flowpath that is not locked, sealed, or otherwise secured in position, is fn its correct position.

The flowpath fiothe RWTto the RCS via a single HPSI pump shall only be establishedif:

(a) the RCS pressure boundaty does not exist, or (b) RCS pressure boundary integrity exists and no charging pumps are operable.

In the latter case:

1) all charging pumps shall be disabled;

2) heatup and cooidown rates shall be limited In accordance with Figure 3.1-1b; and 3) at RCS temperatures below 115'F, any two of the followingvalves in the operable HPSI header shall be verifled dosed and have their power removed:

HCV4616 HCV<626 HCV<636 HCV4646 HCV<617 HCV<627 HCV<637 HCVQ647 ST. LUCIE-UNIT1 SI4 14 Amendment No. 69, fH,99, 94,468,%H,+H 152

3.12.3 At least one charging pump or high pressure safety injection pump'n the boron injection flowpath required OPERABLE pursuant to Specification 3.1Z.1 shall be OPERABLE and capable of being powered from an OPERABLE emetgency bus.

MODES 5 and 6.

hQXKRf:

With no charging pump or high pressure safety injection pump* OPERABLE, suspend all operations involving CORE ALTERATIONSor positive reactivity changes until at least one of the required pumps Is restored to OPERABLE status.

4.12.3 At least one of the above required pumps shall be demonstrated OPERABLE by verifying the charging pump develops a flow rate of greater than or equal to 40 gpm or the high pressure safety injection pump develops a total head of greater than or equal to 2571 ft.

when tested pursuant to Specification 4.0.5.

The flowpath from the RWT to the RCS via a single HPSI pump shall be established only If:

(a) the RCS pressure boundaty does not exist, oi (b) RCS pressure boundaty integrity exists and no charging pumps are operable.

In the latter case:

1) ail charging pumps shall be disabled;

2) heatup and cooldown rates shall be limited in accordance with Figure 3.1-1b; and 3) at RCS temperatures below 115'F, any two of the followingvalves in the operable HPSI header shall be verwed closed and have their power removed:

HCVQ616 HCV4626 HCV4636 HCV<646 HCV<617 HCV<627 HCV<637 HCV<647 ST. LUCIE-UNIT1 3/4 1-12 Amendment No. 69, N; 99, 494,+%,~

1 52

2.

Declared inoperable and satisfy SHUTDOWN MARGINrequirements of Specification 3.1.1.1. Afterdeclaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 for up to 7 days per occurrence with a total accumulated time of ~ 14 days per calendar year provided ail of the following conditions are met:

a)

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the CEAs in the group with the inoperable CEA shall be aligned to within 7.5 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on COLR Figure 3.1-2; the THERMALPOWER level shall be restricted pursuant to SpeciTication 3.1.3,6 during subsequent operation.

b)

The SHUTDOWN MARGINrequirement of SpecNcation 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , Otherwise, be in at least HOT STANDBYwithin the next 6 houts.

e, With one fulllength CEA misaligned from any other CEA in its group by15 or more inches, operation in MODES 1 and 2 may continue provided that the misaligned CEA is positioned within 7.5 inches of other CEAs in its group in accordance with the time constraints shown in COLR Figure 3.1-1 a.

f.

With one full-length CEA misaligned from any other CEA in its group by 15 or more inches beyond the time constraints shown in COLR Figure 3.1-1 a, reduce power to c 70% of RATED THERMALPOWER prior to completing ACTIONf.1 or fZ.

Restore the CEA to OPERABLE status within its specNed alignment requirements, or 2.

Declare the CEA inoperable and satisfy the SHUTDOWN MARGIN requirements of Spec Tiication 3.1.1.1. Afterdeclaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided:

a)

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the CEAs In the group with the inoperable CEA shall be aligned to within 7.5 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on COLR Figure 3.1-2; the THERMALPOWER level shall be restricted pursuant to Spec Tiication 3.1.3.6 during subsequent

. operation.

ST. LUCIE-UNIT1

$4 1-21 Amendment No. &,~ 152

APO N

b)

The SHUTDOWN MARGINrequirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Otherwise, be In at least HOT STANDBYwithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

g.

With more than one fulllength CEA inoperable or misaligned from any other CEA in its group by 15 inches (indicated position) or more, be in HOT STANDBYwithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

h.

With one full-length CEA inoperable due to causes other than addressed by ACTIONa above, and inserted beyond the long term steady state insertion limits but within its above specified alignment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6.

4.1.3.1.1 The position of each fulWength CEA shall be determined to be within7.5 inches (indicated position) of all other CEAs in its group at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time Inteivals when the Deviation Circuit and/or CEA Block Circuit are inoperable, then verify the indMdual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.12 Each fulllength CEA not fullyInserted shall be determined to be OPERABLE by inserting it at least 7.5 inches at least once per 92 days.

4.1.3.1.3 The CEA Block Circuit shall be demonstrated OPERABLE at least once per 92 days by a functional test which verifies that the circuit prevents any CEA from being misaiigned from all other CEAs in its group by more than 7.5 inches (indicated position).

4,1.3.1.4 The CEA Block Circuit shall be demonstrated OPERABLE by a functional test which verifies that the circuit maintains the CEA group overlap and sequencing requirements of Specification 3.1.3.6 and that the circuit prevents the regulating CEAs from being inserted beyond the Power Dependent Insertion Limitof COLR Figure 3.1-2:

a.

Prior to each entty into MODE2 from MODE 3, except that such verification need not be performed more often than once per 92 days, and

~

g I

b.

Atleast once per 6 months.

The licensee shall be excepted from compliance during the startup test program for an entiy into MODE 2 from MODE 3 made in association with a measurement of power defect.

ST. LUCIE-UNIT 1 3/4 1-22 Amendment No. &,98,W,

+P,469, 1 52

RDS FT 32.3 The calculated value of F, shall be within the limits specTiied in the COLR.

PEUMIIUM MODE1'.

~CI(+:

With FT not within limits, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

a.

Be in at least HOT STANDBY,or b.

Reduce THERMALPOWER to bring the combination of THERMALPOWER and Ff to within the limits of COLR Figure 3.2-3 and withdraw the full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THERMALPOWER limitdetermined from COLR Figure 3,2-3 shall then be used to establish a revised upper THERMALPOWER level limiton COLR Figure 3.2A (truncate Figure 3.2-4 at the allowable fraction of RATED THERMALPOWER determined by COLR Figure 3,2-3) and subsequent operation shall be maintained within the reduced acceptable operation region of COLR Figure 3.2A.

42,3.1 4.2,3.2 The provisions of SpecTiication 4.0.4 are not applicable.

Fs shatt be catcutatedby the exPresstcn FJ =F,(1+ Ts) when F, ts calculated with a non-full core power distribution analysis code and shall be calculated as F, = F, when calculations are performed with a fullcore power distribution analysis code.

F, shall be determined to be within its limitat the following intervals.

a.

Prior to operation above 70 percent of RATEDTHERMALPOWER after each fuel loading.

b.

Atleast once per 31 days of accumulated operation in MODE 1, and c.

Within four hours ifthe AZIMUTHALPOWER TILT(Tq) is > 0.03.

See Special Test Exception 3.10.2, ST. LUCIE-UNlT1 3/42-9 Amendment No. 97, 82, 48, 65,

~,152

3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limitlines shown on Figures 3.4-2a, 3.4-2b and 3.4N during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.

Atalltimes. ¹ hQllQH:

With any of the above limits exceeded, restore the temperature and/or pressure to withinthe limits within 30 minutes; perform an analysis to determine the effects of the outwf-limitcondition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T<< to less than 200'F within the following30 hours in accordance with Figures 3A-2b and 3,4.

When the flowpath from the RWT to the RCS via a single HPSI pump is established per 3.12.1 or 3.1.2.3 and RCS pressure boundary integrity exists, the heatup and cooldown rates shall be established in accordance with Fig.3.1-1b.

¹ During hydrostatic testing operations above system design pressure, a maximum temperature change in any one hour period shall be limited to 5'F.

ST. LUCIE-UNIT 1 3J44-21 Amendment No, 4, fH,~, 152

Pages 3/4 &4through 3/4 6-9 have been DELETED.

Page 3/4 6-10 is the next valid page.

ST. LUCIE-UNIT1 3/4 6-3 Amendment No. 88,4&7,+8, ] 5p

Pages 3/4 6-21 throvgh 3/4 6-22 have been DELETED.

Page 3/4 6-23 is the next vaiid page.

ST. LUCIE-UNIT1 3/4 6-20 j

Amendment No. 85, +HI,449, ) 52

I E

V U

3.6.5.1 The containment vessel to annulus vacuum relief valves shall be OPERABLE with an actuation setpoint of 2,25 a 0.25 inches Water Gauge differential.

MODES1,2,3and4.

hQXLQH:

With one containment vessel to annulus vacuum relief valve inoperable, restore the valve to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following30 houts.

4.6.5.1 No additional Surveillance Requirements other than those required by Speclcation 4.0.5 and at least once per 3 years verifythat the vacuum relief valves open fullywithin 8 seconds at 2.25 ~ 0.25 inches Water Gauge differentiaL ST. LUCIE-UNIT1 3/4 6-26 Amendment No. ee, 152

3.9.4 The containment penetrations shall be in the following status:

a.

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

A minimum of one door in each airlock is dosed, and C.

Each penetration 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 except forvalves that are open on an intermittent basis under administrative control, or Be capable of being dosed by an OPERABLE automatic containment isolation valve, or 3.

Be capable of being dosed by an OPERABLE containment vacuum relief

valve, During CORE ALTERATIONSor movement of irradiated fuel within the containment.

hQIIQH:

With the requirements of the above specTiication not satisfied, immediately suspend all operations involving CORE ALTERATIONSor movement of irradiated fuel in the containment. The provisions of SpeciTication 3.0,3 are not applicable, 4.9.4 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of and at least once per 7 days during CORE ALTERATIONSoi movement of irradiated fuel in the containment by:

a.

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

Testing of containment isolation valves per the applicable portions of Specifications 4.6.3.1.1 and 4.6,3.1 Z.

ST. LUCIE-UNIT 1 Amendment No. +7,152

EF 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 oveHoad cut offlimitof ~ 3000 pounds.

During movement of CEAs or fuel assemblies within the reactor pressure vessel.

hQILQH:

With the requirements for crane OPERABILITYnot satisfied, suspend use of any inoperable manipulator crane from operations invoMng the movement of CEAs and fuel assemblies within the reactor pressure vessel.

4.9,6 The manipulator crane used formovement of CEAs or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of such operations by performing a load test of at least 2500 pounds and demonstrating an automatic load cut off before the crane load exceeds 3000 pounds.

ST. LUCIE-UNIT1 Amendment No. 152

3.102 The group height, insertion and power distribution limits of Specifications 3,1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 32.3 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMALPOWER is restricted to the test power plateau which shall not exceed 85% of RATEDTHERMALPOWER, and b.

The limits of SpecTiication 3.2.1 are maintained and determined as specified in Specification 4.1 022 below.

Bl: MODES1and2, hQIIQ5:

With any of the limits of Spec Tiication 3.2.1 being exceeded while the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 32.3 and 3.2.4 are suspended, either.

a, Reduce THERMALPOWER sufficiently to satisfy the requirements of Specification 3.2.1, or b.

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

4.102.1 4.10.22 The THERMALPOWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Spec Tiications 3.1.1.4, 3.1.3.1, 3.1.3,5, 3.1.3.6, 3.2,3, or 3.2,4 are suspended and shall be verifiied to be within the test power plateau, The linear heat rate shall be determined to be within the limits of SpecTiication 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of SpecTiications 4.2.1.4 during PHYSICS TESTS above 5% of RATED THERMALPOWER in which the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 32,3, or 3.2.4 are suspended.

ST. LUCIE-UNIT1 3/4 10-2 Amendment No.~,~, +49...

dies,152

3.112.5 The concentration of oxygen in the waste gas decay tanks shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

Atalltimes.

hGXlQH:

a.

With the concentration of oxygen in the waste gas decay tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b.

With the concentration of oxygen in the waste gas decay tank greater than 4% by volume and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the system and immediately commence reduction of the concentration of oxygen to less than or equal to 2% by volume.

c, The provisions of SpeciTications 3.0.3 and 3.0.4 are not applicable.

4.11.2.5.1 The concentration of oxygen in the waste gas decay tank shall be determined to be within the above limits by continuously monitoring the waste gases in the on seivice waste gas decay tank.

4.112.5.2 With the oxygen concentration in the on service waste gas decay tank greater than 2% by volume as determined by SpecTiication 4.11.2.5.1, the concentration of hydrogen in the waste gas decay tank shall be determined to be within the above limits by gas partitioner sample at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ST. LUCIE-UNIT 1 3f4 11-14 Amendment No. 69, +48, 1 52

T N

fhe boron addition capability after the plant has been placed in MODES 5 and 6 requires either 3650 gallons of 2.5 to 3.5 weight percent boric acid solution (4371 to 6119 ppm boron) from the boric acid tanks or 11,900 gallons of 1720 ppm borated water from the refueling water tank to makeup for contraction of the primaiy coolant that could occur ifthe temperature is lowered from 200'F to 140'F.

The restrictions associated with the establishing of the fiowpath from the RWT to the RCS via a single HPSI pump provide assurance that 10 CFR 50 Appendix G pressure/temperature limits will not be exceeded in the case of any inadvertent pressure transient due to a mass addition to the RCS.

If RCS pressure boundaiy integrity does not exist as defined in Specification 1.16, these restrictions are not required. Additionally, a limiton the maximum number of operable HPSI pumps is not necessaiy when the pressurizer manway cover or the reactor vessel head is removed.

V B The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGINis maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels.

The ACTIONstatements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met.

The ACTIONstatements applicable to an immovable or untrippable CEA and to a large

'misalignment (z 15 inches) of two or more CEAs, require a prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrippable CEA, the loss of SHUTDOWN MARGIN.

For small misalignments (c 15 inches) of the CEAs, there is 1) a small degradation in the peaking factors relative to those assumed in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect on the time dependent long term power distributions relative to those used in generating LCOs and LSSS setpoints for DNBR and linear rate, 3) a small effect on the available SHUTDOWN MARGIN,and 4) a small effect on the ejected CEA worth used in the safety analysis.

Therefore, the ACTION statement associated with the small misalignment of a CEA permits a one hour time Inteival during which attempts may be'made to restore the CEA to within its alignment requirements prior to initiating a reduction in THERMALPOWER, The one hour time limitis sufficient to (1) identify causes of a misaligned CEA, (2) take appropriate corrective action to realign the CEAs, and (3) minimize the effects of xenon redistribution.

Overpower margin is provided to protect the core in the event of a large misalignment (a 15 Inches) of a CEA. However, this misalignment would cause distortion of the core power distribution. This distribution may, in turn, have a sign Tiicant effect on (1) the available SHUTDOWN MARGIN,(2) the time4ependent long-term power distributions relative to those used in generating LCOs and LSSS setpoints, and (3) the ejected CEA worth used in the safety analysis. Therefore, the ACTIONstatement associated with the large misalignment of the CEA requires a prompt realignment of the misaligned CEA, ST. LUCIE-UNIT1 B 3/41-3 Amendment No. 9F, W, fH,94152

I

p-The low temperature overpressure protection system (LTOP) is designed to prevent RCS overpressurization above the 10 CFR 50 Appendix G operating limitcurves (Figures 3A-2a and 3,4-2b) at RCS temperatures at or below 304'F during heatup and 281 'F during cooldown. The LTOP system is based on the use of the pressurizer powerwperated relief valves (PORVs) and the implementation of administrative and operational controls.

The PORVs aligned to the RCS with the low pressure setpoints of 350 and 530 psia, restrictions on RCP starts, limitations on heatup and cooldown rates, and disabling of nonessential components provide assurance that Appendix G Pfl limits willnot be exceeded during normal operation or design basis overpressurization events due to mass or energy addition to the RCS. The LTOP system APPLICABILITY,ACTIONS, and SURVEILLANCEREQUIREMENTS are consistent with the resolution of Generic Issue 94, 'Additional Low-Temperature Overpressure Protection for Ught-Water Reactors,'ursuant to Generic Letter 9046.

Reactor Coolant System vents are provided to exhaust noncondensible gases ancVor steam ftam the primary system that could inhibit natural circulation core cooling. The OPERABILITYof at least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function.

The redundancy design of the Reactor Coolant System vent systems serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant System vent system are consistent with the requirements of Item Il.b.1 of NUREG%737, 'ClariTication of TMIAction Plan Requirements,'ovember 1980.

ST, LUCIE-UNIT1 B 3/44-15 Amendment No. 68, es, EH; 464,482,152

The limitation on minimum boron concentration ensures that: 1) the reactor willremain subcritical during CORE ALTERATIONS,and 2) a uniform boron concentration is maintained for reactivity control in the water volumes having direct access to the reactor vessel.

The limitation on Kis sufficient to prevent reactor criticalitywith all fulllength rods (shutdown and regulating) fullywithdrawn.

The OPERABILITYof the wide range logarithmic range neutron fluxmonitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

The requirements on containment penetration closure and OPERABILITYensure that a release of radioactive material within containment willbe restricted from leakage to the environment. The OPERABILITYand closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

ln accordance with Generic Letter 91%8, Removal of Component LIsts from the Technical Spec Tiications, the opening of locked or sealed dosed containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions willnot preclude access to dose the valves and that this action willprevent the release of radioactivity outside the containment.

The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facilitystatus or core reactivity condition during CORE ALTERATIONS.

0 The OPERABILITYrequirements of the cranes used for movement of fuel assemblies ensures that:

1) each crane has sufficient load capacity to lifta fuel element, and 2) the core internals and pressure vessel are protected from excessive liftingforce in the event they are inadvertently engaged during liftingoperations.

ST. LUCIE-UNIT1 B 3/4 9-1 Amendment No. 68,46e, 3 52

a.

Minimum annular space = 4 feet b.

Annulus nominal volume = 543,000 cubic feet c.

Nominal outside height (measured from top of foundation base to the top of the dome) = 230.5 feet d.

Nominal inside diameter = 148 feet e.

Cylinder wall minimum thickness = 3 feet f.

Dome minimum thickness = 2,5 feet g.

Dome inside radius -112 feet 5.2.2 The containment vessel is designed and shall be maintained fora maximum internal pressure of 44 psig and a temperature of 264'F.

52.3 Penetrations through the containment structure are designed and shall be maintained in accordance with the original design provisions contained in Sections 3.8.2.1.10 and 6.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.3 5.3.1 5.3.1.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 176 fuel rods clad with ZircaIoy4. Each fuel rod shall have a nominal active fuel length of between 134.1 and 136.7 inches.

Individual fuel assemblies shall contain fuel rods of the same nominal active fuel length. Fuel assemblies shall be limited to those designs that have been analyzed using NRC approved methodology and shown by tests or analyses to comply with fuel design and safety criteria. The initialcore loading shall have a maximum enrichment of 2.83 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading.

Except for special test as authorized by the NRC, all fuel assemblies under control element assemblies shall be sleeved with a sleeve design previously approved by the NRC.

ST. LUCIE-UNlT 1 Amendment No. 84,44, V6, 444> f 5p

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UNITED STATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, D.C. 205554001 FLORIDA POWER 8 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. 89 License No.

NPF-16 The Nuclear Regulatory Commission (the Commission) has found that:

A.

B.

C.

D.

E.

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

dated May 29.

1997. complies with the standards and requirements of the Atomic Energy Act of 1954.

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

2.

3.

Accordingly. Facility Operating License No. NPF-16 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 Amendment No. 89

, are hereby incorporated in the license.

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

This license amendment is effective as of its date of'ssuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY CONHISSION Frederick J.

Hebdon. Director Project Directorate II-3 Division of Reactor Projects

- I/II, Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

September pp, igg7

ATTACHMENT TO LICENSE AMENDMENT NO.8>

TO FACILITY 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 amendment number and contain vertical lines indicating the area of change.

The corresponding over leaf pages are also provided to maintain document completeness.

Remove Pa es XVIII XIX B 2-1 B 2-4 B 2-5 B 2-6 3/4 1-1 3/4 1-19 3/4 2-14 3/4 3-26 3/4 6-3 3/4 6-21 3/4 6-26 3/4 8-7 3/4 9-6 3/4 11-14 B 3/4 2-2 Insert Pa es XVIII XIX B 2-1 B 2-4 B 2-5 B 2-6 3/4 1-1 3/4 1-19 3/4 2-14 3/4 3-26 3/4 6-3 3/4 6-21 3/4 6-26 3/4 8-7 3/4 9-6 3/4 11-14 B 3/4 2-2

EAT 0

The restrictions of this safety limitprevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centertine fuel melting willoccur. 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 ccelant saturation temperature.

Operation above the upper boundaiy of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant shay reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMALPOWER 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 fluxand the location of DNBforaxially uniform and non-uniform heat flux distributions, The local DNB heat flux ratio, DNBR, defined as the ratio of the heat fluxthat would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to the DNB-SAFDLof 1.28 in conjunction with the Extended'tatistical Combination of Uncertainties (ESCU). 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 willnot occur and is chosen as an appropriate margin to DNBfor all operating conditions.

The curves of Figure 2.1-1 show conseivative loci of points of THERMALPOWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating forwhich the DNB-SAFDLis not violated forthe 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 limitthe maximum value of reactor inlet temperature.

Reactor operation at THERMALPOWER levels higher than 112% of RATED THERMALPOWER is prohibited by the high power level trip setpoint speciTied in Table 2.1-1. The area of safe operation is below and to the left of these lines.

The conditions forthe Thermal Margin Safety Limitcurves 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 Liming Conditions for Operation, the Variable Overpower Trip and the Power Dependent Insertion Limits, assure that the Spec Tiied Acceptable Fuel Design Limits on DNB and Fuel Centerline Melt are not exceeded during normal operation and design basis Anticipated Operational Occurrences, ST. LUCIE-UNIT2 B 2-1 Amendment No. 8,45

v A Reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions which are too rapid to be protected by a Pressurizer Pressure - High or Thermal Margin/Low Pressure Trip.

The Variable Power Level High trip setpoint is operator adjustable and can be set no higher than 9.61% above the indicated THERMALPOWER level. Operator action is required to increase the trip setpoint as THERMALPOWER is increased, The trip setpoint is automatically decreased as THERMALPOWER decreases.

The trip setpoint has a maximum value of 107,0% of RATED THERMALPOWER and a minimum setpoint of 15.0% of RATED THERMALPOWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMALPOWER level at which a trip would be actuated is 112% of RATED THERMALPOWER, which is the value used in the safety analysis.

ri Pr ur-1 The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam line safety valves, provides Reactor Coolant System protection against oveg)ressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2375 psia which is below the nominal liftsetting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves.

The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than the DNB-SAFDLof 1.28, in conjunction with ESCU methodology.

The trip is initiated whenever the Reactor Coolant System pressure signal drops below either 1900 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher ofbT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIALSHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHALPOWER TILTand the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function, In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.

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

The Thermal Margin/Low Pressure trip setpoints are derived from the core safety limitsthrough application of appropriate allowances for equipment response time, measurement uncertainties and processing error. The allowances include: a variable (power dependent) allowance to compensate forpotential power measbrement error, an allowance to compensate for potential temperature measurement uncertainity; an allowance to compensate for pressure measurement error; and an allowance to compensate forthe time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit.

ST. LUCIE-UNIT2 B24 Amendment No.e,69 89

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.

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, The 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 fiow. This setting was used with an uncertainty factor of 30 psi in the safety analyses.

Gen The Steam Generator Level-Low trip provides protection against a loss of feedwater flowincident and assures that the design pressure of the Reactor Coolant System willnot be exceeded due to loss of the steam generator heat sink. This specified setpoint provides allowance that there willbe sufficient water inventory in the steam generator at the time of the trip to provide sufficient time for any operator action to initiate auxiliaty feedwater before reactor coolant system subcooling is lost. This trip also protects against violation of the spec Tiied acceptable fuel design limits (SAFDL)for DNBR, offsite dose and the loss of shutdown margin forasymmetric steam generator transients such as the opening of a main steam safety valve or atmospheric dump valve.

The Local Power Density-High trip, functioning from AXIALSHAPE INDEXmonitoring, is provided to ensure that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a consequehce of axial power maidistributions. A reactor trip is initiated whenever the AXIALSHAPE INDEXexceeds the allowable limits of Figure 2,2-2. The AXIALSHAPE INDEXis calculated from the upper and lower excore neutron detector channels.

The calculated setpoints are generated as a function of THERMALPOWER level with the allowed CEA group position being inferred from the THERMALPOWER level, The trip is automatically bypassed below 15% power.

The maximum AZIMUTHALPOWER TILTand maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints.

In addition, CEA group sequencing ln accordance with the Spec Tiications 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 occurrence prior to a Power Level-High trip is assumed.

ST. LUCIE-UNiT2 B 2-5 Amendment No. BB, 89

F ND I

I 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 available. The trip is delayed 10 minutes following a reduction in flow to below the trip setpoint and the trip does not occur ifflowis restored before 10 minutes elapses.

No credit was taken forthis trip in the safety analysis.

Its functional capability at the specwed trip setting is required to enhance the overall reliabilityof the Reactor Protective System.

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 administratively enforced startup rate limit. The trip is not credited in any design basis accident evaluated in UFSAR Chapter 15; however, the trip is considered in the safety analysis in that the presence of this trip function precluded the need for specific analyses of other events initiated from subcritical conditions, a

w-The Reactor Coolant Row - Lowtrip provides core protection against DNB in the event of a sudden sign Tiicant decrease in RCS flow. The Reactor trip setpoint on low RCS flowis calculated by a relationship between steam generator differential pressure, core inlet temperature, instrument errors and response times. When the calculated RCS flowfalls 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.

ST. LUCIE-UNIT2 B2%

Amendment No. 46,89

BO WNMA Gl 3.1.1.1 The SHUTDOWN MARGINshall be greater than or equal to 5000 pcm.

MODES 1,2', 3 and 4.

hQXLQH:

With the SHUTDOWN MARGINless than 5000 pcm, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 1720 ppm boron or equivalent until the required SHUTDOWN MARGINis restored.

4.1.1.1.1 The SHUTDOWN MARGINshall be determined to be greater than or equal to 5000 pcm:

a.

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

Ifthe inoperable CEA is not fullyinserted, and is immovable as a result of excessive friction or mechanical interference or Is known to be untrippable, the above required SHUTDOWN MARGINshall be verified acceptable with an increased allowance forthe withdrawn worth of the immovable or untrippable CEA(s),

b.

When in MODE 1 or MODE 2 with Keffgreater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3.1.3.6.

c.

When in MODE 2 with Keffless than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticalityby verifying that the predicted critical CEA position is within the limits of Specmcation 3.1.3.6.

See Special Test Exception 3.10.1.

ST. LUCIE-UNIT2 Sf4 M Amendment No. &,69

Q~Qg: (Continued) e.

With one full-length CEA misaligned from any other CEA in its group by more than 15 inches beyond the time constraints shown in Figure 3,1-1a, reduce power to c 70% of RATED THERMALPOWER prior to completing ACTIONe.1 or e.2.

1.

Restore the CEA to OPERABLE status within its specified alignment requirements, or 2.

Declare the CEA inoperable and satisfy SHUTDOWN MARGINrequirement of Specification 3.1.1.1.

Afterdeclaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided:*

a)

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the remainder of the CEAs in the group with the inoperable CEA shall be aligned to within 7.0 inches of the inoperable

= CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3,1-2; the THERMALPOWER level shall be restricted pursuant to Spec Tiication 3.1.3.6 during subsequent operation.

b)

The SHUTDOWN MARGINrequirement of SpecTiicatlon 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Otherwise, be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

f.

With one or more full-length CEA(s) misaligned from any other CEAs in its group by more than 7.0 inches but less than or equal to 15 inches, operation in MODES 1 and 2 may continue, provided that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the misaligned CEA(s) is either:

1.

Restored to OPERABLE status within its above specified alignment requirements, or 2.

Declared inoperable and the SHUTDOWN MARGINrequirement of Spec Tiication 3.1.1.1 is satisfied. After declaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided:

a)

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the remainder of the CEAs In the group with the inoperable CEA shall be aligned to within 7,0 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1-2; the THERMALPOWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

b)

The SHUTDOWN MARGINrequirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Otherwise, be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

g.

With one fulldength CEA inoperable due to causes other than addressed by ACTIONa., above, and inserted beyond the Long Term Steady State Insertion Limits but within its above specified alignment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of SpeciTication 3.1.3.6.

Ifthe pre-misalignment ASI was more negative than Z.15, reduce power to z 70% of RATED THERMALPOWER or 70% of the THERMALPOWER level prior to the misalignment, whichever is less, prior to completing ACTION e.2.a) and e2.b).

ST. LUCIE-UNIT2

$41-19 Amendment No. 8,89

3.2,5 The following DNB-related parameters shall be maintained within the limits shown on Table 3.2-2.

a, Cold Leg Temperature b.

Pressurizer Pressure c.

Reactor Coolant System Total Flow Rate d,

AXIALSHAPE INDEX hQILOOI:

With any of the above parameters exceeding its limit, restore the parameter to within its limitwithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMALPOWER to s 5% of RATEDTHERMALPOWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.2.5.1 4,2.5,2 Each of the parameters of Table 3.2-2 shall be veried to be within their limits by instrument readout at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The Reactor Coolant System total flowrate shall be determined to be within its limitby measurement't least once per 18 months.

Not required to be performed until THERMALPOWER is a 90% of RATEDTHERMALPOWER.

ST. LUCIE-UNIT2

$42-14 Amendment No. 89

1~~AANL.~

I tc dl ADI Tl N I

ME INSTRUMENT MINIMUM CHANNELS OPERABLE APPLICABLE MODES ALARM/TRIP MEASUREMENT SETPOINT RANGE ACTION PROCESS MONITORS (Continued) c.

Noble Gas Effluent Monitors Reactor AuxiliaryBuilding Exhaust System (Plant Vent Low Range Monitor) li.

Reactor AuxiliaryBuilding Exhaust System (Plant Vent High Range Monitor) iii.

Steam Generator Blowdown Treatment Facility Building Exhaust System Iv.

Steam Safety Valve Discharge¹ 1/steam header v.

Atmospheric Steam Dump Valve 1/steam header Discharge¹ vi.

ECCS Exhaust 1,2,3&4 1,2,3&4 1,2,3&4 1,2,3&4 1,2,3&4 1,2,3&4 10 -10 pCVcc 10 -10 pCVcc 10 -10 ICVcc 10 -10 pCVcc 10 -10 IjCVcc 10 -10 pCVcc 27 27 27 27

'**The Alarm/Trip Setpoints are determined and set in accordance with the requirements of the Offsite Dose Calculation Manual.

¹ The steam safety valve discharge monitor and the atmospheric steam dump valve discharge monitor are the same monitor.

ST. LUCIE-UNIT2 3/43-26 Amendment No. 25, H, 89

Pages 3/4 &4through 3/4 64 have been DELETED.

Page 3/46-9 is the next valid page, ST. LUCIE-UNIT2 3/4 6-3 Amendment No. 86,66, f$,89

~

Pages 3/4 6-22 through 3/4 6-23 have been DELETED.

Page 3/4 6-24 ls the next valid page.

ST. LUCIE-UNIT2 3/4 6-21 Amendment No. f$,68,88, 6g

V 3.6.5 The primary containment vessel to annulus vacuum relief valves shall be OPERABLE with an actuation setpoint of 9.85 a 0.35 inches water gauge.

MODES 1, 2, 3 and 4.

hQILQH:

With one primaiy containment vessel to annulus vacuum relief valve inoperable, restore the valve to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following30 hours.

4.6,5 No additional Surveillance Requirements other than those required by Specification 4.0.5.

ST. LUCIE-UNIT2 8/4 6-26 Amendment No. 60, Sg

c)

Verifyingthat all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a safety injection actuation signal.

7.

Verifyingthe diesel generator operates forat least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

During the first2 hours of this test, the diesel generator shall be loaded within a load band of 3800 to 3985 kW'nd during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded within a load band of 3450 to 3685 kW'.

The generator voltage and frequency shall be 4160 a 420 volts and 60 a 12 Hz within 10 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test.

8.

Verifyingthat the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 3935 kW.

9.

Verifyingthe diesel generator's capability to:

a)

Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power.

b)

Transfer its load to the offsite power source, and c)

Be restored to its standby status.

10.

Verifyingthat with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizes the emergency loads with offsite power.

11.

Verifyingthat the fuel transfer pump transfers fuel from each fuel storage tank to the engine-mounted tanks of each diesel via the installed cross connection lines.

¹ This band is meant as guidance to avoid routine overloading of the engine. Variations in load in excess of this band due to changing bus loads shall not invalidate this test,

This test may be conducted in accordance with the manufacturer's recommendations concerning engine prelube period.

ST. LUCIE-UNIT2 3/48-7 Amendment No. 89,69,78, 89

EFV Li GO ERA ON TO 3.9.6 The manipulator crane shall be used for movement of fuel assemblies, WMlolwithout CEAs, and shall be OPERABLE with:

a.

A minimum capacity of 2000 pounds, and b.

An overload cut off limitof less than or equal to 3000 pounds.

During movement of fuel assemblies, with or without CEAs, withinthe reactor pressure vessel.

hQIlQH:

With the requirements for crane OPERABILITYnot satisfied, suspend use of any inoperable manipulator crane from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel.

4.9.6 The manipulator crane used for movement of fuel assemblies, with or without CEAs, within the reactor pressure vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of such operations by performing a load test of at least 2000 pounds and demonstrating an automatic load cut offbefore the crane load exceeds 3000 pounds.

ST. LUCIE-UNIT2 Amendment No. 89

~

C 3.112.5 The concentration of oxygen in the waste gas decay tanks shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

Atall times.

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a, With the concentration of oxygen in the waste gas decay tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b.

With the concentration of oxygen in the waste gas decay tank greater than 4% by volume and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the system and immediately commence reduction of the concentration of oxygen to less than or equal to 2% by volume.

c.

The provisions of Specifications 3,0,3 and 3.0.4 are not applicable.

4.112,5.1 The concentration of oxygen in the waste gas decay tank shall be determined to be within the above limits by continuously monitoring the waste gases in the on seivice waste gas decay tank.

4.11'.5,2 With the oxygen concentration in the on service waste gas decay tank greater than 2% by volume as determined by Specification 4.112.5,1, the concentration of hydrogen in the waste gas decay tank shall be determined to be within the above limits by gas partitioner sample at least once per 24 houts.

ST. LUCIE-UNIT2 3/411-14 Amendment No.~, 89

assumptiqns used in establishing the LInear Heat Rate, Thermal Margin/Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid, An AZIMUTHALPOWER TILT> 0.10 is not expected and ifitshould occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The requirement that the measured value of T be multiplied by the calculated values of F, and F to determine F~ and F~~ is applicable only when F, and F are calculated with a non-full core power distribution analysis code. When monitoring a reactor core power distribution, F, or F with a fullcore power distribution analysis code the azimuthal tiltis explicitlyaccounted foras part ofhce radial power distribution used to calculate F and F~

The Suiveillance Requirements forverifying that F<, F, and T, are withintheir limits provide assurance that the actual values ofF, F, and T, do not exceed the assumed values. Verifying F~

and F, after each fuel loading prior to exceeding 75% of RATED THERMALPOWER provides additional assurance that the core was properly loaded.

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

The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of a 1.28 in conjunction with ESCU methodology throughout each analyzed transient.

The 12-hour periodic surveillance of these parameters through Instrument readout is sufficient to ensure that the parameters are restored within their limitsfollowing load changes and other expected transient operation. The 18-month periodic measurement of the RCS total flowrate is adequate to detect flowdegradation and ensure correlation of the flowindication channels with measured flowsuch that the indicated percent flowwillprovide sufficient veriifiication of flowrate on a 12-hour basis.

ST. LUCIE-UNIT2 B 3/4 2-2 AmendmentNo.B 89

ML17229A481

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