ML17223B258

Amend 109 to License DPR-67,deleting Facility Unit 1 TS 3/4.2.2, Total Planar Radial Peaking Factor-F(txy) & All Refs

ML17223B258
Person / Time
Site:	Saint Lucie
Issue date:	08/10/1991
From:	Berkow H Office of Nuclear Reactor Regulation
To:
Shared Package
ML17223B259	List: ML17223B258 ML17223B260
References
NUDOCS 9109170328
Download: ML17223B258 (19)

Text

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UNITED STATES NUCLEAR REGULATORY COIVIIVIISSION WASHINGTON, D.C. 20555 FLORIDA POWER

& LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

109 License No.

DPR-67 1.

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

A.

The application for amendment by Florida Power

& Light Company, (the licensee) dated April 17,

1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assur ance (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 >iith 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.

9109'170328 910910 PDR ADOCN, 05000335,',

P PDR,

i 2.

Accordingly, Facility Operating License No.

DPR-67 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending 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. 109, are hereby incorporated in the license.

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

3.

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

FO E

NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

H bert N. Ber ow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 109 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 com-pleteness.

Remove Pa es Ia IV 1-7 3/4 2-2 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-11 3/4 10-2 B 3/4 2-1 B 3/4 2-2 Insert Pa es Ia IV 1-7 3/4 2-?

3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-11 3/4 10-2 B 3/4 2-1 B 3/4 2-2

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 Action..

1.2 Axial Shape Index...

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1.3 Azimuthal Power Tilt...

1.4 Channel Calibration.

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1.5 Channel Check...,.........................

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1.6 Channel Functional Test..

1-2 1.7 Containment Vessel Integrity......................

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1-2 1.8 Controlled Leakage..

. 1-2 1.9 Core Alteration.............................,................

1-2 1.10 Dose Equivalent I-131.

1.11 E Average Disintegration Energy.

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1.12 Engineered Safety Features Response Time.....................

1-3 1.13 Frequency Notation............................

1-3

l. 14 Gaseous Radwaste Treatment System..............

1.15 Identified Leakage...........................................

1-3 1-4 1.16 Low Temperature RCS Overpressure Protection 1.17 Member(s) of the Public.....................

Range............

1-4 1-4 1.18 Offsite Dose Calculation Manual (ODCM).......................

1-4 1.19 Operable

- Operability.......................................

1-5 1.20 Operational Mode - Mode....................

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1.21 Physics Tests

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ST.

LUCIE - UNIT 1

Amendment No. 27 82 Q 69

INDEX DEFINITIONS SECTION PAGE 1.23 Process Control Program (PCP).

1.24 Purge - Purging.

1.25 Rated Thermal Power 1-5 1-5 1-6 1.26 Reactor Trip System

Response

Time 1.27 Reportable Event.

1.28 Shield Building Integrity........

1.29 Shutdown Margin.............

1.30 Site Boundary..........

1.31 Source Check...................

1-6 1-6 1-6 1-6 1-6 1-6 1.32 Staggered Test Basis.....

.33 Thermal Power..............................................

1 1-7 1-7 1.34 Unidentified Leakage.

1.35 Unrestricted Area.

1-7 1-7 1.36 Unrodded Integrated Radial Peaking Factor - Fr....

. 1-7 1.37 DELETED 1-7 ST.

LUCIE - UNIT 1

Ia Amendment No.

$9,ggl, gp, 1O9,

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.0 APPLICABILITY..............................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1.1 BORATION CONTROL...

Shutdown Margin - T

) 200'F.......

Shutdown Margin - T 200'F....

Boron Dilution...........

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Moderator Temperature Coefficient Minimum Temperature for Criticality.

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3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-4 3/4 1-5 3/4 1-7 3/4.1.2 BORATION SYSTEMS..............

Flow Paths - Shutdown................................

Flow Paths - Operating............

Charging Pump - Shutdown..........

Charging Pumps - Operating.

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Boric Acid Pumps - Shutdown.......

Boric Acid Pumps - Operating......

Borated Water Sources

- Shutdown..

Bor ated Water Sources

- Operating..................

3/4 1-8 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-13 3/4 1-14 3/4 1-15 3/4 1-16 3/4 1-4 3/4.1. 3 MOVABLE CONTROL ASSEMBLIES.................

Full Length CEA Position.........

Position Indicator Channels......

CEA Drop Time.....................

Shutdown CEA Insertion Limit.....

Regulating CEA Insertion Limits......................

3/4 1-20 3/4 1-20 3/4 1-24 3/4 1-26 3/4 1-27 3/4 1-28 ST.

LUCIE - UNIT 1

Amendment No97,

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4.2 POWER DISTRIBUTION LIMITS PAGE 3/4.2.1 LINEAR HEAT RATE 3/4 2-1 3/4.2.2 DELETED 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR-T F

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3/4 2-6 3/4 2-9 3/4.2.4 3/4

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~ 5 DNB PARAMETERS 3/4 2-13 AZIMUTHAL POWER TILT - T............................

3/4 2-11 q

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION...................

3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION......................................

3/4 3-9 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitorl ng Incore Detectors Seismic Instrumentation..............................

Meteorological Instrumentation..........

Remote Shutdown Instrumentation Fire Detection Instrumentation.......................

Accident Monitoring Instrumentation......

Radioactive Liquid Effluent Monitoring Instrumentation........................

3/4 3-21 3/4 3-21 3/4 3-25 3/4 3-27 3/4 3-30 3/4 3-33 3/4 3-37 3/4 3-41 3/4 3-45 Radioactive Gaseous Effluent Monitoring Instrumentation......................................

3/4 3-50 3 4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION....,...

3/4 4-1 3/4.4.2 SAFETY VALVES -

SHUTDOWN.............................

3/4 4-2 3/4.4.3 SAFETY VALVES - OPFRATING.......................

.... 3/4 4-3 ST.

LUCIE - UNIT 1

IV Amendment No. g7,

$7,

$g,

$7,p p 1O'

t DEFINITIONS STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASiS shall consist of A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the scecified t st interval into n equal subintervals, and Tne testing of one system',

subsystem, train or other designated component at the beginning of each subinterval.

THER/NL PO~ ER 1.33 THERf1AL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

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~ UNIDEHTIFI ED LEAKAGE 1.34 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.35 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY

)access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

UHRODDED INTEGRATED RADIAL PEAKING FACTOR - Fr 1.36 The UHRODDED INTEGRATED RADIAL PEAKING FACTOR is the ratio of the peak pin power to the average pin power in an unrodded core, excluding tilt.

ST.

LUCIE - UNIT 1

1-7 Amendment No. 59i,

109,

TABLE 1.1 FRE UENCY NOTATION NOTATION 4/t5*

S/U N.A.

~FRE UENCY At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> At least once per 7 days At least 4 per month at intervals of no greater than 9 days and a minimum of 48 per year At least once per 31 days At least once per 92 days At least once per 184 days At least once per 18 months Prior to each reactor startup Completed prior to each release Not applicable

• For Radioactive Effluent Sampling

• • For Radioactive Batch Releases Only ST.

LUCIE - UNIT 1

1-8 Amendment No.gg/,69

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued C.

Ver ifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2, where 100 percent of maximum allowable power represents the maximum THERMAL POWER allowed by the following expression:

MxN where:

1.

M is the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination,.

2.

N is the maximum allowable fraction of RATED THERMAL POWER as determined by the F

curve of Figure 3.2-3.

T 4.2.1.4 Incore Detector Monitorin S stem

- The incore detector monitor-ing system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:

a.

Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 days of accumulated operation in MODE l.

b.

Have their alarm setpoi nt adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alarms:

l.

A measurement-calculational uncertainty factor of 1.07, 2.

An engineering uncertainty factor of 1.03, 3.

A THERMAL POWER measurement uncertainty factor of 1.02.

PIf the core system becomes inoperable, reduce power to M x N within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and monitor linear heat rate in accordance with Specification 4.2.1.

ST.

LUCIE - UNIT 1

3/4 2-2 Amendment No. 77, g7, gg,5$ @. 14~

109,

THIS PAGE INTENTIONALLY LEFT BLANK ST.

LUCIE - UNIT 1

3/4 2-5 Amendment No. g7, gg, gg, 63

POWER DISTRIBUTION LIMITS DELETED ST.

LUGIE - UNIT 1

3/4 2-6 Amendment No. 27,82 EH.N.

109,

PO ER DISTRIBUTION LIMITS DELETED ST.

LUCIE - UNIT 1

3/4 2-7 Amendment No. g7,5$, 109,

c' H

I Q

1.0

1. 7, 1.0)

UNACCEPTABLE OPERATION REGION O

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,M D

0.9 0.8 0.7 ACCEPTABLE OPERATION REGION (1.78,0.9 tD A.

ID et O

0.6

l. 70

l. 71 1.72 1.73 1.74 1.75 Measured F>T 1.76 1.77 1.78 FIGURE 3.2-3 Allowable Combinations Of Thermal Power And F<T

POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - T LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL,POWER TILT (T

) shall,not exceed 0.03.

q APPLICABILITY'ODE 1*

ACTION:

a.

With the indicated AZIMUTHAL POWER TILT determined to be

~ 030 but

< 0. 10, either correct the power tilt within two hours or determine within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per subsequent 8 )ours, that the TOTAL INTEGRATED RADIAL PEAKING FACTOR (F

) is within the limits of Specification 3.2.3.

b.

With the indicated AZIMUTHAL POWER TILT determined to be

> 0. 10, operation may proceed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided that the TOTAL INTEGRATED RADIAL PEAKING FACTOR (F

) is within the limits of Specification 3.2.3.

Subsequent operation for. the purpose of measurement and to identify the cause of the tilt is allowable provided the THERMAL POllER level is restricted to

< 20K of the maximum allowable THERMAL POWER level for.the existing Reactor Coolant Pump combination.

SURVEILLANCE REQUIREMENT 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.

4.2.4.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit by:

a.

Calculating the tilt at least once per 7 days when the Subchannel Deviation Alarm is OPERABLE,

• See Special Test Exception 3.10.2.

ST.

LUCIE - UNIT 1

3/4 2-11 Amendment No. 9,

$7, $9.

109,

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued b.

Calculating the tilt at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the Subchannel Deviation Alarm is inoperable, and c.

Using the incore detectors to determine the AZIMUTHAL POWER TILT at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when one excore channel is inoperable and THERMAL POWER is ) 75/ of RATED THERMAL POWER.

ST.

LUCIE - UNIT 1

3/4 2-12 Amendment No. 27,

SPECIAL TEST EXCEPTIONS GROUP HEIGHT INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION I

3.10.2 The group height, insertion and power distribution limits of S peci ficati on s 3.1.1. 4, 3.1.3.1, 3.1.3. 2, 3.1.3. 5, 3. 1.3. 6, 3.2. 3, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and b.

The limits of Specification 3.2.1 are maintained and deter-mined as specified in Specification 4.10.2.2 below.

APPLICABILITY:

MODES 1

and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.2.3 and 3.2.4 are suspended, either:

a.

Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or b.

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

SURVEILLANCE RE UIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS.in which the requirements of Specifications 3.1.1.4,

3. 1.3. 1, 3. 1.3.2,

3. 1.3.5,

3. 1.3. 6, 3.2. 3 or 3.2.4 are suspended and shall be verified to be within the test power plateau.

4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5X of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.4, 3.1.3.1,

3. 1.3.2,

3. 1.3.5,

3. 1.3.6, 3.2. 3 or 3.2.4 are suspended.

ST.

LUCIE - UNIT 1

3/4 10-2 Amendment No.

$7, IOgy

3/4. 2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a

LOCA, the peak temperature of the fuel cladding will not exceed 2200'F.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provides adequate monitoring of the core power distribution and is=-capable of verifying that the linear heat rate does not exceed its limits'he Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2.

In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assump-tions are made:

1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3e6 are satis'fied,

2) the AZIMUTHAL POWER TILT restrictions of Specifica-tion 3.2.4 are satisfied, and 3) the TOTAL INTEGRATED RADIAL PEAKING FACTOR does

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not exceed the limits of Specification 3.2.3."

The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1.

The setpoints for these alarms include allowances, set in the conservative directions, for 1) a measurement-calculational uncertainty factor of 1.07,

2) an engineering uncertainty factor of 1.03, 3) a THERMAL POWER measurement uncertainty factor of 1.02.

3/4.2.3 and 3/4.2.4 TOTAL INTEGRATED RADIAL PEAKING FACTOR - FAND AZIMUTHAL POWER TILT - T The limitations on F

and T

are provided to ensure that the assump-r.

q tions used in the analysis for establishing the Linear Heat Rate and Local Power Density-High LCOs and LSSS setpoints and ST.

LUCIE - UNIT 1

B 3/4 2-1 Amendment No. )7/,P,PP,]P, Id. 109,

POWER DISTRIBUTION LIMITS BASES the DNB Margin LCO, and Thermal Margin/Low Pressure LSSS setpoints remain vali( during operation at the various allowable CEA group insertion limits.

If For Tq exceed their basic limitations, operation may continue under the additional restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat Rate, Thermal Margin/Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid.

An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

Th'e requirement that the measured value of (1+Tq) be multiplied by the calculated value of Fr to determine F$ is applicable only when Fr is calculated with a non-full core power distribution analysis.

With a full core power distribution analysis code the azimuthal tilt is explicitly accounted for as part of the radial power distribution used to calculate Fr.

t The surveillance requirements for verifying that FTr and T

are within their limits provide assurance that the actual values of FT anl Tq do not exceed the assumed values.

Verifying Fg after each fuel loading prior to exceeding 75K of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

3 4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of opera-tion assumed in the transient and accident analyses.

The limits are consis-tent with the safety analyses assumptions and have been analytically demon-strated adequate to maintain a minimum DNBR of > 1.22 throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

ST.

LUCIE - UNIT 1

B 3/4 2-2 Amendment No. g7, gg, Q, gg, IOg,