ML17229A288

Amend 150 to License DPR-67,modifying Specifications for Selected cycle-specific Reactor Physics Parameters to Refer to Facility Core Operating Limits Rept for Limiting Values

ML17229A288
Person / Time
Site:	Saint Lucie
Issue date:	04/01/1997
From:	Hebdon F NRC (Affiliation Not Assigned)
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ML17229A287	List: ML17229A288
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NUDOCS 9704030132
Download: ML17229A288 (41)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20555-0001 LORIDA POWER 5 LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE PLANT UNIT NO.

1 MENDMENT TO FACILITY OPERATING LICENSE Amendment No. i5O License No.

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

A.

B.

C.

D.

E.

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

dated December 9,

1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in, conformity with the application, the provisions of the Act, and the rules and regulations of the

'ommission; 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.

9704030i32 97040i 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. 150, 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:

April 1, 1997 Frederick J.

Hebdon, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

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ATTACHMENT TO LICENSE AMENDMENT NO.

150 TO FACILITY OP RATING 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.

emove Pa es I

XV 1-2 3/4 1-5 3/4 1-21 3/4 1-22 3/4 1-23 3/4 1-28 3/4 1-30 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-13 3/4 2-14 3/4 2-15 3/4 9-1 B 3/4 1-1 B 3/4 1-4 B 3/4 2-1 B 3/4 9-1 6-19 Insert Pa es I

XV 1-2 3/4 1-5 3/4 1-21 3/4 1-22 3/4 1-23 3/4 1-28 3/4 1-30 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-9 3/4 2-13 3/4 2-14 3/4 9-1 B 3/4 1-1 B 3/4 1-4 B 3/4 2-1

,B 3/4 9-1 6-19 6-19a

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I DEFINITIONS INDEX SECTION PAGE 1.0 DEFINITIONS 1.1 Action...

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1,2 Axial Shape Index......

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1.3 Azimuthal Power Tilt

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1.4 Channel Calibration... ~.............,

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1,5 Channel Check.... ~.................. ~....

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1.6 Channel Functional Test.......

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. 1-2 1.7 Containment Vessel Integrity

.. 1-2 1.8 Controlled Leakage 1-2 1.9 Core Alteration.....

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1.9a Core Operating Limits Report (COLR).... ~.........................

1-2 1 ~10 Dose Equivalent I-131

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1 3 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 1.14 Gaseous Radwaste Treatment System.. ~..............,..... ~..... 1-3 1.15 Identified Leakage...,...............,.......... ~.............

1-4 1 ~16 Low Temperature RCS Overpressure Protection Range..... ~...........

1-4 1.17 Member(s) of the Public........ ~...........,... ~... ~.........

1-4 1.18 Offsite Dose Calculation Manual (ODCM)

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1.19 Operable - Operability

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1.20 Operational Mode - Mode............ ~.................,........

1-5 1,21 Physics Tests

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1.22 Pressure Boundary Leakage ST. LUCIE-UNIT1

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AmtIq@nent No. 27; 82, 66, se I DU

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C INDEX ADMINISTRATIVECONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION...

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6.7 SAFETY LIMITVIOLATION....... ~....

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6-12 6.8 PROCEDURES AND PROGRAMS,.....

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6-13 6.9 REPORTING RE UIREMENTS 6.9.1 ROUTINE REPORTS.................

1 Startup Report Annual Reports.............

Monthly Operating Reports..

Annual Radioactive Effluent Release Report

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6-15b 6-15b 6-16 6-16a 6-17 Annual Radiological Environmental Operating Report.......,

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6-18 Core Operating Limits Report (COLR)........

6.9.2 SPECIAL REPORTS 6-19 6 ".ga 6.10 RECORD RETENTION........,......,.........

6-20 6.11 RADIATIONPROTECTION PROGRAM 6-21 6.12 HIGH RADIATIONARE

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6.13 PROCESS CONTROL PROGRAM....,,

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6-23 6.14 OFFSITE DOSE CALCULATIONMANUAL........................

6-23 ST. LUCIE-UNIT1 XV Amendment Nq.~, &,69, eS, ~,~ >>U

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'EFINITIONS CHANNEL FUNCTIONALTEST

'.6 A CHANNEL FUNCTIONALTEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.,

CONTAINMENTVESSEL INTEGRITY 1.7 CONTAINMENTVESSEL INTEGRITYshall exist when:

a.

All containment vessel penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed position except for valves that are open on an intermittent basis under administrative control.

b.

All containment vessel equipment hatches are closed and sealed, c.

Each containment vessel air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and e.

The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGEshall be the seal water flow supplied from the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE 1.9a CORE ALTERATIONshall be the movement or manipulation of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel.

Exceptions to the above include shared (4 fingered) control element assemblies (CEAs) withdrawn into the upper guide structure (UGS) or evolutions performed with the UGS in place such as CEA latching/unlatching or verification of latching/

unlatching which do not constitute a CORE ALTERATION. Suspension of CORE ALTERATIONSshall not preclude completion of movement of a component to a safe position.

OPERATING LIMITS REPORT COLR The COLR is the unit-specific document that provides cycle specific parameter limits for the current operating reload cycle. These cycle-specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.9.1.11.

Plant operation within these limits is addressed in individual Specifications.

ST. LUCIE-UNIT1 1-2 Ameg~ent No. 60, 448,

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I REACTIVITYCONTROL TEMS hllODERATOR TEMPERATURE COEFFICIENT LIMITINGCONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be maintained within the limits specified in the COLR. The maximum positive limit shall be:

a.

Less positive than+7 pcm/'F whenever THERMALPOWER is 5 70% of RATED THERMALPOWER, and b.

Less positive than+2 pcm/'F whenever THERMALPOWER is > 70% of RATED THERMAL POWER, APPLICABILITY:

MODES 1 and 2*¹.

ACTION:

With the moderator temperature coefficient outside any one of the above limits, be in HOT STANDBYwithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, SURVEILLANCEREQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by.confirmatory measurements.

MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

With K,h 1.0.

¹ See Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 1-5 Amendment No. 87; 88, 86 150

ii REACTIVITYCONTROL TEMS

'FULL LENGTH CEA POSITION continued LIMITINGCONDITION FOR OPERATION continued 2.

Declared inoperable and satisfy SHUTDOWN MARGIN requirements of Specification 3.1.1.1.

After declaring 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 5 14 days per calendar year provided all of the following conditions are met:

a)

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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 THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; otherwise, be in at least HOT STANDBYwithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

e.

With one full length CEA misaligned from any other CEA in its group by 15 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 3.1-1a.

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 3.1-1a, reduce power to 6 70% of RATED THERMALPOWER prior to completing ACTION f.1 or f.2.

1.

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

Declare the CEA inoperable and satisfy the SHUTDOWN MARGIN requirements of Specification 3.1.1.1.

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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

ST. LUCIE - UNIT 1 3/4 1-21 Amendment No. &"50

0

REACTIVITYCONTROL TEMS

.>I FULL LENGTH CEA POSITION continued LIMITINGCONDITION FOR OPERATION continued b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

g.

With more than one full length 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

h.

With one full-length CEA inoperable due to causes other than addressed by ACTION a 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.

SURVEILLANCEREQUIREMENTS 4.1.3.1.1 4.1.3.1.2 4.1.3,1.3 4.1.3.1.4 The position of each full-length CEA shall be determined to be within 7.5 inches (indicated position) of all other CEAs in its group at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Deviation Circuit and/or CEA Block Circuit are inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Each full-length CEA not fully inserted shall be determined to be OPERABLE by inserting it at least 7.5 inches at least once per 92 days.

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 misaligned from all other CEAs in its group by more than 7.5 inches (indicated position).

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 entry into MODE 2 from MODE 3, except that such verification need not be performed more often than once per 92 days, and b.

At least once per 6 months.

,* The licensee shall be excepted from compliance during the startup test program for an entry 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. 44, QO, 74;

~ 150

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DELETED ST. LUCIE - UNIT 1 3/4 1-23 Amendment No. W )gP

REACTIVITYCONTROL TEMS REGULATING CEA INSERTION LIMITS LIMITINGCONDITION FOR OPERATION 3.1.3.6 The regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits specified in the COLR (regulating CEAs are considered to be fully withdrawn when withdrawn to at least 129.0 inches) with CEA insertion between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits restricted to:

a.

5 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval, b.

~ 5 Effective Full PoWer Days per 30 Effective Full Power Day interval, and I

c.

5 14 Effective Full Power Days per calendar year.

APPLICABILITY:

MODES 1 and 2*¹,

ACTION:

a.

With the regulating CEA groups inserted beyond the Power Dependent Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either:

1.

Restore the regulating CEA groups to within the limits, or 2.

Reduce THERMALPOWER to less than or equal to that fraction of RATED THERMALPOWER which is allowed by the CEA group position and insertion limits specified in the COLR.

b.

With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits for intervals

> 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval, except during operation pursuant to the provisions of ACTION items c. and d. of Specification 3.1.3.1, operation may proceed provided either:

1.

The Short Term Steady State Insertion Limits are not exceeded, or 2.

Any subsequent increase in THERMALPOWER is restricted to 5 5% of RATED THERMALPOWER per hour.

See Special Test Exceptions 3.10,2 and 3.10.5.

With K,h 1.0.

ST. LUCIE - UNIT 1 3/4 1-28 Amendment No.@

150

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DELETED ST. LUCIE - UNIT 1 3/4 1-30 Amendment No. &2, 48,,150

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3/4.2 POWER DIST IBUTION LIMITS LINEAR HEAT RATE LIMITINGCONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed the limits specified in the COLR, APPLICABILITY:

MODE 1.

ACTION:

H With the linear heat rate exceeding its limits, as indicated by four or more coincident incore channels or by the AXIALSHAPE INDEX outside of the power dependent control limits of COLR Figure 3.2-2, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:

a.

Restore the linear heat rate to within its limits within one hour, or b.

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

SURVEILLANCE REQUIREMENTS 4.2.1.1

.4.2.1.2 4,2.1.3 The provisions of Specification 4.0.4 are not applicable.

The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring system.-

Excore De ec or Monitorin S stem - The excore detector monitoring system may be used for monitoring the linear heat rate by:

a.

Verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the full length CEAs are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Limit of Specification 3.1.3.6.

b.

Verifying at least once per 31 days that the AXIALSHAPE INDEX alarm setpoints are adjusted to within the limits shown on COLR Figure 3.2-2.

ST. LUCIE - UNIT 1 3/4 2-1 Amendment No. aa, 150

'OWER DISTRIBUTION SURVEILLANCEREQUIREMENTS continued c.

Verifying that the AXIALSHAPE INDEX is maintained within the allowable

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limits of COLR Figure 3.2-2, where 100 percent of maximum allowable power represents the maximum THERMALPOWER allowed by the following expression:

MxN where:

1.

M is the maximum allowable THERMALPOWER level for the existing Reactor Coolant Pump combination.

2.

N is the maximum allowable fraction of RATED THERMALPOWER as determined by the F ~ curve of COLR Figure 3.2-3.

4.2.1.4 Incore Detector Monitorin S stem'- The incore detector monitoring system may be used for monitoring the linear heat rate 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 1.

b.

Have their alarm setpoint adjusted to less than or equal to the limits shown on COLR Figure 3.2-1.

¹ If the incore 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 4.2.1.3.

ST. LUCIE - UNIT 1 3/4 2-2 Amendment No. 47; 87; 82, sa, ss, vo, e@,434,43',

1.50

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Pages 3/4 2-4 (Amendment 106), 3/4 2-5 (Amendment 63), and 3/4 2-6 through 3/4 2-8 (Amendment 109) have been deleted from the Technical Specifications.

The next page is 3/4 2-9.

ST. LUCIE - UNIT 1 3/4 2-3 Amendment No. 27; 88, 48',

m, m,84,150

POWER DISTRIBUTION L TS TOTALINTEGRATED RADIALPEAK NG FACTOR - Fi LIMITINGCONDITION FOR OPERATION 3.2.3 The calculated value of F, shall be within the limits specified in the COLR.

APPLICABILITY:

MODE 1'.

ACTION:

With F, not within limits, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

a.

Be in at least HOT STANDBY, or b.

Reduce THERMALPOWER to bring the combination of THERMALPOWER and F, 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 THERMAL POWER level limit on COLR Figure 3.2-4 (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 3.2-4.

SURVEILLANCEREQUIREMENTS

. 4.2.3.1 4.2.3.2 The provisions of Specification 4.0.4 are not applicable.

F," shall be calculated by the expression F, = F,(1 + 7 ) when F, is calculated with a non-full core power distribution analysis code and shall be calculated as F,

= F, when calculations are performed with a full core power distribution analysis code, F, shall be determined to be within its limit at the following intervals.

a.

Prior to operation above 70 percent of RATED THERMALPOWER after each fuel loading, b.

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

Within four hours if the AZIMUTHALPOWER TILT (T,) is ) 0.03.

See Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 2-9 Amendment No. 87; 88, 48, s3150

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'OWER DISTRIBUTION TS DNB PARAMETERS LIMITINGCONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a.

Cold Leg Temperature b.

Pressurizer Pressure c.

Reactor Coolant System Total Flow Rate d.

AXIALSHAPE INDEX APPLICABILITY:

MODE 1.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.2.5,1 4.2.5,2 Each of the parameters of Table 3.2-1 shall be verified to be within their limits by instrument readout at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement* at least once per 18 months.

Not required to be performed until THERMALPOWER is 2 90% of RATED THERMAL POWER.

ST. LUCIE - UNIT 1 3/4 2-13 Amendment No. 97.,

150

l g

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J TABLE 3,2-1 DNB MARGIN LIMITS

= Parameter Cold Leg Temperature Pressurizer Pressure Reactor Coolant Flow Rate AXIALSHAPE INDEX Four Reactor Coolant Pumps Operating 5 549'F 2 2225 psia*

2 345,000 gpm, COLR Figure 3.2-4 Limit not applicable during either a THERMALPOWER ramp increase in excess of 5% of RATED THERMALPOWER or a THERMALPOWER step increase of greater than 10% of RATED THERMALPOWER.

ST. LUCIE - UNIT 1 3/4 2-14 Amendment No. 87; 48,

~,44I.,

150

3/4.9

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'ORON CONCENTRATION REFUELING "RATIONS LIMITINGCONDITION 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 within the limitspecified in the COLR.

APPLICABILITY'ODE6'.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at 2 40 gpm of 1720 ppm boron or its equivalent to restore boron

- concentration to within limits.

SURVEILLANCEREQUIREMENTS 4.9.1.1 4.9.1.2 The boron concentration limit shall be determined prior to:

a.

Removing or unbolting the reactor vessel head, and b.

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

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. LUCIE - UNIT 1 3/4 9-1 Amendment No. 408, 150

3/4 1

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'ASES REACTIVITY TROL SYSTEMS 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,,, The most restrictive condition occurs at EOL, with T,~ at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 3600 pcm is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN required by Specification 3.1.1.1 is based upon this limiting condition and is consistent with FSAR accident analysis assumptions, For earlier periods during the fuel cycle, this value is conservative.

With T, g 5 200 F the reactivity transient resulting from a boron dilution event with a partially drained Reactor Coolant System requires a 2000 pcm SHUTDOWN MARGIN and restrictions on charging pump operation to provide adequate protection.

A 2000 pcm SHUTDOWN MARGIN is 1000 pcm conservative for Mode 5 operation with total RCS volume present, however LCO 3.1.1.2 is written conservatively for simplicity.

3/4.1.1.3 BORON DILUTIONAND ADDITION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration changes in the Reactor Coolant System.

A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 11,400 cubic feet in approximately 26 minutes.

The reactivity change rate associated with boron concentration changes will be within the capability for operator recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT MTC The limiting values of the MTC ensure that the assumptions for the MTC used in the accident and transient analyses remain valid through each fuel cycle, Determination of MTC at the specified conditions ensures that the maximum positive and/or negative values of the MTC will not exceed the limiting values.

ST. LUCIE - UNIT 1 B 3/4 1-1 Amendment No. W, 46, 48, ea,se, 150

REACTIVITYCONTROL SYSTEMS BASES 3/4.1.3 MOVABLECONTROL ASSEMBLIES (continued)

The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA. Conformance with these alignment requirements brings the core, within a short period of time, to a configuration consistent with that assumed in generating LCO and LSSS setpoints, However, extended operation with CEAs significantly inserted in the core may lead to perturbations in 1) local bumup, 2) peaking factors, and 3) available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

The requirement to reduce power in certain time limits, depending upon the previous F', is to eliminate a potential nonconservatism for situations when a CEA has been declared inoperable.

A worst case analysis has shown that a DNBR SAFDL violation may occur during the CEA misalignment if this requirement is not met. This potential DNBR SAFDL violation is eliminated by limiting the time operation is permitted at FULL POWER before power reductions are required.

These reductions will be necessary once the deviated CEA has been declared inoperable.

The time allowed to continue operation at a reduced power level can be permitted for the following re'asons:

1.

The margin calculations that support the Technical Specifications are based on a steady-.

state radial peak of F, > the limits of Specification 3.2.3, 2;

When the actual F, 5 the limits of Specification 3.2.3, significant additional margin exists.

3.

This additional margin can be credited to offset the increase in F', with time that can occur following a CEA misalignment.

4.

This increase in F, ls caused by xenon redistribution.

5.

The present analysis can support allowing a misalignment to exist without correction, if the time constraints and initial F, limits of COLR Figure 3.1-1a are met.

Operability of the CEA position indicators (Specification 3.1.3.3) is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit. The CEA "Full In" and "Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fullywithdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the "Full In" or "Full Out" limits.

ST. LUCIE - UNIT 1 8 3/4 1-4 Amendment No. 48, 88, V4,

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150

3/4.2 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, The Excore Detector Monitoring System performs this function by continuously monitoring the AXIALSHAPE INDEXwith the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIALSHAPE INDEX is maintained within the allowable limits specified in the COLR.

In conjunction with the use of the excore monitoring system and in establishing the AXIALSHAPE INDEX limits, the following assumptions are made:

1) the CEA insertion limits of Specifications 3,1.3,5 and 3.1.3.6 are satisfied, 2) the AZIMUTHALPOWER TILT restrictions of Specification 3.2.4 are satisfied, and 3) the TOTAL INTEGRATED RADIALPEAKING FACTOR does 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 specified in the COLR. The setpoints for these alarms include allowances, set in conservative directions, for 1) a measurement-calculational uncertainty factor, 2) an engineering uncertainty factor, 3) a THERMALPOWER measurement uncertainty factor.

3/4.2.3 and 3/4,2.4 TOTAL INTEGRATED RADIALPEAKING FACTOR - F" AND AZIMUTHALPOWER TILT-T The limitations on F, and T, are provided to ensure that the assumptions 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, 82, 68, es,vo,ee,~,

150

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'/49 REFUELING RATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitation on minimum boron concentration ensures that:

1) the reactor will remain 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 K, is sufficient to prevent reactor criticality with all full length rods (shutdown and regulating) fullywithdrawn.

3/4.9.2 INSTRUMENTATION The OPERABILITYof the wide range logarithmic range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9,3 DECAY TIME 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.

3/4.9A CONTAINMENTPENETRATIONS The requirements on containment penetration closure and OPERABILITYensure that a release of radioactive material within containment will be 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.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

3/4.9.6 MANIPULATORCRANE OPERABILITY The OPERABILITYrequirements of the cranes used for movement of fuel assemblies ensures that:

1) each crane has sufficient load capacity to lift a fuel element, and 2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

ST. LUCIE - UNIT 1 B 3/4 9-1 Amendment No. 66 150

II

h qe ADMINISTRATIVECONT LS

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t ANNUALRADIOLOGICALENVIRONMENTALOPERA ING REPORT (continued) 6.9.1.9 6.9.1.10 At least once every 5 years, an estimate of the actual population within 10 miles of the plant shall be prepared and submitted to the NRC.

At least once every 10 years, an estimate of the actual population within 50 miles of the plant shall be prepared and submitted to the NRC.

6.9.1.11 CORE OPERATING LIMITS REPORT COLR a.

Core operating limits shall be established prior to each reload cycle, or prior to any.remaining portion of a reload cycle, and shall be documented in the COLR for the following:

Specification 3.1

~1.4 Specification 3.1.3.1 Specification 3.1.3.6 Specification 3.2.1 Specification 3.2.3 Specification 3.2.5 Specification 3.9.1 Moderator Temperature Coefficient Full Length CEA Position - Misalignment > 15 inches Regulating CEA Insertion Limits Linear Heat Rate.

Total Integrated Radial Peaking Factor - F T DNB Parameters Refueling Operations - Boron Concentration b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as described in the following documents or any approved Revisions and Supplements thereto:

1.

WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Westinghouse Proprietaty) 2.

NF-TR-95-01, "Nuclear Physics Methodology for Reload Design of Turkey Point 8 St. Lucie Nuclear Plants," Florida Power & Light Company, January 1995.

3.

XN-75-27(A), Rev. 0 and Supplement 1 through 5, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Rev. 0 dated June 1975, Supplement 1 dated September 1976, Supplement 2 dated December 1980, Supplement 3 dated September 1981, Supplement 4 dated December 1986, Supplement 5 dated February 1987.

4.

ANF-84-73(P), Rev. 3, "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuel Corporation, dated May 1988.

5.

XN-NF-82-21(A), Rev. 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, dated September 1983.

6.

ANF-84-93(A), Rev. 0 and Supplement 1, "Stearnline Break Methodology for PWR's," Advanced Nuclear Fuels Corporation, Rev, 0 dated March 1989, Supplement 1 dated March 1989.

ST. LUCIE-UNIT 1 6-19 Amendment No. M, 69, 86,

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'DMINISTRATIVECON S

CORE OPERATING LIMITS REPORT (continued) 7.

XN-75-32(A), Supplements 1, 2, 3, and 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, dated October 1983.

8.

XN-NF-82-49(A), Rev; 1 and Supplement 1, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Advanced Nuclear Fuels Corporation, Rev.

1 dated April 1989, Supplement 1 dated December 1994.

9.

XN-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, dated October 1983.

10. XN-NF-621(A), Rev. 1, "Exxon Nuclear DNB Correlation of PWR Fuel Design," Exxon Nuclear Company, dated September 1983.

11

~ EXEM PWR Large Break LOCA Evaluation Model as defined by:

a)

XN-NF-82-20(A), Rev.

1 and Supplements 1 through 4, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, ail dated January 1990, b)

XN-NF-82-07(A), Rev. 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, dated November 1982.

c)

XN-NF-81-58(A), Rev. 2 and Supplements 1 through 4, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, Rev. 2 and Supplement 1 and 2 dated March 1984, Supplements 3 and 4 dated June 1990.

d)

XN-NF-85-16(A);Volume 1 through Supplement 3; Volume 2, Rev.

1 and Supplement 1, "PWR 17x17 Fuel Cooling Tests Program," Exxon Nuclear Company, all dated February 1990.

e)

XN-NF-85-105(A), Rev. 0 and Supplement 1, "Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs,"

Exxon Nuclear Company, all dated January 1990.

c.

The core operating limits shall be determined such that all applicable limits (e.g fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC within the time period specified for each report.

ST. LUCIE - UNIT 1 6-19a Amendment No. 150

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