ML17229A161

Proposed Tech Specs 1.9a Re Core Operating Limits Rept

ML17229A161
Person / Time
Site:	Saint Lucie
Issue date:	12/09/1996
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17229A160	List: ML17229A159 ML17229A161 ML17229A162
References
NUDOCS 9612120454
Download: ML17229A161 (61)

Text

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment Core 0 eratin Limits Re ort COLR ATTACHMENT3 ST. LUCIE UNIT 1 MARKED-UP TECHNICALSPECIFICATION PAGES INDEXPage I INDEXPage XV Page 1-2 Page B3/41-1 Insert - A Page B3/41-4 Page 3/41-5 Page B 3/42-1 Page 3/4 1-21 Page B 3/49-1 Page 3/4 1-22 Page 3/4 1-23 Page 6-19 Page 3/4 1-28 Insert - B (2 pages)

Page 3/4 1-30 Page 3/42-1 Page 3/42-2 Page 3/42-3 Page 3/42-4 Page 3/42-5 Page 3/42-6 Page 3/42-7 Page 3/42-8 Page 3/4 2-9 Page 3/42-13 Page 3/42-14 Page 3/42-15 Page 3/49-1

'Mi2i20454 qhi20905000335 PDR ADOCK P PDR

INDEX DEF I NIT IOHS SECTION PAGE 1.0 DEFINITIONS l..1 Action.. ~ ~ \ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ 1 1 1.2 Axial Shape Index................................,;... 1-1 1.3 Azimuthal Power Tilt....................................... 1-1 1.4 Channel Calibration.......................................... 1-1 1.5 Channel Check..... o ~ ~ ~ ~ ~ ~ p ~ ~ ~ ~ ~ ~ ~ ~ ~

1.6 Channel Functional Test...........................:.......... 1-2 1.7 Containment Vessel Integrity....... ... ... . .............. 1-2 lo8 Controlled Leakage........................................... 1-2

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1 p5 1.11 E Average Disintegration 'Energy.....'......................... 1-3 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... ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 1.18 Offsite Dose Calculation Manual (ODCM)........... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 4 1.19 Operable - Operability....................................... 1-5 1.20 Operational Mode - Mode....................................... 1-5 1.21 Physics Tests..... ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ o ~

1.22 Pressure Boundary Leakage .... o ~ ~ ~ ~ ~ ~ ~ ~ 1 5 ST. LUCIE - UNIT 1 Amendment Ne. 27,Xl NPP.

~ ~ ~ ~ y t INDEX AOM IN I STRATIVE CONTROLS SECTION PAG 6.6 R PORTAB VENT ACTION................:............ .. ~ ~ 6-12 6 1 SAFETY LIMIT VIOLATION................................. 6-12

6. PROC R PROGRAMS............................ .... 6-13 6.9 R POR R REMENTS 6.9.1 ROUTINE REPORTS...................................,... 6-15 1 Startup Report.................... ... 61Kb I Annual Reports........................................ 6-16 Monthly Operating Reports............;...... 6-16a Annual Radioactive Effluent Release Report.. 6-17 Annual Ra ol ic E i ment l eratin Ort o ~ ~ 6-1 Cora. O aWo+6 4m'~ gg 6,~ ecygg.g ~ ~ o \ \ ~ o &-('I 6 .9..2 SPECIAL REPORTS.'.............

6.10 R 0 R NTION............... ooo ~ ~ o ~ o ~ oooo ~ o ~ ~ o ~ ~ ~ ~ ~ oo 6 20

6. RAG ATION PR TECTION PROGRAM... ~ 000 ~ ~ ~ o@00 ~ ~ ~ ~ ~ 000 ~ ~ ~ ~ ~ ~ 621 6.12 HIGH ATION AREA....-.........-......---..-..-.-.---- 6-22 6.13 PROCESS CONTROL PROGRAM.............-...--.---..-.-..-.- 6-23 6.14 OFFSIT OS CALCULATI N MANUA ........

ST. LUCIE - UNIT 1 o d tll

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1.6 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.

1.7 CONTAINMENTVESSEL INTEGRITY shall exist when:

a. All containment vessel penetrations required to be dosed during accident conditions are either.

1. Capable of being dosed 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 as provided in Table 3.6-2 of Specification 3.6.3.1,

b. All containment vessel equipment hatches are dosed 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.

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

1.9 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.

Q+56g,<-p sAe~d oA HGrY fhG.E ST. LUCIE - UNlT 1 1-2 Amendment No. 69,

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment CoreO eratin LimitsRe ort COLR INSERT - A art of L-96-298 Attachment 3 CORE OPERATING LIMITSREPORT COLR 1.9a The COLR is the unit-'specific document that provides cycle specific parameter limits for the current operating reload cycle. These cycle-specific paramctcr limits shall be dctcrmined for each reload cycle in accordance with Specification 6.9.1.11. Plant operation within these limits is addressed in individual Specifications.

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REACTIVITY CONTROL SYSTEMS ITIawkttined withe 4he hen+s speafied MODERATOR TEMPERATURE COEFFICIENT i~ +he. C LR. "Che. wag,~@~

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LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall

a. Less positive than +7 pcm/'f whenever THERMAL POMER is < 7N of RATED THERMAL POMER,

b. Less positive than +2 pcm/'F whenever THERMAL POMER is > 70K of RATED THERMAL POMER, and APPLICABILITY: MODES 1 and 2*8 ACTION:

With the moderator temperature coefficient outside any one of the above limits, 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 4.1.1.4.1 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 cenparison with the above limits.

• Mith K ff > 1.O.

fSee Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 1-5 Amendment Nn. gF, IP,PN

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REACTIVITY CONTROL SYSTEMS FULL LENGTH CEA POSITION Continued)

LIMITING CONDITION 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 ( 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 wh~le maintaining the allowable CEA sequence and insertion limits shown on 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 STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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 COLR shown in Figure 3.l-la.

With one full-length CEA misaligned from any other CEA in its group by 15 or more inches beyond the time constraints shown in Figure 3.1-1a, reduce power to < 70K of RATED THERMAL POWER prior o completing ACTION f.l or f.2.

Restore the CEA to OPERABLE status within its specified align-ment 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 oq Fi ure 3.1-2; the THERMAL POWER level shall be restricte pursuant o Specification 3.1.3.6 during subsequent oper'ation.

ST. LUCIE - UNIT 1 3(4 1-21 Amendment No.

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R A IVITY CONTROL SYSTEMS F LENGTH C POSITION Conti nued LIMITING CONDITION FOR OPERATION Continued b) The SHUTDOMN 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 />.

Mith more than one full length CEA inoperable or misaligned from any other CEA in its group by'5 inches (indicated position)'r more, 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 />.

h. kith 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.

SURVEILLANCE RE UIREMENTS 4.1.3.1.1 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 />.

4.1.3.1.2 OPERABLE by Each inserting it't leastnot7.5fully inserted full-length CEA inches shall at least be determined once per 92 days.

to be 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 misaligned 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'pecification 3.1.3.6 and that the

,circuit prevents the regulating CEAs from being inserted beyond the Power Dependent Insertion Limit of Figure 3.1-2:

• a. Prior to each entry into MODE 2 from MODE 3, except that such veri&cation 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 NODE 2 from MODE 3 made in association with a measurement of power defect.

ST. LUCIE - UNIT 1 3/4 1-22 Amendment No. 4+,89, 74,P@

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REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS ~lied iwWe. CtR.

LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating CEA groups shall be limited to the withdrawal sequence and.to the insertion limits elan-e (regulating gJa~

tdd Id d t I y 1ly 1thd

-when withdrawn to at least 129e0 inches) with CEA insertion between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits restricted to:

a. < 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

c. < 14 Effeccive Full Power Days per calendar y'ear.

APPLICABILITY: .NODES 1* and 2*$ .

ACTION

a. Mith 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:

Restore the regulating CEA groups to within the limits, or

2. Reduce THERMAL POMER to less than or equal to that fractsan of RATED THERMAL POMER which is allowed b the CEA ro position ep648~~ov ttp8'. <<azh igsagstsn ltggtH s cl$ te4 tn +ha, CC7LR

b. Mith 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 tffe provisions of ACTION items

c. and d. of Specification 3.1.3.1, operation may proceed pro-vided either:

1. Th ih tT it dyi I I tt

~~g-8 are not exceeded, or

2. Any subsequent increase in THERMAL POMER is restricted to < 5X of RATED THERMAL POMER per hour.

'I See Specia1 Test Exceptions 3.10.2 and 3.10.5.

g Mith K eff ST. LUCIE - UNIT 1 3/4 1-28 Amendment No. Pf

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3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 :The linear heat rate shall not exceed the limits Mown-en-f~

mpeci(ical'i~ %ha, CC> LR APPLICABILITY: NODE 1 .

ACTION:

Hith the linear heat rate exceeding its limits, as indicated by four or more coincident incore channels or by the AXIAL SHAPE INDEX outside of the power dependent control limits of Figure 3.2-2, within 15 minutes corrective action to reduce the linear heat rate to within the 'nitiate limits and either: ea~

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

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

SURVEILLANCE REQUIREMENTS 4.2.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.1.2 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with

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either the excore detector monitoring system or with the incore detector monitoring system.

4.2.1.3 Excore Detector Monitorin S stem - The excore detector moni-toring system may be used f it rin he ee~ew v44en by:

4oeox e~t vWe Verifying at least once per ours 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 AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limits shown on Figure 3.2-2.

CoL ST. LUCIE - UNIT 1 3/4 2-1 Amendment No.

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c. Verifying that the AXIALSHAPE INDEX is maintained within the allowable limits o Igure 3.2-2, where 100 percent of maximum allowable power

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represents the maximum THERMALPOWER allowed by the following expression:

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 THERMAL POWER as determined by the F, curve of Figure 3.2-3.

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4.2.1.4 ' The incore detector monitoring system may be used for monitoring the coo~ewe istRbutioa by verifying that the incore detector Local Power Density alarms: Ligecxv. he~ vote.

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 Figure 3.2-1.

COLA.

0 lf 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. +V,M,GB, 69,65. W,~,~,~8

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7.0 UNACCEPTABLE OPERATION 3.6. 0 gg gA 15.0 14.0 AC OPERATION 13 0 FIGURE 3 2-1 PEAK LINEAR HEAT RATE VS BURNUP

>" Pages 3/4 2-4 (~endmcnt 106), 3/4 2-5 (Amendment 63), and 3/4 2-6 through 3/4 2-g (remend ent 109).

have been deleted from the Tcchnical Specifications. Thc next page is 3/4 2-9.

ST. ZUCXE UNIT 1 3/4 2-3 AMENDMENT NO. R7,82>M~TH.7%~ 4

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POWER OISTRIBUTION LIMITS OELETEO ST. LUCIE - PIT 1- Amendment No.-27-,N, 109,

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POWER DISTRIBUTION LIMITS TOTAL INTEGRATED RADIAL PEAKING FACTOR - FT LIMITING CONDITION FOR OPERATION 3.2.3 The calculated value of FT shall be mA4hIn Ate 4'ncIAm +ci(ied ce f4e COLR APPLICABILITY: MODE 1*.

ACTION: Aa't &Naia (hMy95 Mith FT , 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 THERMAL POWER to bring the combination of THERMAL POMER and F to within the limits of Figure 3.2-3 and withdraw the full ength CEAs to or beyond e Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THERMAL POWER limit determined ro igure 3.2-3 shall then be used to establish a revised upper THERMAL POWER 1 vel limit o Figure 3.2-4 (truncate Figure 3.2-4 at the allowable fraction o RATED THERMAL POWER determined by Figure 3.2-3) and subsequent operation shall be maintained within the ".educed acceptable operation region of Figure 3.2-4.

SURVEILLANCE RE UIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not, applicable.

4.2.3.2 FT shall be calculated by the expression FT = Fr(l+Tq) when F is calculated with a non-full core power distribution analysis code and squall be calculated as FT = Fr when calculations are performed with a full core power distribution analysis code. Fr shall be determined to be'within its limit at the following intervals.

a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,

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

c. Within four hours if the AZIMUTHAL POMER TILT (Tq) is ) 0.03.

• See Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 2-9 Amendment No. g7, Ng, AN,~

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0 POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION 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. AXIAL SHAPE 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 THERMAL POWER to ( 5X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.5.1 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 />.

4.2.5.2 The Reactor Coolant System tqtal flow rate shall be determined to be within its limit by measuremen~t least once per 18 months.

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~Ora~tfn Cold Leg Temperature ~549 F Pressurfzer Pressure ~ 2225 psia~

Reactor Coolant Flow Rate p 345,000 gpm AXIAL SHAPE INDEX Ffgure 3.2-4 COLE Lfmft not applfcable durfng efther a THERNL POMER ramp fncrease fn excess of 5X of RATED THERMA!. POMER or a THERNL POWER step fncrease of greater than 10% of RATED THERMAL POWER.

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3/4. 9 REFUELING OPERATIONS ouikhia We heeH steceeIIied.

BORON CONCENTRATION ~~+4< coLR, LIMITING CONDITION FOR OPERATION 3.9.1 With'he reactor vessel head unbolted or. removed, the boron concentration of all filled portions of the Reactor Coolant System and the refuelin cavity shall be maintained

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a. a Keff of 0.95 or less, which includes a 0 m conservati ce for uncerta

b. A boron ion of > 1720 ppm, es a 50 ppm ervative allowance for uncertaintie .

APPLICABILITY: MODE 6*.

ACTION:

Mith the requirements of the above specification not satisfied, imediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at > 40 g of 1720 boron or its equivalent u ~oron eoneen e eel e eee ppl e SURVEILLANCE RE UIREMENTS

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Ae. bororl ~Amenti'&ion llwlH 4.9.1.1. lI a nN4iens sha11 be determined prior to:

a~ Removing or unbolting the reactor vessel head. and

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

4.9.1.2 The boron concentration of the refueliag cavity shal'f 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 NODE 6 when the reactor vessel head is unbolted or removed.

ST. LUCIE - UNIT 1 3/4 9-1 Amendment No. ++

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3 4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 ADORATION CONTROL 3 4.1.1.1 and 3 4.1.1.2 SHUTDOMN MARGIN A sufficient SHUTDOMN NARGIN 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 1imits, and 3) the reactor will be maintained sufficiently subcritica1 to preclude inadvertent criticality in the shutdown condition.

SHUTDOMN MARGIN requirements vary, throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T . The most r estrictive condition occurs at EOL,'with T at no $ Nd operating temperature, and is associated with a postu)Hed 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. Mith T < 200'F, the reactivity transient resulting from a boron dilBfon event with drained Reactor Coolant. System requires a 2000 pcm SHUTDOWN a'artially MARGIN and restrictions on charging pump operation to provide adequate protection. A 2000 pcm SHUTDOWN NARGIN is 1000 pcm conservative for Node 5 operation with total RCS volume present, however LCO 3.1.1.2 is written conservatively for .simplicity.

3/4.1.1.3 BORON DILUTION AND ADDITION A minimum flow rate of at least 3000 GPN provides adequate mixing, prevents strat'.fication and ensures that reactivity changes will be gradual during boron concentration changes in the Reactor'oolant System.

A flow rate of at least 3000 GPN 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 NTC a.55UW +loAS limiting values for the in the accident III~

The NTC used and ansient anal es we 0 ERNA~ME~e~~X ERlllt~llSI~

Determination of MTC at the specified conditions ensures that the maximum positive and/or negative values of the NTC will not exceed the limiting values.

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REACTIVITY CONTROL SYSTEMS BASES 3 4..3 HOVAB CQNTRO ASSEHBL S 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 burnup,- 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 h that a DNBR SAFDL violation may occur during e-s -on our~ -e e CEA mis-alignment if this requirenent is not met. This potential DNBR SAFDL violation is eliminated by limiting the time operation is permitted at FULL POMER 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 reasons:

1. The margin calculations that support the Technical S ecificatio are based on a steady-state radial peak of F' . ~e u>ti scf-5 <~ Iec4~ 9.2.9.

Mhen the actual F< . significant additional margin exists.

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

This increase in F~ is caused by xenon redistribution.

ent anal sis can support allowing a misalignment-to exis

~e mut without correction, if the initial of the CGA position indicators (Specif tion 3.1.3.3) is F'perability required to determine CEA positions and thereby ensure mpliance with the CEA alignment and insertion limits and ensures proper ope tion of the rod block circuit. The CEA 'Full In'nd 'Full Out'imits p ide an additiona1 inde-pendent means for determining the CEA positions whe the CEAs are at either their fully inserted or fully withdrawn positions. Therefore, the ACTION statements applicable to inoperable CEA position ndicators permit continued operations when the positions of CEAs with inop able position indicators can be verified by the Full In" or 'Full Out" lim k(Ne. eog54dcue45 4+ Itmt+s of

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4 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.

INDEX'with The Excore Detector Monitoring System performs this function by continuously monitoring the g

AXIALSHAPE the OPERABLE quadrant symmetric excore neutron flux detectors

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and verifying that the AXIALSHAPE INDEX is maintained within the allowable limi I u

~ In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made: 1} the CEA insertion'limits of Specifications 3.1.3.5 and 3.1.3e6 are satisfied, 2) the AZIMUTHALPOWER TILT restrictions of Specification 3.2e4 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 limi u . 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.

r 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

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/4.9 'REFUELING OPERATIONS BASES 3/4.9.1 80ROH CONCENTRATION The limitatio on minimum boron concentration ensures that:

1) the reactor will remain subcritical during COR ALTERATIONS. and 2) a.

uniform boron concentration is maintained fcr r activity control in the water volumes havin direct access to the reactor vassal. The 'limitation oii K mf is sufficient to prevent r actor criticality with aff. fullf length rods shutdown and regulating) fully withdrawn.

3/4.9. 2 INSTRUMENTATION The OPERABILITY of 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 D'iCAY TIME The m'.nlmum requirement for reactor subcri"icality prior to movement of irradiated F'uel assemblies in the reactor pressure vessel ensures that sufficient time has el >psed to allow the radioactive decay of the short lived fission products. This decay time is consistent with .he assumptions used in the accident analyses.

3/4.9.4 CONTAINHENT PENETRATIONS The requiranents on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restnctions,are sufficient to restrict radioactive material.re ease from a fuel element rupture based upon the '1ack of containment pressurization potential while in the REFUELING f400E.

3 4.9.5 COfhl&NICATIONS The requirement for communicati'ons capability ensures that refueling sta-tion personnel can be promptly informed cf sigrtficant changes in the facility status. or core reactivity condi tion during COP.= ALTERATIONS.

3/4.9.6'QGPULATOR CRANE OPERABILITY Th 0?LRABILITY requirements of the cranes used for movement of fuel asseriblies ensures that: 1) each crane has su=ficient load capacity to list a fuel e'.~~cot, and 2) the, core internals and pressure vessel are protected from excessive lifting force in the event th y are i'nadvertently engaged during 1'.fing operations.

ST. LUCIE - UNIT 1 B 3/4 9-1 Amend-.ent No.~

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ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (Continued) 6.9.1.9 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.

6.9.1.10 At least once every 10 years, an estimate of the actual population within 50 iles of the plant shall be r re and submitted to the NRC.

SPECIAL REPORTS

~~ ~ Vexr Z. ~s 6.9.2 Special reports shall be submitted to the NRC within the time period specified for each report.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment Core 0 eratin Limits Re ort COLR INSERT-B Pa e1of2 artofI 96-298 Attachment3 6.9.1.11 CORE OPERATING LIMITSREPORT COLR 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 Moderator Temp'erature Coefficient Specification 3.1.3.1 Full Length CEA Position - Misalignment >. 15 inches Specification 3.1.3.6 Regulating CEA Insertion Limits Specification 3.2.l. Linear Heat Rate Specification 3.2.3 Total Integrated Radial Peaking Factor - F,~

Specification 3.2.5 DNB Parameters Specification 3.9.1 Refueling Operations - Boron Concentration The analytical methods used to dctcnninc thc core operating limits shall bc those previously reviewed and approved by the NRC, as dcscribcd in thc following documents or any approved Revisions and Supplements thereto:

1. WCAP-11596-P-A, "Qualification of thc PHOENIX-P/ANC Nuclear Design System for Prcssurizcd Water Reactor Cores," June 1988 (Westinghouse Proprietary)

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

3; XN-75-27(A), Rev. 0 and Supplcmcnt 1 through 5, "Exxon Nuclear Neutronics Design Methods for Prcssurizcd Water Reactors," Exxon Nuclear Company, Rev. 0 dated Junc 1975, Supplcmcnt 1 dated September 1976, Supplement 2 dated Deccmbcr 1980, Supplement 3 dated Scptcmber 1981, Supplcmcnt 4 dated Dcccmbcr 1986, Supplcmcnt 5 dated February 1987.

4. ANF-84-73(P), Rcv. 3, "Advanced Nuclear Fuels Methodology for Prcssurizcd 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 Scptcmbcr 1983.

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

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

8. XN-NF-8249(A), Rcv. 1 and Supplement 1, "Exxon Nuclear Company Evaluation Model EXEM PWR Small Brcak Model," Advanced Nuclear Fuels Corporation, Rev. 1 dated April 1989, Supplement 1 dated Dcccmber 1994.

I I St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment CoreO eratin LimitsRe ort COLR INSERT - B Pa e2 of 2 art of I 96-298 Attachment 3

9. XN-NF-78-44(A), "A Generic Analysis of thc 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 Scptcmbcr 1983.

11. EXEM PWR Large Brcak LOCA Evaluation Model as deflncd by:

a) XN-NF-82-20(A), Rcv. 1 and Supplcmcnts 1 through 4, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, all dated January 1990.

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

c) XN-NF-81-58(A), Rev. 2 and Supplements 1 through 4, "RODEX2 Fuel Rod Thcrmal-Mechanical Response Evaluation Model," Exxon Nuclear Company, Rcv. 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), Rcv. 0 and Supplcmcnt 1, "Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs," Exxon Nuclear Company, all dated January 1990.

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

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

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