ML17309A913

Proposed Tech Specs Modifying Specifications for Selected cycle-specific Reactor Physics Parameters to Provide Reference to St Lucie Unit 2 COLR for Limiting Values

ML17309A913
Person / Time
Site:	Saint Lucie
Issue date:	12/29/1997
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17309A912	List: ML17309A911 ML17309A913 ML17309A914 ML17309A915
References
NUDOCS 9801060187
Download: ML17309A913 (46)

Text

II St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Core 0 eratin Limits Re ort COLR ATTACHMENT3 to L-97-280 ST. LUCIE UNIT 2 IIJIARKED-UP TECHNICAL SPECIFICATION PAGES INDEX Page I INDEX Page XIX INDEX Page XXI Page 1-2 INSERT - A Page 3/4 1-5 Page 3/4 1-18 Page 3/4 1-19 Page 3/4 1-19a Page 3/4 1-20 Page 3/4 1-26 Page 3/4 1-28 Page 3/4 2-1 Page 3/4 2-2 Page 3/4 2-3 Page 3/4 2-4 Page 3/4 2-5 Page 3/4 2-6 Page 3/4 2-7 Page 3/4 2-9 Page 3/4 2-11 Page 3/4 2-12 Page 3/4 2-15 Page 3/4 9-1 Page B 3/41-4 Page B 3/42-1 Page B 3/4 9-1 Page 6-20 INSERT - B (4 pages)

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INDEX DEFINITIONS I,gg.'f. OPEIVailNG- iWllTo RCPORT (CC7LN)

SECTION AGE

1. 0 DEF INITIONS 1.1 ACTIONo ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ ~ ~ ~

1.2 AXIAL SHAPE INDEX.......

1.3 AZIMUTHAL POWDER TILT............

1. 4 CHANNEL CALIBRATION....

1.5 CHANNEL CHECK..................--- ~ ~ ~ o \ ~ ~ ~ ~ ~ o 1.6 CHANNEL FUNCTIONAL TEST... - .. -. - - -. - .. -. - - - ..- 1-2 1.7 CONTAINMENT VESSEL INTEGRITY. 1-2 1.8 CONTROLLED LEAKAGE...... - - - -- 1-2 1.9 T ERATIONooo ~ ~ ~ ~ ~ o ~ ~ ooo ~ ~ oooo ~ ~ o ~ oo ~ ~ ~ ~ ~ ooo ~ ~ ~ ~ ~ ~ 1-2 1 10 DOSE EQUIVALENT I-131...... 1-3 E-AVERAGE DISINTEGRATION ENERGY............. 1-3 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME FREQUENCY NOTATION. 1-3 1 14 GASEOUS RADAR/ASTE TREATMENT SYSTEM....... 1-3 1.15 IDENTIFIED LEAKAGE 1-3

1. 16 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE........... 1-4

1. 17 MEMBER(5} OF THE PUBLIC. - . - - .. - - - . -- 1-4 1.18. OFFSITE DOSE CALCULATION MANUAL (ODCM}.. o ~ o 1-4 1.19 OPERABLE - OPERABILITY...'- - - ~ ~ - ~ ~ ~ ~ ~ -- - ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ 1-4

1. 20 OPERATIONAL MODE - MODE . ~ ~ ~ ~ 1-4

1. 21 PHYSICS TESTS.............--.--- 1-4

1. 22 PRESSURE BOUNDARY LEAKAGE 1-5

1. 23 PROCESS CONTROL PROGRAM.......--oo 1-3 1.24 PURGE - PURGING. 1-5 1.25 RATED THERMAL POMER. 1-3

l. 26 REACTOR TRIP SYSTEM RESPONSE TIME. 1-3

1. 27 REPORTABLE QP<---..-. o ~

1.28 SHIELD BUILDING INTEGRITY ~ o o ~ o 1.29 SHUTDOWN MARGIN. 1-6

1. 30 SITE BOUNDARY.... ~ ~ ~ ~ ~ o ~ ~ 1-6 ST LUCIE - UNIT 2 Anendnen>> NQ.

ggg~TI N PAQg 6.5.2 COMPANY NUCLEAR REVIEW BOARD .. ...... ... 6-9 F UNCTION ~ ~ ~~~~~~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ ~ ~~~ ~ ~ ~~0 << ~00 ~ ~ ~ ~ 0 ~ eoet ~ 6-9 C OMPOSITION. ~ ~ ~~~~~~ ~ ~ ~~ ~~0 ~1 ~ ~~ ~~~~1 ~ ~~ \~ V ~

Al TERNATES ~ 0 ~ ~~ 10 ~ 10 ~ 0 ~ ~ ~0~~ ~~ ~~ ~ ~ ~ ~0 CONSULTANTS ~ ~0 0~0 ~~~ 0 ~1~ 0 ~ ~0~~0 ~ ~0 ~0 ~ ~ t ~~ ~ ~~ 6 1 0 M EETING FREQUENCY ....,......~......,....A....................,-------- ...6-10 Q UORUM0000 ~ oooeooes 10 0 110 ~ ootoo ~~0~ to ~ 0 ~ ao ~ ~ ~ ~ ~ ~ 00 to ~~0~ 0~ ~~ ~~~~~~~~~~~~~~ ~0~0~~~~~~~~~~~~~~~~ 100 ~~ .... 6-10 R EVIEWOO ~0~~~~~~~~~1~0~0~0~ 001 ~~~ ootoo ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ to ~ ~ eoeooooo ~ Oo ~ ~ \ ~ Oeeo ~ 1\0 ~ \ ~ ~0~ \ ~ 0 ~ ~ 0 ~ ~ ~ \<<te ~ ~ 0 ~~~~ ~~ 1 ~~0 ~~ 6 ~ 1 AUDITS 00 ~e~ ~~ 0 ~ ~ ~ ~ ~ ~~~~~~0 ~ 0~0~~ ~~ ~~0~~~~ ~ ~ ~ ~0~~~~~ ~~ ~ ~ 111 ~0 ~ 0 ~~ 1 ~0~ 6 ~ ~

AUTHORITY.......... ~ .. ~~~~~~0 ~~ \~0 ~ 0 ~ 1~ ~~~~~~~ 6 12 R ECORDS 0 ~ ~ ~ ~ ~ ~ 0000 ~ ~ 1 ~~~~~0~ *~ ~ ~~~ . 6-12 TECHNICALREVIEW RESPONSIBILITIES.......................................... .6-12 RP RAB A Tl N 0000 ~~~~~0~ 0000 ~ 001000 ~ 0 ~ ~ ~ ~ 000 ~ 0 ~ 1 ~ 00000 ~ ~ ooto ~ ooto ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ eeooeooeeooea 6 1 7 AF IMlTVl Tl ~ ~ ~ ~ 0 ~ 0 ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ \~0~~~ ~~~ 0 ~0 0 aors ~~ ~ 00 ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ 0 ~ ~ ~ ~~~~ ~~~~~~ ~ ~0~0 ~ ~ 0~~~~~ 6 13 PR D AND PR M ~0 ' ~~~~ ~0 ~~0 \~ ~ ~~~ ~0 1 ~ ~ ~0 ~ ~ 0 ~ 0 ....6-13 REPORTIN R M NT e~ ~1 ~ ~0 ~ ~ ~~ ~ \ ~ ~ ~~ ~0 ~~0~0 ~~~0 ~0 ~~ ~ ~ 6 16 6.9.1 R OUTINE REPORTS ot ~ ~ oooooootttooo ~ 0 ~ 0000 ~ ~ ~ ~ 0000 ~ 00 ~ 0000 ~~~0~1~~~0~ 0 ~ ootooo ~ ~ 0 ~ 0 ~ 0 ~ 0 ~ 000 ~ ~ 00 ~ ~ ~ ~ 00 ~~ too ~~~~ 6 1 6 S TARTUP REPORT................,...............................................................6 16 ANNUAL REPORTS .............,...............,........................................,..................,....,... 6-1 6 MONTHLYOPERATING REPORTS .... ..,.......................,....,.....,........., ~......,....,6-17

~

ANNUALRADIOACTIVEEFFLUENT RELEASE REPORT ...................................... 6-1 8 ANNUALRADIOLOGICALENVIRONMENTALOPERATING REPORT.....'............. 6-1 9 6.9.2 S PECIAL REPORTS ~~~~~ 0000000 ~~~~~~~~ 00001 ~ 00 ~ 0 ~ 0 ~ 000000 ~ ~~~ ~ 000 ~ 0000 000 ~ eooooo ~ otooooo ~ 000 ~ ~ ~ 000 ~ ~ ~ 000 000 v v

~0 ~0~0~ tto ~ 0 ~ ~ ~ ~ 0 ~ oto ~ 00 ~ 0 ~ 0 ~ ~ ~ ~ 0 ~ 00 ~ 1 ~ \ ~ 0 ~ ~ 0 ~ 00 ~ 0 ~ ~ ~ ~ 0 ~ 0 ~ 0 ~ 0 ~ 0 ~ oootoo ~ 00 ~ 0 ~ ~ ~ ~ ~ ~ 10 ~ 0 ~ ~~~~ \ 6+v

~ 00 ~ 0 ~ ~ ~ 1 ~ ~ ~~ ~ ~0 0~ 0 ~ 0 ~ 00 ~ 00 ~ 0 ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~~~~~~~0~~ 6 21

<<+ toqggPvgg genS Cfoez 4><<)

IIII ST. LUCIE - UNIT 2 XIX

~ ggg~~ Amendment No. W, SP,

.".CODEX

'! ST OF r .'GURES FlGURE PAGE 2.1-1 REACTOR CORE THERMAL HARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING..................... 2-3 2.2-1 LOCAL POWER DENSITY - HIGH TRIP'SETPOINT PART 1 lg (FRACTION OF RATED'HERMAL POMER VERSUS gkz) ..-.......... 2-7 2o2 2 LOCAL POMER DENSITY - HIGH TRIP SETPOINT PART 2 (gRp VERSUS Yq) ~ ~ ~ ~ ~ o o o ~ o ~ o ~ ~ ~ o ~ o o o ~ ~ o ~ o ~ ~ o ~ ~ o ~ ~ ~ ~ ~ 2-8 2.2-3 THERMAL HARGIN/LOM PRESSURE TRIP SETPOINT PART 1 (Yg VERSUS Aq) o ~ ~ o ~ o ~ ~ ~ ~ ~ o o ~ o ~ ~ ~ o ~ o ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~

2.2-4 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS ARE).--..-....... 2-10 B 2.1-1 AXIAL POMER DISTRIBUTION FOR THERMAL MARGIN SAFETY LIMITS.. B 2-2 3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AS A FUNCTION OF STORED BORIC ACID CONCENTRATION...................... 3/4 1-15 3.1-1a looo ~ o ~ o ~ ~ o ~ ~ ~

'( 3.i-r olhtCi'oBHfHlt+NR

-REAC reR-Ceem~UMPS-ePEaaRN .........................

3.2-1 URNU ooo ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~

3~2 2 3~2 3 me~eHeaeae eVa -AN pp Xyo0 ~ ~ ~

4.2-1 DELETED.

3.2-4 BEXMPHVRW UfHtEAHBR 558 uHBPMH o ~ ~ ~ o ~ ~ o ~ o o o o ~ o o o ~ o o o o o ~ ~ ~ ~ ~ ~ ~ ~ ~ o 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMITS VERSUS PERCENT OF RATED THERMAL POMER MITH THE PRIHARY COOLANT SPECIFIC ACTIVITY >1 uCi/GRAM DOSE EQUIVALENT I-131................;....................... 3/4 4-28 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 15 EFPY, HEATUP AND CORE CRITICAL................... 3/4 4-31a ST. LUC IE - UNIT 2 XXI Amendment No. S, ~

DEFINITIONS CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST 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.

CONTAINMENT VESSEL INTEGRfTY 1.7 CONTAINMENTVESSEL INTEGRITY shall exist when:

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

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 positions, except for valves that are open on an intermittent basis under administrative control.

b. Alf 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.

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

CORE Al TERATION 1.9 CORE ALTERATION shall be the movem'ent 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 ALTERATIONS shall not preclude completion of movement of a component to a safe position.

ST. LUCIE - UNIT 2 1-2 Amendment No. 86

0 St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Core 0 eratin Limits Re ort COLR INSERT - A Part of I 280 Attachment 3 CORE OPERATING LIMITS REPORT COLR 1.9a The COLR is the unit-specific document that provides cycle specific parameter limits for the current operating reload cyde. 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.

pi REACTIVITY CONTROL SYSTEMS Qa,ta(o.tlag bJSliirl '@4 t<i~~

Spc(/eh in. 4hz COLR. %e.

MODERATOR TEMPERATURE'OEFFICIENT ~a~jllkulVl fO5'I/OIL 4&lb '54Lli 4C LIMITING CONDITION FOR OPERATION 3.1,1.4 The moderator temperature coefficient (MTC) shall b :

.idled Less positive than +5 pcm/ F at < 70K RATED THERMAL POMER, sA~

b. Less positive than +3 pcm/ F at > 70K RATED THERMAL POME egal'-i-ve-than 3 ERM~OMER.

APPLICABILITY: MODES 1 and 2*$

PeLa it-ACTION:

Mith the moderator temperature coefficient outside any one of the above limits, 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 />.

SURVEILLANCE REQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. NTC measured values shall be extra'polated and/or compensated to permit direct comparison with the above limits.

4.1.1.4.2

~ ~ ~ ~ The MTC shall be determined at the following frequencies

~

and THERMAL POMER conditions during each fuel cycle:

a. Prior to initial operation

~

above 5X of RATED THERMAL POMER, after each fuel loading.

b. At any THERMAL POMER, within 7 EFPD after reaching a RATED THERMAL POMER equil-.brium boron concentration of 800 ppm.

c. At any THERMAL POMER, within 7 EFPD after reaching a RATED THERMAL POMER equilibrium boron concentration of 300 ppm.

• Mith K,<< greater than or equal to 1.0.

SSee Special Test Exceptions 3.10.2 and 3.10.5.

ST. LUCIE UNIT 2 3/4 1-5 Amendment No. k4;-BS,

REACTIVITY CQNTROL SYST:-MS 3/<. 1. 3 MOVABLE CONT ROL AS 5 EMBL'S E

CEA POSITION LIMITING CONDITION FOR OPERATION

3. 1.3." The CEA Block Circuit and all full-length (shutdown and regula ing)

CEAs which are inserted in the core, shall be OPERABLE with each CEA of a given group positioned within 7.0 inches (indicated position) of all other CEAs in its group.

APPLICABILITY: MODES 1" and 2".

ACTION:

a. With one or more full-length CEAs inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied wi hin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 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 />.

b. With the CEA Block Circuit inoperable, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

1. With one CEA position indicator per group inoperable take action per Specification 3.1.3.2, or

2. With the group overlap and/or sequencing interlocks inope. able maintain CEA groups 1, 2, 3 and 4 fully withdrawn and the CEAs in group 5 to less than '5.~ insertion and place and mairtain CEA drive sys m;n either the "Manual" or 'Off" oosizion, or

3. Be in at least HOT STANDBY.

C. With more than one full-length CEA inoperable or misaligned from any other CEA in its group by more than 15 inches (indicated position),

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 />.

d. With one full-lengtn Ci'A misaligned frdm any othe. CEA in its group by more than 15 inches, operation in NODES 1 and 2 may continue, "rovided that the misaligned CEA is positioned within 15 inches of the other CEAs in its group in accordance with the time constraints shown in Figure 3.1-1a.'

~.c X

See Special Tes- Exceptions 3.10.2, 3.10.4, and 3.10.5.

ST. LUCIE - iJNIT 2 3/4 1- jB Amendm~n. loci

1

TIVI

~~5: (Continued)

e. With one full-length CEA misaligned from any other CEA in its group by more than 15 inches beyond the time constraints shown i > igure 3.1-1 a, reduce power to s 70% of RATED THERMAL POWER prior to mpletin GT} .1 or e.2.

'+s GDLR.

Restore the CEA to OPERABLE status within its s ie a ignment requirements, or

2. Declare the CEA inoperable and satisfy SHUTDOWN MARGIN requirement of Specification 3.1.1.1. After declanng 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 7.0 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on+igure 3.1-2; the THERMAL POWER level shall be

~ ~

restricted purs ant to Spe 'on 3.1.3.6 during subsequent operation. Lg b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 houis.

Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With one or more fulldength CEA(s) misaligned from any other CEAs in its group by more than 7,0 inches but less than or equal to 15 inches, operation in MODES 1 and 2 may continue, provided that within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the misaligned CEA(s) is either.

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

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

a) Within 1 hour1.157407e-5 days <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 7,0 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown ongigure 3.1-2; the THERMALPOWER level shall be restricted purs hnt to Sp tion 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 the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> .

g. With one fulldength CEA inoperable due to causes other than addressed by ACTION a., above, and inserted beyond the Long Term Steady State Inseition 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.

• lf the pre-misalignment ASI was more negative than Z.15, reduce power to s 70% of RATED THERMALPOWER or 70% of the THERMAL POWER level prior to the misalignment, whichever is less, prior to completing ACTION eZ.a) and e.2.b).

ST. LUCIE. UNIT 2 $4 1-19 Amerxhnent No. 8

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KTIgE- (Continued)

h. Nth one full-length CEA inoperable due to causes other than addressed by ACTION a., above, but within its above specified alignment requirements and either fully withdrawn or within the Long Term Steady State Insertion Limits if group 5, operation in MODES 1 and 2 may continue.

in full-length CEA SURVEILLANCE RE UIREMENTS 4.1.3.1.1 The position of each full-length CEA shall be determined to be within 7.0 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 Each full-length CEA not fully inserted in the core shall be determined to be OPERABLE by movement of at least 7.0 inches in any one direction at least once per 92 days.

4.1.3.1.3 The CEA Block Circuit shall be demonstrated OPERABLE at least once per 92 days by a functinna1 test which verifies that the circuit prevents any CEA from being misaligned from all other CEAs in its group by more than 7.0 inches (indicated position).

4.1.3.1.4 The CEA Block Circuit shall be demonstrated OPERABLE by a functional test which verifies that the circuit maintains the CEA group overlap and sequencing requirements of Specification 3.1.3.6 and that the circuit prevents the regulating CEAs from being inserted beyond the Power Dependent Insertion Limit of 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, t

and

b. At least once per 6 months.

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

ST. LUCIE - UNIT 2 3/4 '1-20 Amendment No. 8;

REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION

3. 1.3.'6 The regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits shown on Figure 3 1-2 (regulating CEAs are con-sidered to be full withdrawn in ac ordance with Figure 3.1-2 when withdrawn to greater than or equal to 129.0 nches), with EA insertion between the Long Term Steady State Insertion Limi t tion Limits restricted to: 'ow and P we Dependent COt g

a. Less than or equa'l to 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. Less than or equal to 5 Effective Full Power Days per 30 Effective Full Power Day interval, and

c. Less than or equal to 14 Effective Full Power Days per calendar year.

APPLICABILITY: NODES 1" and 2"P.

ACTION:

a. Mith the regulating CEA groups inserted beyond the Power Dependent Insertion Limits, except tor surveillance testing pursuant to

~ ~

Specification 4.1.3.1.2, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

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

2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed b the CEA grou position b.

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Mith the regulating CEA groups inserted between the Long Term Steady g, '~< g~LQ State Insertion Limits and the Power Dependent Insertion Limits for intervals greater than 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, operation may proceed provided either:

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

2. Any subsequent. increase in THERMAL POWER is restricted to less than or equal to 5 of RATED THERMAL POMER per hour.

See Special Test Exceptions,3.'l0.2, 3. 10.4, and 3. 10.5.

Mith K ff greater than or equal to l. 0.

ST. LUCIE - UNIT 2 3/4 1"26

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I-0,20 ~(ONG TE(IM--~ -Sl T TE(llVI STEAOY EADY STATE STATE IN S E III ION

~ 0.10 INSEIITION I.IMIT LIMIT 3 I I

GllOlll'5 l 20 40 60 80 100 0 20 40 GO 80 100 0 20 40 IfNI (136)(100.0)(81.6) (51.4) (21.2) (0) (136)(108.8)(81.6) (51.4) (21.2) (0) (136)(108.8)(81.6) (54 (21.2) 0) 0 20 40 60 80 100 0 ?0 40 60 GII 100 (136) (108.8) (81.6) (51.4) (21. 2) (0) (136) (108.0) (81. 6) (54. 4) (21.2) (0) 5 CEA INSEIITION (INCIIES CEA WIT IltlllAWN)

3/4. 2 POMER DISTRIBUTION LIMITS 3/4 2;1 LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2. 1 The linear heat rate shall not exceed the limits APPLICABILITY: MODE l.

ACTlQM:

Mith 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 initiate cor" rective action to reduce th li he t rate to within the, limits and either:

Cong.

a. Restore the linear heat rate to within its limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or

b. Be in at least HOT STANDBY within the next 6 hours.

SURVEILLANCE RE UIREHENTS 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 either the excore detector monitoring system or with the incore detector monitoring system.

4.2.1.3 Excore Detector Nonitorina System The excore detector monitoring system may be used for monitoring the by:

<neo,t

a. Verifying at least once per,~M hours that the full-lenath 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 leas once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limit shown on Figure 3.2-2.

COLO.

ST. LUCIE - UNIT 2

c. Verifying that the AXlALSHAPE lNDEX is maintained within the allowable limits ofgRgure 3.2-2, where 100% of maximum allowable power represents the maxim ~m THERMAL POWER allowed by the following expression:

MxN CO LIZ.

where:

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

2. N is the maximum allowable fraction of RATED THERMALPOV/ER as determinedby the FY~ curve of Figure 3.2-3.

CGI P.

4.2.1A ' The'incore detector monitoring system may be used for monitoring the b e'n t 'heincoredetector Local Power Density alarms: /sneer k+ ~+~

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.

lf 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 2 3/4 2-2 Amendment No. M,

I/

'I

15.0

~~

~ <<f

~ ~ I

' I ul gg ~ p 14.0 ~ ~ f ~ 4

~l ~ ~

I I "UNACC EPTABLE OPERATION-Qw MO ZQ

~z 1S.0

~U ~ ~

QJ +

~ ~

C .ABLE OP ERATlON cZ

~ 12.0

~ ~ ~ ~ ~

~ ~ ~

11.0 BOL EOL Rgure 3Z-1 Allowable peak linear heat rate vs burnup t 4 3,

~/q z

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q. Q a

(~) g~y~ bedew

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'i ST. LUCIE - UNlT2 3/4 2-3

~'

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)

4

le 1.0 UNACCEPTABLE OPERATION REGIO

(-0.10, 0.08 (0 15, O.BB) 0.8 ACCEPTABLE QPERATIO REGIQ K

0.6 G ( o.5o. O.eo) (o.so, Q.eo>

0.4

-0. 6 0.4 -0. 2 Q.o 0.2 Q PERIPHE AXIAL PE INDEX (SOT MM LE HALO% 4 30 2 2 AXI L SHAPE INDEX VS. RA TI N PF flAXIi~Ut AL OMABLE POMER LEYEL PER S EC FICATION .~.1.3 Sj LU Am ndmen. %0. 4

r iuvi<r.

. O>mDLE cWB N Tl,.'~',lL Ppl<ER AND F, F I

I I I  !

I I I UN/lGCE!'TtlOL'E PIIOGTION OEGION I I I  !

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(>..7:, x.o). I o

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l. 65 I, 70 1. 7b 1.00 I~EA~UllEO F

THIS PAGE IHTE 0 LT GLAHK ST. CTE=U 3 endment-No. U'

~ \ ~

+ P7MUT QQ ~ ogf~ ~

3/4 2.2 TOTAL PLANAR RAD-'A~ P=AK~NG FAC!ORS - F XV LIMITING CONDITION . OR OPERATION

. 3.2.2 The calculated value of 'y shall F be APPLiCABILITY: MODE 1". Ai44e Spp.aig'acL

~w +la III%i+5 COLR ~

ACTIDN:

we wi~<z lim<W Mith F , within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

ao Reduce THERMAL POWER to bring +he combination of THERMAL POMER and to wi hin he limits of Figure 3.2-3 and withdraw the full length CEAs to or beyond the ong Tern Steady Sta.e Insertion Limits of Specification 3.1.3.6; or COLA

b. Be in HOT STANDBY.

SURVEILLANCE REOUIR~>ENTS 4.2.2.~ The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F Txy shall be calculated by the expression F T = F (:~T ) when xy F

xy is calculated with a non-full core power dis ribution xy q analysis code and shall be calculated as i = F when calculations are performed with a full xy xy core power distribution analysis code. F xv shall be determined to be wi hin i+s limit at the ollowirg intervals:

Prior to operation above 70'~of RATED THERMAL POWER after each fuel loading,

b. At leas once per 31 davs of accumulated operat'.on in MODE 1, and C. Wi hin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the AZIMUTHAL POWER TiLT (T ) is ) 0.03.

See Special Test Exception 3. i0. 2.

ST. LUC "" - UNiT 2 3/4 2-7 Amendment No-

~

~

Jl

~ r

POWER DISTRIBUTION LIMITS TOTAL INTEGRATED RADIAL PEAKING FACTOR - F LIMITING CONDITION FOR OPERATION 3-2.3 The calculated value of F, shall be APPLICABILITY: MODE I*. wi4io 4e t~'t~

m(eo<gecL tin+he COLS.

ACTION: co+ w)Aht~ (utes With F. . . within 6 hours either:

a. Be in at least HOT STANDBY, or

b. Reduce THERMAL POWER to bring the combination of THERMAL POWER and F to within the limits of Figure 3.2-3 and withdraw the full-C<L~ length,CEAs to or beyond the Long Term Steady State Insertion UNits of Specification 3.1.3.6. The THERMAL POWER limit deter-5 mined from Figure 3.2-3 shall then be used to establish a revised CO L~ . y u er THERMAL POWER level limit on Figure 3.2-4 (truncate Fi

.2-4 at the allowable fraction of RATED THERMAL POWER determined CQL~

cat P.

~ b Figure 3.2-3) and subsequent operation sbali be maintained within the reduced acceptable operation region of Figu cot P SURVEILLANCE RE UIREMENTS .

4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F, shall be calculated by the expression F, = F,(1+T,) 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! of RATED THERMAL POWER after each fuel loading,

b. At least once per 31 days of accumulated operation in MODE 1, and C. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the AZIMUTHAL POWER TILT (T,) is > 0.03.

• See Special Test Exception 3.10.2.

ST. LUCIE - UNIT 2 3/4 2-9 Amendment No ~Qd

I k

A r

n 1

r r.

news ST. LUCiiE - UNi7 7 3/4 2-11 amendment Ho.

I I o 1.0 U .:CE>TABLE  ! -O.OB,;. !0. ISA ~ ~ 40J OP- A.ioN S REGi CC CCEPTABLE g 0 OPERATION REGION c 0 6

!-o.so, 0 ) (O.SO, 0.6S) cC i 0 4

-G. 6 0.4 -0.2 0.0 0.2 O. 4 PERIPHERAL AXIA  :.;N =X, SOT APP I LR SELO 40% POSER)

FIG 3.2-4 AXIAL SHAPE X OPERATING LIMITS WITH FOUR REAC OR COOLANT PUMPS OPERATING LUCIE- 3/4 2-12 Am d ent No. 42

TABLE 3. 2-2 ONB MARGIN LIMITS FOUR REACTOR COOLANT PUMPS PARAMETER OPERATING Cold Leg Temperature (Narrow Range) 535 F" < T < 549 F Pressurizer Pressure 2225 psia"' <

PPZ<R

< 2350 psia" Reacto~ Coolant Flow Rate > 363,GOO-gpm I AXIAL SHAPE INOEX Figure 3.2-4

~nLR Applicable only if power level > 70~ RATED THERMAL POWER.

Limit not applicable during either a THERMAL POWER ramp increase in excess of SX of RATED THERMAL POWER or a THERMAL POWER step increase of greater than 10K of RATED THERMAL POWER.

ST. LUCIE - UN?T 2 3/4 2" 15 Amendme~t No I

3/4. 9 REFUELING OPERATIONS 3/4.9. 1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System an h r uel canal sha aintaine u d m.

APPLICABILITY MODE 6~.

&Ah a lee ~<wL4 Ie

- mcoL

<~equi ACTION:

Mith the requirements of the above specification not satisfied, -'mmediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater t a or equal to 40 m of a solution containing 1720 m boron or uce -

s - n-o~qu

+ ~~~~

"~ ~ I ~

4m~ <<oIAw~+~&icAA, Q mi'4IAIn <Lwi45,

~

~re~4eq SURVEILLA'NCE REOUIREMENTS 4.9.1.1 or~ ave-thea shall be determined prior to: 4C!P I x c roA COACea+Yah.tm I,<m<+

I i ~

aO Removing or unbolting the reactor vessel head, an

b. Mithdrawal of any full length CEA in excess of 3 feet from its fully inserted position within the reactor pressure vessel.

4.9. 1.2 The boron concentration of the reactor coolant system and the refue)ing canal shall be determined by chemical analysis at least once per 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 whenever fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed.

ST. LUCIE -'UNIT 2 3/4 9-1 kNCN/me~I'

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES Continued Overpower margin is provided to protect the core in the event of a large misalignment (> 15 inches) of a CEA. However, this misalignment would cause distortion of the core power distribution. This distribution may, in turn, have a significant effect on (1) the available SHUTD01A MARGIN, (2) the time-dependent long-term power distributions relative to those used in generating LCOs and LSSS setpoints, and (3) the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement ass'ociated with the large misalignment of a CEA requires a prompt realignment of the misaligned CEA.

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 bring 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 to reduce power in certain time limits depending upon. the previous requirement 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 second hour after the CEA misalignment this requirement is not met. This potential DNBR SAFDL violation is eliminated if 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. This time allowed for continued operation at a reduced power level can be permitted for the following reasons:

+4. ((m'g~s S act 1c~kion ~ < ~ +ha. Lni<ar S eoi

1. The margin calculations which support the Technical Specifications are based on a steady-state radial peak of F r

2. When the actual F<, 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 r is caused by xenon redistribution.

5. The present ana1ysis can support a11owing a a i n t to exist Ci without correction, if the

~ca.fe +iWa. Coy g~sn.t-S av h'Wtki~l ry YC QLQ Fi um S.l. La are W~,

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

Q ~ ~ II ra 0

II'

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 are 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 INDEX with the'OPERABLE quadrant symmetric excore neutron flux detectors~i~L" and verifying that. the AXIALSHAPE INDEX is maintained within the allowable limits of

+Lg Figure 3.2-2. 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 PLANAR RADIALPEAKING FACTOR does not exceed the limits of Specification 3.2.2.

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

'igure 3.2-1. The setpoints for these alarms include allowances, set in the conservative directions, for (1) a measurement~lculational uncertainty factor, (2) an engineering uncertainty factor, (3) an allowance for axial fuel densification and thermal expansion, and (4):a THERMAL POWER measurement uncertainty factor.

kaattlLtwd8 The limitations on FT 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 remain valid during operation at the various allowable CEA group insertion limits. The limitations on F~ and T, are provided to ensure that the assumptions used in the analysis establishing the DNB Margin LCO, the Thermal Margin/Low Pressure LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. If F~~, F~ or T, 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 ST. LUCIE- UNIT2 B 3/4 2-1 Amendment No. M,

3/4.9 REFUELING OPERATIONS BASES 3/4. 9. 1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a .

uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The value for Keff includes a ff 2X delta k/k conservative allowance for uncerta nties. Similarly, the bo~on concentration value ncludes a conservative uncertainty allowance of 50 ppm boron.

~pg~jfjgJ lA AIR COLA 3/4.9.2 INSTRUMENTATION The OPERABILITY of the startup 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 TINE 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 safety analyses.

3/4.9. 4 CONTAINMENT BUILDING PENETRATIONS The requirements 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 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.

ST. LUCIE - UNIT 2 B 3/4 9"1 ~enJ~aA +.

II ADMINISTRATIVE CONTROLS C

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 miles of the plant shall be prepared and submitted to the NRC.

~5~5 ~ >~uJQ o~ Q~~L 4 PASSE, SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC within the time period specified for each report.

6. 10 RECORD RETENTION In addition to the applicable record retention requirements of Title IO, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least 5 years:

a ~ Records and logs of'nit operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections,.

repair and replacement of principal items of equipment related to nuclear safety.

C. All REPORTABLE EVENTS.

d. Records of surveillance activities; inspections and calibrations required by these Technical Specifications.

e. Records of changes made to the procedures required by Specification 6.8.1.

ST. LUCIE UNIT 2 6-20. Amendment No. 43-;Hi

St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Core 0 eratin Limits Re ort COLR lNSERT-B Pa e1 of 4 Partof L-97-280 Attachment3 6.9.1.11 CORE OPERATING LIMITS REPORT 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 Temperature Coefficient Specification 3.1.3.1 Movable Control Assemblies - CEA Position Specification 3.1.3.6 Regulating CEA Insertion Limits Specification 3.2.1 Linear Heat Rate Specification 3.2.2 Total Planar Radial Peaking Factors - F~~

Specification 3.2.3 Total Integrated Radial Peaking Factor- F,~

Specification 3.2.5 DNB Parameters - Axial Shape Index Specification 3.9.1 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 Proprietary)

2. NF-TR-95-01, "Nuclear Physics Methodology for Reload Design of Turkey Point & St.

Lucie Nuclear Plants," Florida Power 8 Light Company, January 1995.

3. CENPD-199-P, Rev. 1-P-A, "C-E Setpoint Methodology: CE Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems," January 1986.

4. CENPD-266-P-A, "The ROCS and DIT Computer Code for Nuclear Design," April 1983.

5. CENPD-275-P-A, "C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers," May 1988.

6. CENPD-188-A, "HERMITE: A Multi-Dimensional Space - Time Kinetics Code for PWR Transients," July 1976.

7. CENPD-153-P, Rev. 1-P-A, "Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed Incore Detector System," May 1980.

8. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 1: C-E Calculated Local Power Density and Thermal Margin/Low Pressure LSSS for Calvert Cliffs Units I and II," December 1979.

St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Core 0 eratin Limits Re ort COLR INSERT- B Pa e 2 of 4 Part of L-97-280 Attachment3 CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 2:

Combination of System Parameter Uncertainties in Thermal Margin Analyses for Calvert Cliffs Units 1 and 2," January 1980.

10. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 3: C-E Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for Calvert Cliffs Units 1 and 2," February 1980.

CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 1981.

12. Letter, J. W. Miller (NRC) to J. R. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No. 8 to NPF-16 and SER), November 9, 1984 (Approval of CEN-123(F)-P (three parts) and CEN-191(B)-P).

13. CEN-371(F)-P, "Extended Statistical Combination of Uncertainties," July 1989.

14. Letter, J. A. Norris (NRC) to J. H. Goldberg (FPL), Docket No. 50-389, "St. Lucie Unit 2 - Change to Technical Specification Bases Sections '2.1.1 Reactor Core'nd

'3/4.2.5 DNB Parameters'TAC No. M87722)," March 14, 1994 (Approval of CEN-371(F)-P).

15. CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986.

16. CENPD-162-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 1, Uniform Axial Power Distribution," April 1975.

17. CENPD-207-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 2, Non-uniform Axial Power Distribution," December 1984.

18. CENPD-206-P-A; "TORC Code, Verification and Simplified Modeling Methods," June 1981.

19. CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983.

20. CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report," July 1974.

21. CEN-161(B)-P-A, "Improvements to Fuel Evaluation Model," August 1989.

22. CEN-161(B)-P, Supplement 1-P-A, "Improvements to Fuel Evaluation Model,"

January 1992.

23. CENPD-132, Supplement 3-P-A, "Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS," June 1985.

St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Core 0 eratin Limits Re ort COLR INSERT- B Pa e 3 of 4 Part of L-97-280 Attachment 3

24. CENPD-133, Supplement 5-A, "CEFLASH-4A, A FORTRAN77 Digital Computer Program for Reactor Blowdown Analysis," June 1985.

25. CENPD-134, Supplement 2-A, "COMPERC-II, a Program for Emergency Refill-Reflood of the Core," June 1985.

26. CENPD-135-P, Supplement 5, "STRIKIN-II,A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977.

27. Letter, R. L. Baer (NRC) to A. E. Scherer (CE), "Evaluation of Topical Report CENPD-135, Supplement ¹5," September 6, 1978.

28. CENPD-137, Supplement 1-P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model," January 1977.

29. CENPD-133, Supplement 3-P, "CEFLASHQAS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident," January 1977.

30. Letter, K. Kniel (NRC) to A. E. Scherer (CE), "Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P," September 27, 1977.

31. CENPD-138, Supplement 2-P, "PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977.

32. Letter, C. Aniel (NRC) to A. E. Scherer (CE), "Evaluation of Topical Report CENPD-138, Supplement 2-P," April 10, 1978.

33 Letter, W. H. Bohlke (FPL) to Document Control Desk (NRC), "St. Lucie Unit 2, Docket No. 50-389, Proposed License Amendment, MTC Chan e from -27 cm to

~30 cm," L-91-325, December 17, 1991.

34. Letter, J. A. Norris (NRC) to J. H. Goldberg (FPL), "St. Lucie Unit 2- Issuance of Amendment Re: Moderator Temperature Coefficient (TAC No. M82517)," July 15, 1992.

35. Letter, J. W. Williams, Jr. (FPL) to D. G. Eisenhut (NRC), "St. Lucie Unit No. 2, 4, 1984.

36. Letter, J. R. Miller (NRC) to J. W. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cyde 2 License Approval (Amendment No. 8 to NPF-16 and SER), November 9, 1984 (Approval of Methodology contained in L-84-148).

37. Letter, A. E. Scherer Enclosure 1-P to LD-82-001, "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," December 1981.

(c me't ci )

l St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Core 0 eratin Limits Re ort COLR INSERT-B Pa e4of4 Partof L-97-280 Attachment3

38. Safety Evaluation Report, "CESEC Digital Simulation of a Combustion Engineering Steam Supply System (TAC No.: 01142)," October 27, 1983.

39. CENPD-282-P-A, Volumes 1, 2, and 3, Supplement 1-P, "Technical Manual for the CENTS Code," March 1994.

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.

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