Text

Proposed Tech Specs for Unit 1 Cycle 6 Reload
	ML20066E067
Person / Time
Site:	Catawba
Issue date:	01/09/1991
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20066E065	List: ML20066E063; ML20066E067; ML20066E070;
References
	NUDOCS 9101170266
	Download: ML20066E067 (152)
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9101170266 910109 r

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SAFETY LIMITS AND POWER DISTRIBUTION TS MARK-UPS i

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r SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 FIGURE 2.1-la. REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN,0PERATION., 2-2 F ic ue.E 2 1 -1 b KE 4 0-10R CO E S AF(:_ TyLi m iT _ t-C U ($ LOOPS , g

2. 2 LIMITING SAFETY SYSTEM SETTINGS ny 0 9 r R n li o )U -

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES SECTION ,

2.1 SAFETY LIMITS

_ 2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. B 2-2 l 2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINT5............... B 2-3 l

CATAWBA - UNITS 1 & 2 III

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICA8ILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL

-Shutdown Margin - T,yg > 200*F........................... 3/4 1-1 Shutdown Margin - T,yg i 200*F........................... 3/4 1-3 4 Moderator Temperature Coefficient........................ 3/4 1-4 Mi nimus Temperature f or Cri ti cal i ty. . . . . . . . . . . . . . . . . . . . . . 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 Flow Paths - Operating................................... 3/4 1-8 Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - Operating...........................:.r. 3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 ,~'

Borated Water Sources - Operating........................ 3/4 1-12 I-3/4.1.3 MOVABLE-CONTROL ASSEMBLIES Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00................... 3/4 1-16 Posi tion Indication Systems - Operating. . . . . . . . . . . . . . . . . . 3/4 1-17 Posi tion Indication System - Shutdown. . . . . . . . . . . . . . . . . . . . 3/4 1-18 Rod Orop_ Time............................................ 3/4 1-19

-Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Bank Insertion Limits............................ 3/4 1-21 1

3/4.2 POWER DISTRIBUTION LIMITS, 3/4.2.1 AXIAL FLUX OIFFERENCE. . .(.(,/.n i h t . .... ................. 3/4A2-1 m

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR -qF (t h. C.i.t.h......... 3/4 42-5 .[

CATAWBA - UNITS 1 & 2 IV ^n a h nt Nv. 74 (Unli,1) si nhat Nv. 0G(Unit 2) .

' LIMIT!M8 CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PA0E 3/4.2.3 ""CT0" 000L'."T 'MSTE" F L~.' ZTll f"O RISE NOT CHANNEL FACTOR. .W.oi.t .i.7. . NUCLEAR

.................... 3/4A2-9 ENTHAL .

3/4.2.4 QUADRANTPOWERTILTRATIO..(u,n.7..l.)................... 3/4A2-12 3/4.2.5 DN8 PARAMETERS..{..Llni.t..l.)............................ 3 /4 A 2 - 15

- TABLE 3.2-1 DN8 f &RAMET E R S . . .b n t I. .h. . . . . . . . . . . . . . . . 3/4,42-. . . . .16. . . .

A& &lG v A dukCT V 8A-0 jq p p _; y GTMlTNWIGt P6 mea?&cn Loc /S tIV CMMDM)5MCrol . . . ' . co 3/4.3 INSTRUMENTATION p)/d rf g 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TA8LE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM - -

INSTRUMENTATION.......................................... 3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-15 TABLE 3.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP01NTS........................... 3/4 3-27 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-37 ,

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................- 3/4 3-42 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations. . . . . . . . . . . . . . . . 3/4 3-51 TABLE 3-3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS..................................... 3/4 3-52 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-54 Movable Incore Detectors................................. 3/4 3-55

' Seismic Instrumentation..................................- 3/4 3-56 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-57 REQUIREMENTS.............................................

3/4 3-58 Meteorological Instrumentation........................... 3/4 3-59

. 3.;.g.i tjv.n..., ]Gnjt y CATAWBA - UNITS 1 & 2 -

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3/4.2.1 AXIAL FLUX DIFFERENCE (Unit 2).......................... 3/4 B2-1 {

'3/4.2.2 HEAT-FLUX HOT CHANNEL FACTOR'- FQ(z)2)........... (Unit 3/4 B2 5 ;

3/4.2.3 REACTOR COOLANT SYSTEM FLOW TATE AND NUCLEAR ENTHALPY RISE HOT CRANNEL FACTOR (Unit 2)........................ 3/4 B2-9 3/4.2.4 - QUADRANT POWER TILT RATIO (Unit 2 ) . . . . . . . . . . . . . . . . . . . . . . 3/ 4 B2- 12 !

3/4.2.5. DNB PARAMETERS (Unit 2)................................. 3/4 B2-15

. TABLE 3.2-1 DNB PARAMETERS (Unit 2)............................... 3/4 B2-16

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1 BASES 1

SECTION PAGE 3/4.0 APPLICABILITY.................. ,......................... , 8 3/4 0-1 3/4.1 ROCTIVITY CONTROL SYSTEMS 3/4.1.1 BJPATION CONTR0L.......................................... B ?/4 1-l' 3/4.1.2 BORATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1.

3/4.2.1 AXI A L F LU X O I F F E R EN C E . fl4 n.d . l.) . . . . B . .3/4

. . 2-1 3/4.2.2 and 3/4,2,3 HEAT FLUX HOT CHANNEL FACTOR end-MAC40R-HMOL: ANT-EYSTCM 10W -RAT" -AND NUCLEAR ENTHALPY RISE HOT CHAN N E L F ACTO R . f Ll ni.t. h . . . . .c . . . . . . .B.3./ 4. 2. -.2. . . . . . .

H GURE B 3/4 + 1 c T7FICAL INDICATc0 AXIAL rLUx DIFFERE '

Tu t a" A t n W : a . . . . . . . . . . . . . . . . . . . . . . . . . . 0. .3/4-: . . . . 3". . ". . ". .

3/4.2.4 QUADRANTPOWER'TILTRATIO(((

p nit....................... B 3/4 2-5 3/4.2.5 DNB PARAMETERS. B 3/4 2-6 Lu.td ihdtd(U wc\ D .W. l . . . . . . . . . . . . . . . . . . .

3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 3/4,3,3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLA>f LOOPS AND COOLANT CIRCULATION. .. ........ B 3/4 4-1 3/4.4.2 SAFETY VALVES................. ........................... B 3/4 4-1 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES....................... ..................... B 3/4 4-2 3/4.4.5 STEAM GENERATORS................... ...................... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE. . .. . .. .. . ... B 3/4 4 '

3/4.4.7 CHEMISTRY...... ..... ... ... .. . ... ............. . ... 3 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY........ ........ .... .................. B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............... .............. B 3/4 4-7 CATAWBA - UNITS 1 &.2. XIll

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3/4.2.1-AXIAL FLUX DIFFERENCE nit 2)........................B3/42-7 3/4.2.2 and 3/4.2.3 IIEAT FLUX !!0T CilANNEL FACTOR, -

and REACTOR COOLANT SYSTEM FLOW KATE AND NUCLEAR ENTilALPY RISE 110T CilANNEL FACTOR (Unit 2). . . . . . . . . . . . .B 3/4 2-9 3/4.2~.4QUADRANTPOWERTILTRATIO(Unit 2p....................B3/42-11

.3/4.2.5DNBPARAMETERSUnit2p...............................B3/42-12

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-10 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

't 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-kfor four loop operation.

t (t4n M Q evnct 2. . l -l h ( q 4 g 2) glICABILITY: MODES 1.and 2.

ACTIONi-Whenever the point defined by the combination of the highest operating ' loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require--

ments. of Specification 6.7.1. ,

REACTOR COOLANT SYSTEM PRESSURE

~

2.1.2 The Reactor Coolant System pressure snall not exceed 2735 ps'ig.

' APPLICABILITY: MODES 1, 2, 3, 4, and 5. ,

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

MODES 3,~4, and 5:

Whenever the. Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

L L

CATAW8A - UNITS 1 & 2 2-1 '

Figure 2,14 Reactor Core Safety Limits Four Loops in operation g 665 660y Unacceptable Operation ,

655 .

650 2400 psia 1

645t

-640- -

635-630--

p625* 2000 pata . .;

co

.y ro 620--

O i CC .

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.610-- 1915 psia 605- -

600; -

Acceptable 595; . Operation 590--

585--' ,

580I . ,

0.0- 0.2 - 0,4 - 0.6 0.6 1.0 1.2 Fraction of-Rated Thermal Power -

9 CATAWBA - UNITS 1 & 2 2.- 2

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i N t i l I i l l 1 -i- 1 i 1 ;N , i i l UN ACCEPTA8LE F e : l IN i l l l ,OPERATl9N,.. ;r-NiI i Ni i I i-i : I i- i =

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'0- 0.2 0.4 0.6 0.8 1.0 1.2 FR ACTION OF R ATED THERMAL POWER FIGURE 2.1-1 b i

REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION LA IV I T 2.

CATAWBA - UNITS 1 & 2 2- 2 a. ,

k. TA8LE 2.2.-1 g REACTOR TRIP SYSTEM INSTRtiMENTATION TRIP SETPOINTS m

y -TOTAL SENSOR' ALLOWANCE ERROR a

FLINCTIONAL UNIT. (TA) Z (S) TRIP SETPOINT ALLOWA8tE VALUE

<a 1. Hanual Reactor Trip H.A. N.A. N.A. N.A. N.A.

[ 2. Power Range, Neutron Flux fg,90/g

, a. High Setpoint 7. 5 ' .4-Sfr5~9 2 0 1109% of RTP* 1111.2 of RTP*

b. Low Setpoint 8.3 4-Ss-5 71 0 525% of RTP* $27.1% of RTP*

3. Power Range, Neutron Flux, 1. 6 0.5 0 <S% of RTP* with <6.3% of RTP* with High Positive Rate i time r.onstant a time constant 3 2 seconds 3 2 seconds

4. Power Range, Neutron Flux, 1. 6 0.5 0 $$% of RTP* with <6.3% of RTP" with High Negative Rate a time constant a time constant m 32 seconds 32 seconds A 5. Intermediate Range, 17.0 8.4 0 <25% of RIP *

<31% of RTP*

Neutron Flux

6. Source Range, Neutron Flux 17.0 10 . O <105 cps <1.4 x 105 cps

7. Overtemperature AT M t $. 'lY 5 .heeNote1 ee Note 2

~

8. Overpower AT 44.k }>%1.24 Q M1. 7" See Note 3 See Note 4 h g- 9. Pressurizer Pressure-Low 4.0 2.21 1. 5 31945 psig 31938 psig***

, 10. -Pressurizer Pressure-High 7. 5

.A-96= 0,y 0.5 12385 psig $2399 psig

[ ; 11. Pressurizer Water Level-High S.0 2.18 1.5 $92% of instrument $93.8% of instrument e o 1 fio span span m : c' 12. Reactor Coolant Flow-Low sl, TOV. PF.9 w 5: p 2c 5 'A g 1.q u Ig- 0.6 390% of loop minimum measured i g - W ._ ) of loop U I; minimummeasuredflow**[

flow a

EG

ia 33 *RTP = RATED THERMAL POWER W

~ ** Loop minimum measured flow = 96,900 gpm(Unif2). ) %250g pm (Unith <J

- - 1ime constants utilized in the lead-lag controller for Pressurizer Pressure-tow are 2 seconds for lead A and 1 second for lag. Channel calibration shall erisure that these time constants are adjusted to these vaines.

-Ef7=== eab}e=u;:= t 's inn of RID-Bvpan_ System.

__ = ._. - _

+

.bCi kWn; tono '\E4 g TABLE'2.2-1 (Continued)

E .tEACIOR' TRIP SYSTEM INSTRUMENTATION TRIP SETPOINIS

. 'TOIAL- .

SENSOR C .

ALLOWANCE- ERROR.

i'i

.-e FUNCTIONAL UNIT (IA) Z- (S) TRIP SETPOINI ALLOWABLE VALUE

m l ~ '13. Steam Generator Water o.

m Level' Low-Low. '

a. Unit 1 17 14.2 3.5 >17% of span >15.3% of span fron i from 0% to 30% ~0% to.30% RIP *-

RIP

increasing increasing linearly'-

linearly to to 138.3% of span

> 40.0% of span from 30% to 100% RIP

from 30% to 100%

RIP *

'lf

b. Unit 2 11.8 1.7 2.0 >36.8% of narrow 135.1% of narrow range span range span

14. Undervoltage - Reactor- 8.57- 0 1. 0 >77% of bus >76% (5016 volts)

~

Coolant. Pumps voltage (5082 .

volts) with a 0.7s response time kk 15. Underfrequency - Reactor 4.0 0 1. 0 >56.4 Hz with a >55.9 Hz ga Coolant Pumps D.2s response time -

aa gg 16. Turbine Trip hh PP

a. Stop Valve EH N.A. N.A.

N.A. 1550 psig >$00 psig-Pressure Low

w. b. Turbine Stop Valve N.A. N.A. N.A. 31% open' >l% open

"%CC Closure ER 17. Safety Injection Input .N.A. .N.A. N.A. N.A. N.A. ,

eo from ESFi .

SC

' li l l- 6: Alii,1 n " .

._.r _. _ , . --._-m.--- - - -- -

2-_ _-~; a - ._ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ = _ _ _ . _ _ _

I lO0 blitott EB b W l

< n TAbtE 2.2-1 (Continued)

M

> REACIOR 1 RIP SYSTEN INSTRUNENTATION TRIP SElPOINis M

" SENSOR 10TAL

ALLOWAMCE ERROR

,( S) 1 RIP $EIPOINT ALLOWABLE VALUE E FUNCTIONAL UNIT (IA) Z_

^

18. Reactor Trip System

[ Interlocks

" N.A. N.A. N.A. 31 x 10 88 amps >6 x 10 85 amps

a. Intermediate Range Neutron flux, P-6

b. Low Power Reactor Trips Block, P-7

1) P-10 input H.A. H.A. N.A. $10% of RIP * $12.2% of RIP

N.A. N.A. N.A. <10% RIP

Iurbine $12.2% RIP

Turbine j

2) F-13 input Impulse Pressure impulse Pressure Equiva:ent Equivalent "g

c. Powee Range Neutron N.A. N.A. N.A. 148% of RIP * $50.2% of RIP

Flux, P-8

d. -ower Range Neutron N.A. N.A. N.A. $69% of RIP * $70% of RIP" flax, P-9

e. Power Range Neutron N.A. N.A. N.A 310% of RIP

37.8% of RIra f lux , P -10 gg N.A. N.A. $12.2% of RIP" N.A. $10% of RIP

S$ f. Power Range Neutron

@@ flux, Not P-10

.. $12.2% RIP

Yurbine Turbine impulse Chamber H.A. N.A. N.A. 510% RIP

rurbine 5S g.

Pressure, P-13 Impulse Pressure Impulse Pressure zz Equivafent Equivalent N.A. N.A N.A. N.A.

US 19. Reactor Trip Breakers N.A.

nn N.A. M.A. N.A. N.A.

EE 20. Automatic irip and N.A.

?? Interlock Logic me

" RIP = RATED TillaMKL POWER

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=

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N*I ali 3 1 3* I4 1 s l - 4 .t "g 3 ,

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8*

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CATAWBA UNITS 1 & 2 27 ^

.'; = :,,. t.33 ys.r.': t o

t

n IABIE 2.2-1 (Continued)

$ . IAttE NOIATIONS (Continued)

E

> NOIE 1: (Continued)

I' $ 590.8*T (Nominal T,,g allowed by Safety Analysis);

g K3 = 0.001189; P = Pressurizer pressure, psig;

[

es P' = 2235 psig (Hominal ACS operating pressure);

5 = Laplace transform operator, s 8; .

the and f,(al) is a function of the indicated dif ference between top and bottoe detectors of i

power-range newtron los chambers; with (p. ins to be selected based en measured instremment response during plant SIARIUP tests such that:

, (i) for q g between -22.51 and -6.51, ,

Em f,(AI) = 0, where q and g g are percent RAIES IHEW4AL F0WER in the top and bottee halves of the core respectively, and q + g is total IHERetAt POWER in percent et RAIED THEigtAt POWER; (ii) for each percent that the magnitude of g g is more negative than -22.5%, the

- AI Irip 5etpoint shall be autenetically reduced by 3.151% of its value at RAIED a IHElWGAL POWER; and i! !

(iii) f or each percent that the magnitude gf q g is more positive than -6.5%. the el Irip *}

i gy Setpoint shall be automatically reduced by 1.641% of its value at RAIES INElW4Al F0WE R.

p n'-

HDIE 2: The chenael's mani Irip 5etpeint shall not exced its cownted trip setroint by

$k mere than 3. .

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f.D 4'*

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Lo mL T ABLE 2.2-1 (Continued)

D TCIIBTAT18its (c.atinued)

> NolE 3: OVikPOWER AI AI(1+1:5) ( 1 ) ( *15 ) ( I ) ( I )

L (I

5.3) ~ 1 ~ '*(#"}I t 3 (1*15)(I*15)1 2 4 i* ' (I ' E 5) 7 (I

Te5) T d

Where: ai = As detined in Isote 1,

= As defined in Itete I, u s , :: = As defined in tiete 1

~

I = As detined in siete 1

, ,,5

,, g3 = As detined in Ilote 1 n As defined in Ilote 1, AI, =

K. = 1.0704, K3 = 0.02/*F for increasing average temperature and 6 for decreasing average temperature,

dynamic g g = <The function

,sati n, generated by the rate-lag controller for I,

= line constant utilized in the rate-lag controller for I, , s, = 10 s,~

s, I -- = As detined in Ilote 1*

1+u5 I

= As defined in Ilote 1 j

o .IABLE-2.2-1 (Continued)

$ TABLE NOTATIONS (Continued)

I E

@ NOIE 3: (Continued)

=

1 .

' .c.

=

Ks 0.001707/*F for T > 590.8'F and K = 0 for T $ 590.8'F.

, 3 T = As defined in Note 1, .

j -

e. 1"' = Indicated I at RATED THERMAL POWER (Calibration temperature for as avg .

i instrumentation, 5 590.8*F),

i =

5 As defne1 in Note 1, and f2 (al) = 0 for all al.

[

i NOTE 4: The channel's maximp Trip Setpoint shall not exceed its cosouted Irip Setpoint by 7

~ more than % 2.8 Q . {

1

[

4

] D jm P

iln 3,i,l tPfP 1

. p 7;;?

El. E'. :

, JApp Litreglp hit 46d*(Wi& Bypass 4fs tomy i

?

_-., -. . - - . . . ~ . - . . . - - - _ - . - . . - - . . . . . . - - . . . . - . . -

m:owuaps .

/. ' Onif l 1~

q' 33 2 POWER DISTRIBUTION LIMIT $

3/4.2.1 AX!ALFLUXO!FFEMNC,jJF,0J

\ MMITINGCONDITIONFOROPERATION

@ a ceep s.bte /i,nie s :pecifled in +4s. Cott: ctw'r..vG umars a.rt27 (< cuth 3.2.1 TheindicatedAXIALFLUXOIFFERENCE(AFD)shallbemaintainedwithih

' a f a. the allowed operational space as specified in the CORE OPERATING LIMITS REPORT (COLR) for RAOC cperation, or f

M

b. within the target band specified in the COLR about the target flux j i difference during baseload operation, j AP?LICABILITY: H00E 1, above 50% of RATED THERMAL POWER.* (, uni + 1}

ACTION:

33cc.

ForbAOCoperationwiththeindicatedAFDoutsideofthelimits c.

specified in the COLR, (y

1. Either restore the indicated AFD to within the COLR limits within ?

15 ninutes, or .

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , fb. For Base Load operaeion above APL NO with the indicated AXIAL FLUX O!FFERENCE outside of the applicable target band about the target pe/, flux dif ference:

" 1. Either restore the indicated AFD to within the COLR specified I target band limits within 15 minutes, or HO

2. Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes, h{. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR. (.

/

pe /c c e.

See Special Test Exceptions Specification 3.10.2.

ND is the minimum a11ewable (nuclear design) power level for base load

- APL /1 operation and is specified in the CORE OPERATING LIMITS REPORT per 5, Specification 6.9.1.9. - (

CATAWBA - UNITS 1 & 2 3/4 A2-1 3:}i-g t y . ] jUgj} Q

....... - ,. ,~. .. - ,

f Undl f

POWER DISTRIBUTION LIMITS LIMITING CON 0! TION FOR OPERATION __

SURVE!LLANCE REQUIREMENTS 4.2.1.1 The indicated AF0 shall be determined to be within its limits during PCWER OPERATION above 50% of RATED THERMAL POWER by:

{

i

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and .

2) At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status,

b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereaf ter, when the AFD Monitor Alarm is inoper-able. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

~

c. The provisions of Specification 4.0.4 are not applicable.

4.2.1.2 The indicated AFD shall be considered outsics of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the limits.

[4.2.1.3 When in Base Load operation, the target axial flux difference of (

each OPERABLE excore channel shall be determined by measurement at least once Y per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable.-

4.2.1.4 When in Base Load operation, the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining (/

the target flux difference in conjunction with the surveillance requirements of Specification 3/4.2.2 or by linear i torpolation between the most recently mea-sured values and the calculated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable, k

--"--'"""'"^i

3/4 42- 2 CATAWBA UNITS 1 & 2 Q3;-;isi;f[ si;ji @j{ jj l

i I

Unib \

POWER DISTRIBUTION LIMITS 3/4.2.2 NEAT FLUX H0T CHANNEL FACTOR -

LIMITING CONDITION FOR OPERATION 1

4 (,x; f,6) Qm pos.n V 3.2.2 F(Z)shallbelimitedby[thefollowingrelationships:

n l

Fn (Z) $ f K(Z) for P > 0.5 P I (nam E) I QII) I K(Z) for P 1 0. $

Where: F TP = the Fg Limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT

)(

(COLR), !

r - -

" THERMAL POWER , and . i RATED THERMAL POWER fr a(,x,64 '

K(Z) = the normalized F (Z) f;r : ;n;r ::r; 5:i;ht

-specified in th R g, g ,

,,r, g, 4,,g4,yp,,,

APPt.!CABILITY: MODE 1. (lini4- l}

Asplus wN4 i A TION: f4(y,z yy At eA mw t ,

With F (Z) exceeding its_ limit: /

n a.

3 Reduce THERMAL POWER at least 3 for each 3qF (Z) exceeds the limit within 15 minutes and similarly_ reduce the Power Range Neutron l Flux High Trip Setpoints within the next-4 hours; POWER OPERATION l may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION L may proceed provided the Overpower AT Trip Setpoints (value of K4 ) have been reduced at least 1% (in AT span) for each 1% F (Z) exceeds the j .

limit, and f n J/. Identify and c arect the cause of the out of-limit condition prior to increasing THERMAL POWER above the reduced limit required by N a., above; THERMAL POWER may then be increased provided g(Z) is demonstrated through incore mapping to be within its limit, fa CXsCD

( k* N'x,y, e ) :- +h < on uwcd A u 6 !!5* h* l 'h* "" #** #at sJp m ,ts ss .speciN I Y #'#* b> -

-]

CATAWBA - UNITS 1 & 2 3/4 42-5 ^x ad:;nt N;. M (Unit n 0 * ;nd;;nt N;. 00 (Unit. :)

_ . ~ . . _ _ - _ - _

i for Specification 3.2.2 l

. j Attachment 1.

a. Reduce THERMAL POWER at least 1% for each in Te*(X,Y,2) exceeds the I limit within 15 minutes and similarly reduce the Power Range Neutron !

T1ux H1 5 h Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and b, Control the ATD to within new ATO limits which are determined by i reducing the allowable power at each point alon Specification 3.2.1 at least 14 for each in Tc*g the AFD (X,Y,2) exceeds limit thelines of

, limit within 15 minutes and reset the ATD alarm setpoints to the modified limits within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

c. POWER OPERATION say proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent !

POVER OPERATION.may proceed provided the (Norpower AT Trip Setpoints !

(value of X.) have been reduced at least it (in At span) for each it !

Tg*(X Y,2) exceeds the limit, and l I

J 3-- l l

r i 1

1

i I

l, p.

~

i

1 f,,A NY l l

POWER O!STRIBUTION LIMIT $

., SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.44renotapplicable,

?q%%nn wAs w i (* M 4.2.2.2 F;r n?.^0 :p;r;thn, is within its limit by: (2 shall be evaluated to determine [44 Fg (z)

a. Using the movable incere detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

(b. Increasing the measured F (2) component of the power distribution pg g

map by 3% to. account for manufacturing tolerances and further in-

]

creasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied.

c. Satisfying the following relationship:

RTP s M F Fg (2) 1 O

KI*) rP> .5 -

P x w(z)

RTP M F n , gg,)

Fg (t) 1 W(z),x 0.5 g, p -

,5 g whereFh(z)isthemeasuredF(2)increasedbytheallowancesfor 9

manufacturing tolerances 2.nd m6asurement uncertainty, F gRTP is the y Fg limit, K(2) is the normalized F (2) as a function of core height, l 9

P is the relative THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

FhTP , K(z), and W(z) are spec CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. [! ~

d. Measuring gMF (z) according to the following schedule: D

1. Upon achieving ecuilibrium conditions after exceeding by 10% or I

"f*"a , more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was 1sst determined,* or 9

2. At least once per 31 Effective Full Power Days, whichever occurs first.

t

Dur.ing power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a

( power distribution map obtained. f CATAW6A - UNITS 1 & 2 3/442-6

}gff.

. _ . . 3. . . .]. [J s .f..jt

. .Q.,

. , , . _ , r

for Specification 4.2.2.2-- ,

Attachment 21

b. Measuring Fn"(X,Y,Z) at the earliest of:

l '. -At least once per 31 Effective Tull Power Days, or 2, Upon reaching equilibrium conditions after exceeding by 10% or M

more of RATED THIRMAL POWER, the THERMAL POVER at which Pg (X,Y,2) was last determined (2), or

3. At each time the QUADRANT POWER TILT RATIO indicated by the excore detectors is normalized using incere detector measurements.

l 3

4 i

~

mNo additional uncertainties are required in the following equations for TgN (X,Y,Z), because-the limits include uncertainties. _

(2)During power escalation at.the beginning of each cycle, THERMAL POWER may ]

be increased until a power level for extended operation has been achieved and a power distribution map obtained. ,

9 ys p -, -

u

__ yes--tg., %,+ -

ip f

UdIh l POWER 0!STRIBUTION LIMITS SURVEILLANCE RE0V!REMENTS (Continued)

With measurements indicating

^

maximum Ff(z) over I d(z) i has increased since the' previous determination of qEF (z) either of A@c' the following actions shall be taken:

M%

pnLuk.ot '

3 1) F (z) snall be increased by 3 over that specified in Specification 4.2.2.2c., or

{ 2) F (2) shall be measured at least once per 7 Effective Full

! Power Days until two successive maps indicate that maximus F (2) is not increasing.

over Z K(z)

.. With the relationships specified in Specification 4.2.2.2c. above not being satisfied:

l 1) Calculate the percent gF (z) exceeds its limit by the following expression:

S maximum F (z) x W(z). 1 . x 100 fo r P >~ 0. 5 over 2 p RTP /

I q

0 x g(g) q i C -

F

)

l

' Qmaximus -

F (z) x W(z)., Sgj, x 100 f or P < 0. 5 l p ATP (over2 4

/

. 5

K(*) _ j /

2) One of the following actions shall be taken:

l a) Within 15 minutes, control the AFD to within new AFD limits I

which are determined by reducing the AFD limits of Specification 3.2.1 by 1% AF0 for each percent F q(z) exceeds its limits as determined in Specification 4.2.2.2f.1). f Within 8 hout;, reset the AFD alarm setpoints to these mod-ified limits, or b) Comply with the requirements of Specification 3.2.2 for j F (z) exceeding its limit by the percent calculated above, or /

N* t ' c) Ver_ify that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load operation.

CATAWBA - UNITS 1 & 2 3/4A2-7 ' e= t:r t ": . " '"" t 1)

n r.C;rA %. C '%it O

f for Sp3cificaticn 4~,2,2,2 .

Attachment 3:

c. Perferniing the following calculations:

1. For each location,, calculate the 4 margin to the maximum allowable design as follows:

4 Operational Margin - ( 1 -

II)

O

_.) x 1004 k

[To(X,Y,Z)15'

% P.PS Margin - ( 1 - IQ IX'Y'I)

_ ) x 1006 '

( Fe'(X , Y , Z ) )"

where [FoL (X,Y,Z)Ci - and [ Fo L (X,Y, Z)"Eare the operational and RPS design peaking limits defined in the COLR. .

I

2. Find.the minimum operational Margin of all locations examined in !

4.2.2.2,c.1'above. If any margin.is less than zero, then either:

of the following actions shall be taken:

(a) 'Within 15 minutes: ,

(1). Control the AFD to within new.AFD limits that are !

determined byt ,

eed

. (AFD' Limit) g y, - (AFD Limit)C g . .

+[NSLOPE/8F x Margin " ) absoluta-value !

(AFD Limit)

g - (AFD Limit) g y,

- [PSLOPE/83xMargin ) absolute value-

.where Margin " is the minimum margin from 4.2.2.2.c.1, and (2) Vichin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to the modified limits of 4.2.2.2.c.2 a, 'or' (b) . Comply with the ACTION requirements of Specification 3.2.2.

Defined and!specified in the COLR-'per Specification 6,9.1.9,

for Sp3cificotien 4.2.2.2 Attachment 3 (con't):

3.* Find the minimum RPS Margin of all locations examined in 4.2.2.2.c.1 above. If any margin is less than zero, then the following action shall be taken:

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , reduce the Kg value for OTAT by:

g3*dJusted . g3 (') . (gswrt t3) x Margin labsolute value I l

l vhereKARCIN"gf"istheminimummarginfrom4.2.2.2.e.l.

1 u)Dar ined and specified in the COLR per Specification 6.9.1.9.

("g3 value from Table 2.2 1.

}

l l

/

l

for Spacification 4.2.2.2 (Coo 49)

d. Extrapolating the two most recent measurements to 31 Effective Tull Power Days beyond the most recent measurement and if:

[To"(X,Y,Z)) (extrapolated) 2 (ToL (X,Y,2))c' (extrapolated), or (To"(X,Y,2)) (extrapolated) 2 (ToL (X,Y,2))W (extrapolated),

either of the following actions shall be taken:

1. Tn"(X,Y,2) shall be increased by 2 percent over that specified in 4.2.2.2.a. and the calculations of 4.2.2.2.c repeated, or

2. A movable incere detector power distribution map shall be obtained, and the calculations of 4.2.2.2.c.1 shall be performed no later than the time at which the margin in 4.2.2.2.c.1 is extrapolated to be equal to zero.

4

_A

U ni+ \

POWER O!STRIBUTION LIMITS l SURVE!LLANCE REQUIREMENTS (Continued)

The limits specifie's in Specifications 4.2.2.2c., 4.2.2.2e., and l i

i p

g

, l j. 4.2.2.2f. , above am not applicable in the following core plane regions:

A% 1.

2.

Lower core region from 0 to 15%, inclusive Upper core region from 85 to 100%, inclusive _._

~

4.2.2.3 Base Load operation is permitted at powers above APL# if the fonowing conditions are satisfied:

a. Prior to entering Base Load operation, maintain THERMAL POWER above ND APL and less than or equal to that allowed by Specification 4.2.2.2 for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain Base Lead operation surveillance (AFD within the target band about the target flux differ- f pte ence of Specification 3.2.1) during this time period. Base Lead operation is then permitted providing THERMAL POWER is maintained J ND OL ND between APL and APL or between APL and 100% (whichever is most limiting) and FQ turveillance is maintained pursuant to Specification 4.2.2.4. APL OL is defined as:

RTP <

i"I""* 0 x K(Z)

APLOL =

over Z ( ) x 100% -

)

F (Z) x W(2)gt ,

where: F (z) is the measured Fg (z) increased by the allowances for manufacturing to1 % nces and measurement uncertainty. is the FhTP Fq limit, K(z) is the normalized F g(Z) as a function of core height. ?

W(2)gg is the cycle dependent function that accounts for limited power distribution transients encountered during Base Load operation.

RTP F

g , K(z), and W(Z)gg are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. /

b. During Base Lead operation, if the THERMAL POWER is decreased below ND

! APL then the conditions of 4.2.2.3a shall be satisfied before re-entering Base Lead operation.

. 4.2.2.4 During Base Load OperationqF (Z) shall be evaluated to determine if Fq (Z) is within its. limit by:

a. Using the movable incore detectors to obtain a power distribution ND mao at any THERMAL POWER above APL ,

b. Increasing the measured F (Z) component of the power distribution map 9

by.3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied.

NO

.APL is the minimum allowable (nuclear design) power level for Base Load operation in Specification 3.2.1.

CATAW8A UNITS 1 & 2 3/4A2-7a Wacent L. M f; % ,an; L. 5"((Unit MLU&

for Specificatien 4.2.2.2

.teachment4:

e. The ituits in Specifications 4.2.2.2.c and 4.2.2.2.d are not applicable :

in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1. Lover core region from 0 to 156, inclusive.

2. Upper core region from 85 to 100t, inclusive .

4

(

l l

e

Onifl POWER 0!$TRIBUT!0N LIMITS

$URVE!LLANC,j REQUIREMENTS ,(g ntinued)

f. Satisfying the following relationship:

RTP N 0 Daletsj F9 (Z) i f,7 p 3 Apg ND p l p M

where: Fg (Z) is the measured F9 (Z). F is the F limit.

g 9 )

/

K(Z) is the normalized Fg (Z) as a function of core height. P is the relative THERMAL POWER. W(Z) is the cycle dependent function that accounts for limited power dibribution transients encountered during (

Base Load operation. RTP F

g , K(Z), and W(Z)gg are specified in tha (

CORE OPERATING LIMITS REPORT per $pecification 6.9.1.9.

)

d. Measuring F (Z) in conjunction witn target flux difference deter-mination ac ording to the following schedule:

1. Prior to entering Base Load operation after satisfying surveil-lance 4.2.2.3 unless a full core flux map has been,taken in the previous 31 EFP0 with the relative thermal power having been ND maintained above APL for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and

2. At least once per 31 effective full power days.

e. With measurements indicating maximus F (2) over I I K(zd i hasincreasedsincethepreviousdeterminationF((Z)eitherofthe following actions shall be taken: *

1. F (Z) shall be increased by 2 percent over that specified in

.4.2.2,4c, or

2. F (Z) shall be measured at least once per 7 EFPD until 2 [

successive maps indicate that maximus F (z) 3 si not increasing'.

l over z I K(YT l

f. With the relationship specified in 4.2.2.4c above not being satisfied, either of the following actions shall be taken:

1. Place the core in an equilibrium condition where the limit in s 4.2.2.2c is satisfied, and remeasure F (Z), or CATAWBA - UNITS 1 & 2 3/ 42 7b dme r4;e r.t % . 7 ("r.' t 1 F Z;%;'. L. 50 $$ C

- . . _ . .- _. . - . - - - =- - ~ . - . . _ - - . _ - _ - . . - --

hnik \

j POWER DISTRIBUTION LIMITS

$URVEILLANCE REQUIREMENTS (Centinued)

2. Comply with the requirements of $pecification 3.2.2 for helatt Fg (Z) exceeding its limit by the percent calculated with the following expression:

9 M n

((max. over z of ( F (Z) x W(Z)BL ) ) 1 ) x 100 for NDP > APL P

F x K(Z)

g. The limits specified in 4.2.2.4c., 4.2.2.4e., and 4.2.2.Af.

above are not applicable in the following core plan regions:

1. Lower core region 0 to 15 percent, inclusive.

2. Upper core region 85 to 100 percent, inclusive.

(2.2.5 WhenF(Z)ismeasuredforreasonsotherthanmeetingtherecuirements) g of Specification 4.2.2.2 an overall measured F (:) shall be obte ned from a power 9

distribution map and increased by 3% to account for manuf acturing tolerances further increased by 5% to account for measurement uncertainty.

Aspines. wiM Aet.c.s4 m ec.5 i

l i

CATAWBA - UNITS 1 & 2 3/4A2-7c *n nda nt Nv.

f n ada nt N 59(Unit li (U..;i. 2)~

i, for Specification 4,2.2.3 9-d.

l Attachment El

, 4.2.2.3 ithen a full core power distribution map is taken for reasons other

, than meeting the requirements of $pecification 4.2.2.2, an overall r o"(x,Y.Z) shall be determined, then increased by 34 to account for manufacturing i- tolerances, further increased by 54 to account for seasurement uncertainty, t and further increased by the radial local peaking factor to obtain a maximus- 'l n

local peak. This value shall be compared to the limit in Specificacion 3.2.2.

j !

r 4

, 'k i

a i

s F 1.

'.t r

l~

ll i

a

a -e., .,we ,.-w .-.w-..-..-n+,.w.,- ...e,..

--e w ,*.-.,,--r.,. ,3 -,-ww,-,--wey---,,cw.~~-.,-,--ne, y , - , .,-rw+"-..w._.yw-, wr y. -e- -y.e.,--*.-mw

UV$N \

F POWER DISTRIBUTION LIMITS l

f

$URLE!MANCE RQU_!_R,EMENTS._

j 4.2.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2.2 F,y shall be evaluated to determine if qF (2) is within its limit ey: '

4. Using the movable incore detectors to obtain a power distrioution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER, b.

.)

Increasing the measured F,y component of the power distribution map f :

by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, 0

c.

' Comparing above to:

the F,Y computed (F,Y)-obtained in $pecification 4.2.2.2.2b.,

1)' The F,y limits for RATED THERMAL POWER (r,R ) for the appropriate measured core planes given in Specification 4.2.2.2.2e and f.,

.below, and

2) The relationship: x} j /

f

. F, aF (1+0.2(1P)), I l

i { Where F*Y is the limit for fractional THERMAL POWER operation expressed as 4- function of F,RTP y and P is the fraction of RATED ~

THERMAL POWER at which F,y ".u measured.

d. Remeasur.ing F,y according to the following schedule {

h 1)' When F,C is greater than the F x R limit for the appropriate measured core plane but less than the F relationship, additional '

C

.powergistributionmapsshallbetaken d F*Y compared to F*U (

- and F xy either: Y

. a ) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL C '

POWER or greater, the THERMAL POWER at which

F,y was ;

last determined. or /

b) At 1. east once.per 31 EFPD,. whichever occurs - first.

Qe/eb k

} i 1

^ = 3:st O. l(Wit'll CATAW8A UNITS 1 & 2 3/442-7e

'-;,,;.,,,,; Ne. 5 (L a 2) -/ t

_ . _ _ _ . . . _ . _ . . . _ _ ..- __ ~ _ -. _ __ _ . . _ , _ _ _ _ . . - .

(A n i + l 3

POWER Ol$TRIBUT!0N LIMITS

$URVE!LLANCE RE0VIREMENT$ (Continued) i

2) When the F is less than or equal to the F '

imit for the appropriate measured core plane, additional power distribution S

mapsshallbetakenandF,fcomparedtoF and F, at least I or.ca per 31 EFPO.

)

e. T.he F,y limitt for RATED THERMAL POWER (F,E ) shall be provided for all core planes containing Bank "0" control rods and all enrodded core planes in a Radial Peaking Factor Liinit Report per Specifica-

/

tion 6.9.1.9; {

y {

f. The F,y limits of Specification 4.2.2.2.2e., above, are not applicable

{

in the following core planes regions as measured in percent of core i

. height from the bottom of the fuel: '

1) Lower core region from 0 to 15%, inclusive, i

' L

2) Upper core region from 85 to 100%, inclusive. h j i j

3) Grid plane regions at 17.8

2%, 32.1 2 2%, 46. 4 t 2%, 60. 6 2 2% l and 74.9 1 2%, inclusive, and I

' /

., 4)- Core plane regions within : 2% of core height (* 2.88 inches) about the bank demand position of the Bank "D" control rods.

f ,

g. With F exceeding F, , the effects of F,y on F9 (Z) shall be evaluated to determine if F 9 (2) is within its limits. -

1 4.2.2.2.3 When qF (Z) is measured for other than F,y determinations, an overall / !

measured qF (Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to ;

account for measurement uncertainty. ;

i 4

3 O I 8

=.= a =. : a..t a

hvYn k l

POWER O!STRIBUTION LIMITS 3/4.2.3 OCA070" 000LA"! OYST0" TLOU RATC 1"0 NUCLEAR ENTHALPY RISE H0T CHANNEL FACTOR - % er e r)

LIMITING CON 0! TION FOR OPERATION F3.2.3 The combination of indicated Reactor Coolant $ystem total flow rate and)

R shall be maintained within the region of permissible operation specified in the CORE OPERATING LIMITS REPORT (COLR) for four loop operation.

Where:

N

AH

a. R= '

M RTP A ct.nh,,c F3g [1.0 + MF3g 4 0 P)) ()

I

/

THERMAL POWER b* ,

P c.

= RATED THERMAL POWER FhaMeasuredvaluesofFhobtainedbyusingthemovableincere

()

detectors to obtain a power distribution map. The measured valuesofFhshallbeusedtocalculateRsincethefigure

{

specified in the COLR includes penalties for undetected feed- )

water venturi fouling of 0.1% and for measurement' uncertainties l of 2.1% for flow and 4% for incore measurement of Fh,

d. FhP=The FhlimitatRATEDTHERMALPOWER(RTP)specifiedinthe

(

COLR, and f.

mfg = The power factor multiplier specified in.the COLR.

APPLICABILITY:

MODE 1. (u n i4-h ACTION:

fa. Wii.h the combination of Reactor Coolant System total flow rate and R within the region of restricted operation within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reduce the Power Range (

Neutron Flux High Trip Setpoint to below the nominal setpoint by the same amount (% RTP) as the power reduction required by the figure specified in (

the COLR. )

('

' b. With the combination of Reactor Coolant System total flow rate and R within the region of prohibited operation specified in the COLR:

(8(('

a.~a

1. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a) Restore the combination of Reactor Coolant System total flow rate and R to within the region of permissible operation, or b) Restore the combination of Reactor Coolant System total flow rate and R to within the region of restricted operation and comply with action a. above, or f UNITS 1 & 2 3/4%~9 CATAWBA (59@9.}.

".. ," 2'j } U,

, . _ _-. . _ . . .~ _ _ _ . . _ _ _ , __

fer Specification 3.3.3 1

Attachment 1:

3.2.3 Pg(X,Y) shall be limited by irrposing t'.e following relationship:

TMRM (X,Y) $ TARRL (X,Y)

Vhere: Ft.HRM (X,Y) = the reaximum measured radial peak ratio as defined in the CORE OPEPATING 1,IMITS REPORT (CO M).

TaKR'(X,Y) - the reaximum allowable radial peak ratio as defined in the COM, 9

e l

l~

l I

1 I

i for Spocificat6cn 3.2.3 l

Attachment 2:

ACIICH: 1 Vith Tas(X,Y) exceeding its limit: j

-t

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce the allowable THERMAL POWER from RATED THERMAL ;

POWER at least RRHt(D for each 14 that FARR"(X,Y) exceeds the limit, and j

b. Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either: I

1. Restore TAHR"(X,Y) to within the limit of Specification 3.2.3 for .

RATED THERMAL POWER, or

2. Reduce the Power Range Neutron Flux High Trip Setpoint in Table 1 2.21 at least RRH6 for each in that TAHR (X,-Y)

M exceeds that i limit, and ;

c. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the limit of Specification 3.2.3, either:

1. Restore TARR (X,Y) to within the limit of Specification 3.2.3 for M

RATED THERMAL POWER, or

2. Perform the following a~tions:

(a) Reduce the OTAT Kg tenn in Table 2.21 by at least TRH(8) for- 'l each in that TAHR (X,Y) exceeds the limit, and '

(b) Verify through incore mappin5 that FAHR"(X,Y) is restored to vithin the limit for'the reduced THERMAL POWER allowed by (

ACTION a, or reduce THERMAL POWER to less than 5% of RATED THERKAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , l.

{

i

") RRH is the amount of THERMAL POWER reduction required-to compensate for each 14~ that TARR M (X,Y) exceeds TARR'(X,Y), provided in the COLR per l

Specification 6.9.1.9.- f

.$2)TRH is the amount of OTAT 'Kr setpoint reduction required to compensate for each in that FAHR M (X,Y) exceeds the limit. of Specification 3.2.3, provided in the'COLR per; Specification 6.9.1.9.- l 1

1.

I l

l=

!. 'I

f N ni+ \

i POWEP Sl$TRIBUT!0N LIMIT $

3/4.2.3 :A0700 000Li"T n'OT " " LOU OATE "0 NUCLEAR ENTHALPY RISE HOT j

, CHANNEL FACTOR - 74=(e,r)

L1HITING CON 0! TION FOR OPERATION

[ ,-4

,,m i x- , s s I(2iAcceparated in ;

ACTION (Continued) d

O" ##"

c) Reduce THERKAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux High Trip Setpoint to less than or equal to 55% of RATED THERHAL POWER within i the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . __

l

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being within the region of prohibited - .

operation specified in the COLR, verify through incere flux mapping and Reactor Coolant System total flow rate comparison that the com- '

bination of R and Reactor Coolant System total flow rate are restored to within the regions of restricted or permissible operation, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

14 ;

l> ear ts i I

I

. CATAWBA - UNITS 1 & 2 3/442-9a *::ndment Nv. 74(Un;i 1) 4, A;;n;;en; Ne. 50 (Unit 2)- '

(A ni 4 l - - -

1 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued}

i . M /o e.a ]

J/. Identify and correct the cause of the out-of limit condition prior to increasing THERMAL POWER above-the reduced THERMAL POWER limit required by ACTION (b.1.c) and/or o.23, above; subsequent POWER OPERA-TION may Droceed Drovided that :n: ::: tin;ti:n ;f " :nd 'ndi::t:d gjug,M(gy 0;;;ter t;;'.;nt ty;;;; ::t;' ;; r;;; :reldemonstrated, through incore flux mapping :nd ":::t:r C:M : t 6:t:: t;t:1 '!:; c:t: g,. e

n;: ':en, to be within thelr ;i: : :' r::tri:::: ;r ;;r-i::it!: ,

, :::rette specified in the COLk prior to exceeding the following '

THERMAL POWER levels:

/ /) A n;;in;1 50% of RATED THERMAL POWER, z y) A n::'n:1 75% of RATED THERMAL POWER, and J p) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of-attaining greater than or equal to 95%

"Im*4 of RATED THERMAL POWER, au.w .a I

[ SURVEILLANCE REQUIREMENTS -

4 The provisions of Specification 4.0.4 are not applicable.

% .2.3.1 ;

The combination of indicated Reactor Coolant System total flow rate determinedbyprocesscomputerreadingsurdigitalvoltmetermeasurementand]R

{4.2.3.2 shall be deterninsd to be within the regions of restricted or permissible e operation specified in the COLR:

'h, AM a.- Prior to operation above 75% of RATED THERMAL POWER after each fuel M loading, and A e d ame 3 -

b. At least_once per 31 Effective Full Power Days.

(4.2.3.3 The indicated Reactor Coolant System total flow rate shall be verified to be within the regions of restricted or permissible operation specified in

~

L,.

the COLR at least once per 12 hedrs when the most recently obtained value of R, '

obtained per Specification 4.2.3.2, is assumed to exist. ,

(

4.2.3.4 The Reactor Coolant System total flow rate indicators shall be subjected

-to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calc.imetric flow measurement.

4.2.3.S' The . Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

A /Dt ry,saca in Q 6:nce.lt.+ e.

dmert (.:rpuifissuenV.a.5)

A ti.M.h 'ma*-

5 .

3/442-10 ' :nd;;nt Ne. "(Unu U

-CATAW8A: UNITS 1 & 2 Ar.;neent Ne. G (Unit 2) ~

for Specificatien 4.2.3.3 Attact.sent 3:

b. Measuring pgu.R"(X,Y) according to the follwing schedule:

1. Prior to operation above 7h of RATED THERMAL F0VER at the beginning of each fuel cycle, and the earlier of:

2. At least once per 31 Effective Tull Power Days, or

3. At each time the QUADPANT 10VER TILT RATIO indicated by the excore detectors is normalized using incore detector measurements.

9 i

.I

for Specification-4.2.3.2 Attachment 4:=

4.2.3.2 FAHR"(X,Y) shall be evaluated to determine whether Fa(X,Y) is within its limit by:

a. Using the movable incora detectors to obtain a power distribution map at any THERMAL POWE3 greater than 5% of RATED THERMAL POWER.

i y

for-Specification 4.2.3.21 i Attachment 5: ~

c.- Performing the following calculations:

1

1. For each location, calculate the 4 margin' to the 1

maximum allowable design as follows:

M tFu Margin - (1- FARR (X,Y)) x 100%

)

FAHR'(X , Y)

No additional uncertainties are required for !

F4HR M (X,Y), because FAHR'(X,Y) includes uncertainties,

_f i

1

2. Find the minimum margin of all locations examined in-4.2.3 2.c.1 above. If any mar 6i n is less than zero,-

i comply with the ACTION requirements of Specifica'; ion 3.2.3. y

d. Extrapolating the tvo most recent measurements.co 31

' Effective-Full Power Days beyond the'most recent measurement and i : j F4Jul" (extrapolated)-2 FARRL (extrapolated) either of the following. actions shall 1>e.taken: l

1. FAHR M (X,Y) shall be increased by 2 percent over that specified in 4.2.3~.2,a,_and the calculations of 4.2.3.2.c repeated, or 2.- A movable incore detectot power distribution map shall.

be obtained, and the calculations of 4.2,3.2.c shall -l be performed no later than the-tice at which the j margin ~in 4.2.3.2.c is extrapolaced to be. equal ter ,

zero.

i 1

l' L

(AniF(

g-POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2:4 TheOUADRANTPOWERTILTRATIOshallnotexceed1.02[e50%ofRATED (THERMAL POWER./ o n;4 ii -

l APPLICABILITY: MODE 1f*,**

A ACTION:

a, 'With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUA.DRANT POWER TILT RATIO is reduced to within its limit, or b) . THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. .

2. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a) Reduce the QUADRANT POWER TILT RATIO to within its !

limit, or b) Reduce THERMAL POWER;at least 3% from RATED THERMAL POWER for each of indicated QUADRANT POWER TILT RATIO in

f. ca excess o nd similarly reduce the-Power Range Neutron Flux-High ip-Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . -

3, Verify that the-QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding tne limit or reduce THERMAL POWER to-less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip '

Setpoints;to.less than-or equal-to 55% of RATED THERMAL POWER

.within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and '

4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION aoove-50% of RATED THERMAL POWER may proceed provided that the QUADRANT . POWER TILT RATIO is verificd within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%.

or greater RATED THERMAL POWER.

See Special Test Exceptions Specification 3.10.2.

l -; - (1 n se c A et.w, e, e t )

-CATAWBA - UNITS 1 4 2 3/442-1.2

for Specification 3.2.4 Attachment 1:

"Not applicable until calibration of the excore detectors is completed subsequent to refueling.

4

-a 9

l l '.

l'

@(\W \ q l

l l

POWER OISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until. either:

-i a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWERi 2.- Reduce THERMAL' POWER at least 3% from RATED . THERMAL POWER for each 1% of indicated QUADRANT POWER TILT. RATIO in excess of

-within 30 minutes; ,,

3. Verify that the QUADRANT. POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit cr reduce THERMAL

. POWER to less than 50% of RATED THERMAL POWER within the next t 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip

, Setpoints to less than or equal to 55% of RATED THERMAL POWER within the ne:;t 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4. Identify and correct the cau e.of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION l above 50% of RATED THERMAL POWER may. proceed provided that the QUADRANT POWER TILT RATIO .is verified within its-limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% :

or greater RATED THERMAL POWER. ;

c. With the QUADRANT POWER TILT RATIO determined-to exceed 1.09 due to causes other-than the misalignment of-either a. shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until : either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

4 CATAWBA - UNITS 1 & 2 3/442-13

a hl M4 i 1

POWER DISTRIBUTION LIMITS 1

1 LIMITINO CONDITION FOR OPERATION ACTION;(Continued) 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux High Trio Setpoints to less than or equal to 55% of RATED THERMAL POWER within the-next 4 haurs; and-3.- Identify and-correct the cause of the out of-limit condition prior-to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may procesa provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.

y(

I

d. The provisi no.- o f pS cNicatt n J.o M w not, a,pplic ble .

SURVEILLANCE REQUIREMENTS

4. 2. 4.1 The-QUADRANT POWER TILT RATIO shall be determined to be yithin the -

limit above' 50% of RATE ( ?HERMAL POWER by:

a. Calculatilng the ratio at least once per 7 days when-the alarm is OPERABLE, and

b. Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months durir.g steady-state operation when the alarm is inoperable,

c. The provisicss of Specification 4.0.4 are not applicable.

4.2;4.2 The QUADRANT POWER TILT RATIO-snall be determined to be within the; limit ~when'above 75% of RATED THERMAL POWER with one Power Range channel-

. inoperable by using the movable incore detectors ,to confirm that the normalized

-symmetric power distribution, obtained from two sets of four symmetric thimble

-locations or full-core flux map, is consistent.with the indicated _QUADIANT POWER l TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

L CATAWBA - UNITS 1 & 2 3/442-14 'mrewrit No. as (Ur i t 1)

,u rce r.:. N o. u ( Ur. i t 2 )-

t N nIk l POWER DISTRIBtJTION LIMITS

~3/4.2.5 DN8 PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNS related parameters shall be maintained within the limits shown on Table 3.2-1:

a. Reactor Coolant System T,yg, 4

b. . Pressurizer Pressur -

APPLICABILITY: MOD . Reac tor C oole t S ys te r' 7*4*l E /* * < d e..

deniJ.sd in S 2 5 %. ord b. o.bo n O., With e of the *beve parameters ^ exceeding its limit, restore the parameter to within its: limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

hdd'InaeY

@,a+%:he "T

g SURVEILLANCE REQUIREMENTS _

al 4.2.S^ Each of the parameters of Table 3.2-1 shall be verified to be within their limits-at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c\ $nGefY @,

f F

3/442-15 CATAWBA-UNITS 1&2l

i h n[h {

Larf C 5 PM NI8E ** 'f'8 U'!

MtM L, DWE (

pl With the combination of Reactor Co lant System total flow rate and k within

. b. the region of restricted operation within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reduce the Power Range Neutron Flux-High Trip Setpoint to below the nominal setpoint by the same amount

-the COLPa(% RTP) as the oower reduction required by g.e figure epw iii=U in 3,c: S.1-I. rst,eet6 W K

y. With the combination of Reactor Coolant System total flow rate and 4 within C, the region of prohibited operation specified i-g the COLR:

on f"/pra 3.2. ~ / .'

1. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a) Restore the combination of Reactor Coolant System total flow rate and $;,,to,

,, wighgthy region of permissible operation, or b Restore the combination of Reactor Coolant System total flow

~* N )#d 9 ate anc4 to within the region of restricted operation and comply with action a. above, or c) Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER '

and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL' POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , u F, pre 3.2-/

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of i y being within the region of prohibited operation specified Nn th:- COLR, verify -through incccc flux :;pping--

" and D^ccter, Cecient-6ystc; tot:1 fle.; r:t: comperisee.that the com-p ination of7-and Reactor Coolant System ,otal flow rate are restored to within the regions of restricted or permissible operation, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

IO5e b l 4.2.5 h The Reactor Coolant System total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at-least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prier to the performance of the. calorimetric flow measurement.

l ' 4.2.M The Reactor Coolant System total flow rate shall be . determined by precision heat balance measurement at least once per 18 months.

1 .

l l

(d n ik \

1 TABLE 3.2 1 i

DNS PARAMETERS 1

PARAMETER LIMITS Four Loops in Operation Average Temperature Meter Average - 4 channels: < 592*F i.

- 3 channels: {592'F ) 3 Computer Average' - 4 channels: < $93'F I

- 3 channels: (

{593*F i

-Pressurizer Pressure Meter Average - 4 channels:

- 3 channels:

> 2227 psig" E2230psiga

/

Computer Average - 4 channels: > 2222 psig" ' ./ l

- 3 channels: [2224psig"

[

l L-

.)

, Pa.eter Coolad Sis % TcM Flow We. F:yce 3.2-1 /

1 8

l[

i

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED. THERMAL POWER.

3/4A2-16

^ - - ' - " " - ^"*"

CATAWBA - UNITS 1 & 2 {lG ] "' ] " [l ? ] 5 j { R

i Tigure 3.2 1. Reactor Coolant System Total Flow Rate Versus Rated Thera l Power Tour t. oops in operation ((fpl[

-1 388850 A penalty of 0.M for uno.tected feeanter venturt foubo and a Permissible - ;

rn.. w .rn.ni une.ri.iniv of 2. a for operation

-t u .r. inewo.ein ini now.. Region -

.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . .... . ! .* 8: 3 8 6 0 0 0).-

385000- >

i Restricted (ee,3stiso)

E 381150 Operation i

S Region -

.e

~

ca .

. CC : i R '~

9 377300- #'377300Ii .s u.

E ,

. _e

.m. ,

>- f

-CO '

I'3'37

0I

, j373450-T

-o-g.

w Prohibited '

S o Operation ;

+~ 0,msoo) .Re @ -

cll 369600-365750

'061900-86: 88- 90 92 94 96 '98 100 102 Fraction of Rated Therrnal Power CATAWBA WIul+ Z 3/q/p-/}

This page intentionally left blank

l.) f\ $ Y l- l l

-3/4.2 POWER DISTRIBUTION LIMITS

-3/4.2.1_ AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

a. the allowed operational space as specified-in the CORE OPERATING

(

LIMITS REPORT (COLR) for RAOC operation, or ,_

b. within the target band specified in the COLR about the target flux difference during baseload operation.

APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWERA'* [un;-t-i ACTION:

a. For RAOC operation with the Indicated AFD outside of the limits - i specified in the COLR, $- -

1. Either restore the indicated AFD to within the COLR limits within > .

15 minutes, or f

2. - Reduce THERMAL POWER to less than 50% of RATED' THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .-

b. For Base Load operation above APL with the' indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target flux difference:

1. Either restore the indicated AFD to within the COLR specified f target band limits within 15 minutes, or ND

2. Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes,

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR. j t

L *See Special Test Exceptions Specification 3.10.2.

HD

- APL is the minimum allowable (nuclear design) power level for base load operation and is specified in the CORE-0PERATING LIMITS REPORT per s/

Specification 6.9.1.9.

^ = dent Ne " (Unit ik CATAWBA - UNITS 1&2 3/482-1 "

.,nadant Nc. 69 (Unit 2)-

'bnif d F

POWER-0ISTRIBUTION LIMITS-LIMI11NG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 4.2.1.1-- The indicated AFD shall be determined to be within its limits during '

POWER OPERATION above 50% of RATED THERMAL POWER by:

a. Monitoring the indicated AFO for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, j and

2) At least-once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status, b.. Monitoring and logging-the indicated AF0 for.each OPERA 8LE excore channel -at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereaf ter, when the AFD Monitor Alarm is inoper-able. The logged values of the indicated AFD shall be. assumed to exist during the interval preceding each logging.-

, c. The provisions of Specification 4.0.4 are not applicable.-

4.2.1.2 The indicated AFO shall be considered outside of its limits when at i least two OPERABLE excore channels are indicating the AFD to be outside the limits. I !

- 4.2.1.3' When in Base Load operation, the target axial flux difference of C

- each OPERABLE excore channel shall'be determined by measurement at least once per 92 Effective Full Power Days.= The provisions of Specification 4.0.4 are not applicable, ,. ,

- 4.2.1.4 : When-in -Base Load operation, the target flux dif ferenc'e shall be -

up a ed t d at least once per 31 Effective-Full Power. Days by either determining h/ !

the target flux difference in conjunction with the surveillance requirements of ,

- Specification 3/4.2.2 or by linear interpolation between'the most recently mea- l sured values and the calculated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable, 1

l l

CATAWBA - UNITS 1&2 3/462-2 5:ntent 40.29 (Unit 1) fr.:n?T,ent h. n (Unit 2)

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CATAWBA - UNITS 1&2 3/462-3 Amendment No. ;'4 (Unit 1)

Amendment-Nc. 68=(4Jnft-2-)-

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CATAWBA - UNITS 1&2 3/462-4 ^ endent Mc. 39 (Ur.it 1) -

^? ndent "O,31 (Unit 2)

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(2 n . i A ]

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - qF (Zj !

LIMITING CONDITION FOR OPERATION-3.2.2 F g(Z) shall be limited by the following relationships:

Fq (Z) i F f K(Z) for P > 0.5 ' '

P ( i Fq (Z) $

K(Z) for P 1 0.5 Where: _FRTP q = theq F Limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT '

(COLR), 3 p THERMAL POWER , and

~ RATED THERMAL POWER K(Z) = the normalized F (Z) for a given core height-9 specified in the COLR, I APPLICABILITY: MODE 1.

hn M.

' ACTION:'

With F (Z) exceeding its limit:

9

a. Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit 9

within 15 minutes and similarly reduce the Power Range Neutron Flux-High-Trip-Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up.to-a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower:aT Trip Setpoints (value of.K4 ) have-been reduced at least 1% (in AT span) for each 1% Fq(Z) exceeds the limit, and

b. Identify- and correct the cause of the out-of-limit condition prior~

to' increasing THERMAL POWER above the reduced limit required by ACTION a;, above; THERMAL _ POWER may then be increased provided Fq (Z) is demonstrated through incore mapping to be within its limit.

l l

l CATAWBA - UNITS 1&2 3/462-5 -Ani ucs,ent "v. 74 (Unit ib Amenament no. 65 (Un't -2)

!) Yk li f

POWER-0ISTRIBUTION LIMITS 4 SURVEILLANCE REQUIREMENTS t 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 For RAOC operation,qF (z) shall be- evaluated to determine if F (z) q

'is within its limit by: ;

a. Using the movcbie incore-detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

s

b. Increasing the measured F q (z) component of the power distribution map by 3% to account for manufacturing tolerances and further in-creasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied, j

c. Satisfying the following relationship:

RTP -

M F Fq (2) Q x K(z) for P > 0.5 P x W(z)

RTP M F Fq (z) 1 Q x K(z) for P-< 0.5 W(z) x 0.5 - i where'F (2) is the measured Fq (z) increased by the allowances for \

TP manufacturing tolerances and measurement uncertainty, F is the

-F q

limit, K(z).is the normalized Fq(z) as a function of core height, P is-the. relative THERMAL POWER, and-W(z) is the cycle-dependent ,

g ' function that accounts for power distribution transients encountered 1 during normal operation. F TP , K(z), and W(I)=are specified in the CORE OPERATINGELIMITS REPORT per Specification 6.9.1.9.

d. Measuring F (z) according to the following schedule:

1. Upon achieving equilibrium conditions after exceeding by 10% or more.of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or 9

2. -At least once per 31 Effective Full Power Days, whichever occurs l

-first.

l-l *0uring power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

CATAWBA - UNITS 1&2 3/462-6 ^ end=nt-Mc. 74 (Unit-1)

,^ end=rtt Mc. SS (Unit 2) '

hne t A POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

e. With measurements indicating

. maximum F '(z) over z K(z) has increased since the previous determination of F (z) either of the-following actions shall be taken: !

t 1)' FgM (z) shall be increased by 2% over that specified in Specification 4.2.2.2c., or

2) F (z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that ;

maximum FId n

(z) is not increasing, over z K(z)

f. With the relationships specified in Specification 4.2.2.2c. above not being' satisfied:

1) Calculate.the percent F g (z) exceeds its limit by the following '

expression:

/ -

M y

maximum Fg (2) x g(*)* ,x 100 for.P ~> 0.5 ;

over z p RTP L -

f x K(z)

. ,) (

1 ( -

, 8 maximum F (z) x W(z). -IIx 100 for P < 0.5 over z " pRTP / ,

. .5 x K(z)

! 2) One of the following actions shall be taken: '

a) Within 15 minutes, control the AFD to within new AFD limits !

which-are determined by reducing the AFDLlimits of ,

Specification 3.2.1=by 1% AFD for each percent Fg(z) exceeds /l its . limits as determined in Specification 4.2.2.2f.1).

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to these' mod-ified limits,- or

, b) Comply with the requirements of Specification 3.2.2 for Fg (z) exceeding its limit by the percent calculated above, or c) Verify that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load operation.

. CATAWBA - UNITS 1&2 3/482-7 ^:entent S. 74 (5it 1)-

^=cntcat "e. 50 (Unli. 2)

Gifh POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

g. The limits specified in Specifications 4.2.2.2c., 4.2.2.2e., and 4.2.2.2f., above are_not applicable in the following core plane regions:

1. Lower core region from 0 to 15%, inclusive

2. Upper core region from 85 to 100%, inclusive.

4.2.2.3 Base Load operation is permitted at powers above APL NP if the following ( ,

conditions are satisfied:

a. Prior to entering Base Load operation, maintain THERMAL POWER above ND APL and less than or equal to that allowed by Specification-4.2.2.2 e for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain Base Load operation surveillance (AFD within the target band about the-target flux differ-ence of Specification 3.2.1) during this time period. Base Load (/

operation is then permitted providing THERMAL POWER is maintained ND OL ND between APL and APL or between APL and 100% (whichever is most limiting) and FQ surveillance is maintained pursuant to Specification 4.2.2.4. APL BL is defined as:

p RTP APLBL , minimum Q x K(Z) c[

over Z ] x 100%

g M(Z) p q x W(2)BL where: F (z) is the measured Fq (z) increased by the allowances for '

TP manufacturing tolerances and measurement uncertainty. F is-the-Fq limit, K(z) is the normalized F q(Z) as a function of core height.

W(z)BL is the cycle dependent function that accounts for limited power distribution transients encountered during Base Load-operation.

l Ff,K(z),andW(Z)BL are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9,

[

b. During Base Load operation, if the THERMAL POWER is' decreased below I HD

APL then the conditions of 4.2.2.3a shall be satisfied before 1 l re-entering Base load operation.

l 4.2.2.4 During Base Load Operationq F (Z) shall h evaluated to-determine if L F (Z) is within its limit by:

q

a. Using the movable incore detectors to obtain a power distribution HD l map at any THERMAL POWER above APL ,

b. Increasing the measured Fq (Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the L requirements of Specification 3.2.2 are satisfied.

ND

APL is the minimum allowable (nuclear design) power level for Base Load operation in Specification 3.2.1. f)

CATAWBA - UNITS 1 & 2 3/4B2-7a y,hty.yGnity

. . ., _mw o m. w oe u

$n i W l

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

-c. Satisfying the following relationship:

F (Z) $

p RTP p

0 l for P > APL HD

(

where: F (Z) is the measured Fg (Z). -is the Fq limit.

FhTP i K(Z)-is the normalized F q (Z) as a function of core height.. P is the ;

relative THERMAL POWER. W(Z) is the cycle dependent function that

-accounts for limited power dibribution transients encountered during Base Load operation. .FhTP , K(Z), and W(Z)gg are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. 1

d. Measuring F '(Z) in conjunction with target flux -difference deter-mination a: ording to the following schedule:

1. Prior-to entering Base Load-operation after satisfying surveil-lance 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been-maintained above APL HO for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior _to mapping, and ,

At least once per 31 effective full power days, 2.

e. With measurements. indicating maximum ~F (z) l L over z I K(z) 3 has increased since the previous-determination ~F (Z) either of the following actions shall be taken:-

' 1.4 F (Z)-shall be increased by 2 percent over that specified in l 4.2.2.4c, or

'2. F (Z) shall be measured at least once per 7 EFPD_until 2 successive maps . indicate that maximum FM (*) - is not increasing, over z E K(z)_J

f. With the relationship specified in 4.2.2.4c above not being satisfied, either of the following actions shall be taken:

1. Place the core in an equilibrium condition where the limit in 4.2.2.2c is satisfied, and remeasure F (Z), or CATAWBA - UNITS 1 & 2 ^

3/482-7b

^

. dent--h. 74 (Unit 1)

. x ad. Tent "e. 00(Unii ^)

(A rd P 3 1 POWER-DISTRIBUTION LIMITS SURVEILLANCE ~ REQUIREMENTS (Continued)

'2. Comply with'the-requirements of Specification 3.2.2 for F (Z) exceeding its limit by the percent calculated with 9

the' following expression:

-[(max. over z of ( F (Z) x W(Z)BL ) ) -1 ] x 100 APL for P->

NO FhTP x K(Z)

~

P

g. - The limits specified in 4.2.2.4c., 4.2.2.4e., and 4.2.2.4f.-

above are not applicable in_the following core plan regions:

-1. Lower core region 0 to 15 percent, inclusive. .

2 .- Upper core region 85 to 100-percent, inclusive.

4.2.2.5 'When Fq (Z) is measured for reasons other than meeting the requirements

of Specification 4.2.2.2 an overall measured Fq(z) shall be obtained from a power ,

distribution map and increased by 3% to account-for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. ,

.l L

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CATAWBA - UNITS 1 & 2 3/482-7c ?^ en eent 40. 7d(Unit 1)

. =ninnt ."c. 68 (Uni t 2)- .

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CATAWBA - UNITS 1 & 2 3/482-7d ^ -^ d:O n t I

-Amendmen,je39$U"il_ ,

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                                                    ^ endment ha4 (Unit 1)-

CATAWBA - UNITS 1 & 2 3/432-8

                                                    ^ end:ent tic. 50(Unit 2)

I'l i f 1 l POWER DISTRIBUTION LIMITS 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FUR OPERATION 3.2.3 The combination of indicated Reactor Coolant System total flow rate and R shall be maintained within the region of permissible operation specified in g < the CORE OPERATING LIMITS REPORT (COLR) for four loop operation. (, Where: ' N F AH , i

a. R=

FhPgy,0 + mfg (1.0 - P)) THERMAL POWER , ( ' b' P

                      = RATED-THERMAL POWER

c. Fh = Measured values of Fh obtained by us'ing the movable incore f detectors to obtain a power distribution map. The measured values of Fh shall be used to calculate R since the figure specified in the COLR includes penalties for undetected feed- ( :

water venturi fouling of 0.1% and for measurement un:ertainties of2.1%forflowand4%forincoremeasurementofFfg,

d. FRTP= The Fh . limit at RATED THERMAL POWER (RTP) specified in the COLR, and

(

e. 'MFAH= The power factor multiplier specified in -the COLR.

APPLICABILITY: MODE 1(lini-\-A)

    ' ACTION:

a. With the combination of Reactor Coolant System total _ flow rate and R within the region of restricted operation within 6 hours reduce the Power Range Neutron Flux-High Trip Setpoint to below the nominal setpoint by the same amount (% RTP) as the power reduction required by the figure specified in the COLR.. >

(

~~

b. With the combination of Reactor Coolant System total flow rate and R within the region of prohibited operation specified in the COLR:

1. Within 2 hours either:

a) Restore the combination of Reactor Coolant System total flow rate and R to within the region of permissible operation, or b) Restore the enmbination of Reactor Coolant System total flow rate.and R to within the region of restricted operation and comply with action a. above, or CATAWBA - UNITS 1 / 3/4B2-9 ^ ten d ent S. 7d(Unit 1) 5 :n t;nt h. 50(Unit 2)

() n i .-f- M - POWER DISTRIBUTION LIMITS: 3/4.2.3 REACTOR-COOLANT- SYSTEM FLOW RATE _ AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION ACTION (Continued) c) -Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint ,

                   -to less than or equal to 55% of RATED THERMAL POWER within' the next 4 hours,                                                         '

2.- Within 24 hours of initially being within the region of prohibited operation specified in the COLR, verify through incore. flux mapping and Reactor Coolant System total flow rate comparison that the com- 4f bination.of-R and Reactor Coolant System total-flow rate are restored to within the regions of restricted or permissible operation, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours. m i CATAWBA - UNITS 1 & 2 3/462-9a A::nd::nt No. 7d (Unit 1)

                                                             - A::ndment N . 69 (Unit 2)

N ni f S POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)

3. Identify and correct _the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced-THERMAL POWER limit required by ACTION b.1.c) and/or b.2., above; subsequent POWER OPERA-TION may proceed provided that the combination of R and indicated Reactor Coolant System total flow rate are demonstrated, through ,

incore flux mapping and Reactor Coolant System total flow rate comparison, to be within the regions of restricted or permissible _ operation specified in the COLR prior to exceeding the following THERMAL POWER levels: f' a) A nominal 50% of RATED THERMAL POWER, b) A nominal 75% of RATED THERMAL POWER, and c) Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.3.1' The provisions of Specification 4.0.4 are not applicable.

                    '4.2.3.2, The combination of indicated Reactor Coolant System total flow rate determined by process computer readings or digital voltmeter measurement and R shall be determined'to be within the regions of' restricted or permissible
                     -operation. specified .in the COLR:

a. Prior to operation above 75% of RATED THERMAL-POWER after each fuel f

loading, and i b.. At least once_per 31 Effective Full. Power Days.

                    -4.2.3.3 The indicated Reactor Coolant System total flow rate shall be verified to be within the regions-of restricted or permissible operation specified in thelCOLR at least once'per 12 hours when the most recently obtained value of R, L

l obtained per Specification 4.P. 3.2,-is assumed to exist. ([- l '4.2.3.4 The Reactor Coolant System total flow rate indicators shall be subjected to-a CHANNEL CALIBRATION at least once per 18 months. The measurement-

                     -instrumentation shall be calibrated within 7 days prior to the performance of

!. the calorimetric flow measurement. 4.2.3.5 The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once.per 18 months. l l -CATAWBA - UNITS 1 & 2 3/462-10 ^=andment Ne. 74(Urit 1)' l ^ endment 40. 00(Ur.it 2)

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l l CATAWBA - UNITS 1 & 2 3/402-11 ^ .~-. ~u me - th- ,m u , u. m so,

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Unif h e . POWER DISTRIBUTION LIMITS 3/4.2.A QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02 e ^ve SC M RATED-THERMAL POWtt-* APPLICABILITY: MODE 1 %ove. $0% e4 RATED TEWL ANU b* 3 ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. ,

2. Within 2 hours either:

a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux High Trip Setpoints within the next 4 hours,

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 nours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and

4. Identify rnd correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least

(,nce per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.

     *See Special Test Exceptions Specificatior. 2. 4 2.

CATAWBA - UNITS 1 .t 2 3/482-12 v_ - _ -_-__ _ _ ___ -

h ()i h h

                                                          @ (R 0!$TRIBUT!0N LIMITS LIMITING CON 0! TION FOR CPERATION ACTION (Continued)

b. With the QUADRANT POWER TILT RATIO determined to exceed 1 09 due to misalignment of either a shutdown or control rod

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to.less than 60% of RATED THERHAL POWER.

2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1, within 30 minutes; ~ 8' 3

3. Verify that the QUADRANT POWER TILT RATIO is within its limit

           -                                                               within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERkAL PCWER within the next 2 hours 'and reduce the Power Range Neutron Flux-High Trip 5etpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and

4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is. verified within its limit at least once per hour for 12 hours or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c. With the QUADRANT POWER TILT RATIO determined to exr;<.d 1.09 due to causes other than the misalignment o.f either a shui /Jwn 0F control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its_ limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. 9 CATAWBA - UNITS.1 & 2 3/492-13

                                                                                                           =.           ..        .       .    - .- -

                    ,'                                                (/ n i k M POWER 0!$TRIBUT!0N LIMITS, LIMITING CON 0! TION FOR OPERATION ACTION (Continued)

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and

3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the [

Qt)ADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER. ( % pcc4SienS 4 $pedOccch00 3,M are ndcippliCAb d j SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a. Calculating the ratio at least once per 7 days when the alarm is-OPERABLE,'and

b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable,

c. The provisions of Specification 4.0.4 are not applicable. !

i 4.2.4.2 The QUADRANT POWER TILT RAT.IO shall be determined to be within the limit when-above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized

 -symmetric power distribution, obtained from two sets of four symmetric thimble locations or full-core flux map, is consistent with the indicateo QUADRANT POWER         .

TILT RATIO at least once per 12 hours. ' i i CATAWBA UNITS 1 & 2 3/462-14 ^== M ent Nc. 40(Unii 11

                                                                % ne:nt N;. i.i(Unit. 2)
                                + . -     . :.

(.A n i E 1 i POWER DISTRIBUTION LIMITS ! 3/4.2.5 ON8 PARAMETERS

. 4

                     -LIMITING CON 0! TION FOR OPERATION j

t 3.2.? The following DNB related parameters shall be maintained within the 1 limits shown on Table 3.2 1: p.

a. Reactor Coolant System T,yg, and i b. Pressurizer Pressure, j

APPLICABILITY: MODE 1) h ,4 2 cni . [. [ ACTION: ! With any of the above parameters exceeding its limit, restore the parameter to i- within its limit within 2 hours or reduce THERMAL POWER to less than 5% of I: RATED THERMAL POWER within the next 4 hours. e i: e

         \            SURVEILLANCE REQUIREMENTS l.

4.2.5 Each of the parameters of Table 3.2 1 shall be verified to be within

their limits at least--once per 12 hours.

( j l:- c U . , 1 CATAWBA UNITS 1 & 2 3/432-15

blM & 1 1 TABLE 3.2-1 ] DNB PARAMETERS ; i i PARAMETER LIMITS Four Loops ; in Operation i- Average Temperature ,

            -Meter Average         - 4 channels:      < 592'F                                   .
                                   - 3 channels:      [$92'F                                  !

Computer Average - 4 channels: < $93'F

                                   - 3 channels:      7 593'F                                1 Pressurizer Pressure Meter Average         - 4 channels:      > 2227 psig"                        -

{

                                   - 3 channels:      E2230psg*                               f

( Computer t,verage - 4 channels: > 2222 psiga

                                   - 3 channels:      > 2224 psig*                        )   !

g : I I i i i

Limit not applicable during either a THERMAL POWER ramp in excess of'5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

CATAWBA - UNITS 1 & 2- 3/492-16 ? end ent Me. ::,(Unit 1) ;

                                                          . ^end-^nt Me, it,(L' cit 2)        !

1

                                                                                                                                                       - - - - - - - " ' " - -^"   " - - ^ - - - -
         " - JPG,m p me-,6m 44_&a4%.Jde A                    A..--.AA amm.A aeN J. A .Ahu ,4+ A m.Ah - 5eJAAsa-                                    -                             -

As. e 4 4e.e . eaa-4.. h e b SAFETY' LIMITS AND POWER DISTRIBUTION TS BASES MARK-UPS t

l l 2 i ii l }- j.- i. r.- I l t.._....-_4....~ . , . . . . . - . . . . - . . - ~ , . . . _ _ - . . .... . . . . - . . . _ . . . . _ , . . . - . . . . . - _ . -

_ . . _ ~ _. _ _ ._ . _ . _ _ _ _ _ _ . _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . j I J i 2.1 SAFETY LIMITS 4 BASES ln 5(TY j@W 2 .1 1 REACTOR CORE (.Go e b+ h

 - unal The restrictions of this Safety Limit prevent overheating of the fuel and                      l possible cladding perforation which would result in the release of fission l

products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer i coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolent Temperature and Pressure have been related to DNB through the WRB-1 correlation. The WRB-1 DNB correlation has i been developed to predict the DNB flux and the location of DN8 for axially uniform and nonuniform heat flux distributions. The local ON8 heat flux ratio, (DNBR), is defined-as the ratio of the heat flux that would cause ON8 at a particular core location to the local heat flux, and is indicative of the margin to ONB. The DN8 design basis is as follows: there must be at least a 95% probacility that the minimum DNBR of the limiting rod during Condition I and

             !! events _is greater than or ecual to the DNBR 1imit of the DN8 correiation being used (the WRB-1 correlation in this application). The correlation DNBR i

limit is established based on the entire applicable experimental data set such that there is a 95% probability with 95% confidence that DNB will not occur when the minimum ONBR is at the DNBR limit. In meeting this design basis, uncertainties in plant operating parameters, l nuclear and-thermal parameters, and fuel fabrication parameters are co'*idered l statistically such that there is at least a 95% confidence that the minimum ONBR fo the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR' limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. The curves of Figure 2.1-1 show the loci _of points of THERMAL POWER, Reactor Coolant System pressure and average temperature colow which the calculated DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

CATAWBA - UNITS 1 & 2 8 2-1

L i l l 2.1 SAFETY LIMITS BASIS i g This curve is based on a nuclear enthalpy rise hot channel factor, Fg, of 1.49 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F g at reduced power based en the .! expression: N Fg = 1.49 (1 + 0.3 -(1 P)] Where P is the fraction of RATED THERMAL POWER. These limiting heat flux conditions are higher than those calculated for ' the range of al1~ control rods fully withdrawn to the maximum allowable control red insertion assuming the axial power imbalance is within the limits of the f t(AI) function of the Overtemperature trip. When the axial power imtalance is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits. ] 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor vessel, pressurizer, and the Reactor Coolant System piping, vaiscs, anc fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of cesign pressure. The Safety Limit of 2735 psig is therefore consistent witn the design criteria and associated Code requiren nts. The entire Reactor Coolant System is hydrotested at 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation. CATAWBA - UNITS 1 & 2 8 2-2

   ._ _ _ . _     _ _        _ . _ _ _        _ ~ _ . _ _ _ _ _ _                                    _ _ . _ _ _ _ _

g% ~ BASES ._ t I 2.1.1 REACTOR CORE ( per hi+ h l The restrictions of this $4fety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products te the reactor coolant. Overheating of the fuel cladding is prevented i

'              by restricting fuel operation to within the nucleate boiling regime where the heat transfer Coefficient is large and the claccing surf ace temperature is slightly above the coolant Saturation temperature.

Operation above the upper boundary of the nuclette boiling regime could 4 result in excessive cladding temperatures cecause of the onset of caparture from nucleate boiling (DNB) and the resultant sharp recuction in heat transfer coefficient. ONB is not a directly measurable parameter during cperation and therefore THERMAL POWER and related to DNB through theWD p,Racorrelation. gtor Coolant The Temper and Pressure have been ONB correlation has

                                                                                        , ion of CNB for axially been uniformdevelopec        to preciet and nonuniform    heat flux      tnelcistributic DNB flux anc theyThe local DNB heat, flux ratio (DNBR), is cefined as the ratio of the he                          lux that would cause ONB at a particular core location to /the local Sea fivx, and is indicative of the margin to ONB.

g / The ONB cesign bas is as fo110ws: there must be at least a 95% . ! probability that the rr nimum CNBR of the limiting roc curing Concition I and II events is grel.t r .han or ecual to the ONBR 11mit of the ONB correlation being used (the 4 W r correlation in this application). The correlation DNBR limit is establisnec based on the entire applicable experimental cata set such that there is a 95% probability with 95% conficence that ONE will not_ occur when the minimum DNBR is at the ONBR limit, 4 g g g,x,u g ,,gg InmeetingthisdesignCasis,uncertaintiesinplantoperakingparameters, nuclear and thermal parameters, .+ne fuel fabrication parameters 4are consiceree i statistically such that there is at least a 95% conficence that tne minimum i DNBR for the limiting red is greater than or ecual to the ONBR limit. The t uncertainties in the above-p h parameters are used to determine the plant i DNBR uncertainty. This ONBR uncertaintyr ce W nce with th: c r cict s DN04 /s used o H ete establishee a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. The curves of Figure 2.1 1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the cciculated DNBR is no less tnan the cesign ONBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated licuic.

                 -GATAWBA---tittliS-t-& 2                         3 $1 s 4s

1 fQQwhI W '

      --- mm                                                                                                                            -

BASES

                                                                               " ^ - - ~ '                     ~-

N T curve 3 Nbased on a nuclear enthalpy rise het channel factor, F3g, of 1.49tand a reference cosine with a peak of 1.55 for utal power shape. i An allowance is included Mr an increase in F N at reduced power based on the expression! Q de t Wees e Oc+ a d N Q E5 Cor *Ge h d *C k ** K 2 d v.ci a swkI es sse. c' es (on s) and

--)

Fy = 1.49 (1 + 0.3 (1 P)) 7.3 T 7: +,3v.. 7 E.'c

[ Where P is the fraction of RATED THERMAL PCWER. These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control red insertion assuming the axial power imbalance is within the limits of the

       !      f t(AI) function of the OvertemperatureAtrip. When the axial power imoalance

is not within the tolerance, the axial $ower imbalance ef fect on the Over-temperature AT trips will reduce the $ t oints to provide protection consistent with core Safety Limits. 4 f

l 2.lbREACTOR COOLANT SYSTEM DRES$URE k The r c' ion of this Safety Limit protects the inte f the Reactor Ocolant System fro verpressuri:ation a.nd thereby preve e release of racionuclides containe n the reactor coolant from as ing the ccatainment atmosphere, f The reactor vessel, pressurin a .he Reactor Coo'lant System piping, valves, and. fittings are cesigned + ion III of the ASME Code for Nuclear Power Plants which permits a m mum trans At pressure of 110%'(2735 psig) of cesign pressure. The'$aff Limit of 2735 ps therefore consistent with t tne design criteria w associated Code requiremen The entir tor Coolant System is hydrotested at 125% ( t psig) of cesig ssure, to demonstrate integrity prior to initial ope Kron. D h * , n , g [ 9 i CPWA -WITS 112

-M-

3/4.2 POWER DIST't!BUTION LIMITS (f~o.t M <s i t \ BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core greater than or equal to design limit DNBR during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria

f:it ;f MOO" 6 not exceeded, b

The definitions of certain hot channel and peaking factors as used in these specifications are as follows: Fg (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the egg,g g) average fuel rod heat flux, allowing for manufacturing tolerances on u fuel pellets and rods; Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of

 -__                   the integral of linear power along the rod with the highest integrated power to the average rod power.

KCO i$ defia*d " + " " " m'hy d 6 (*Xd) Nt I" '- f ' h* f ' 3/4.2.1 AXIAL FLUX DIFFERENCE gg 74 (xyp) ThelimitsonAXIALFLUX01FFERENCE(AFD)f:::ur:gthat the O F N " RTP 5: d erv:10p:!:f th: F limit /pecified in the CORE OPERATING LIMITS REPORT] ((COLR)f;i::: th: n:r::li::t =t:j p::Mng f::ter 6 not exceeded during either norma'l operation or in the event of xenon redistribution following power changes.V [The full-length Targe x difference is determined at equilibrium xenon may be positioned within the core in y ordance with tions. their respective inser imits and should be ins ei near their normal position for steady-state ope n at high p - evels. The vclue of the target flux difference obtained unde & conditions divided by t;'e fraction of RATED THERMAL POWER is the tar Tux ence at RATED THERMAL '0WER for the associated core burn onditions. Target differences for other THERMAL POWER levels a tained by multiplying the RA ERMAL POWER salue by the appropria ractional THERMAL POWER level. The perio dating of the target difference value is necessary to reflect core burnup hensi ions e - pelst s o.no ne Fas % C N M W Th AFD enveApv_ g ec;fied in w e Cot.R. h as bem c. A p s h d & mens m ma- unwt.ainty . y CATAWBA - UNITS 1 & 2 B 3/4 2-1 Mgd:E} P. M Gfij} U

                                                                                          . .. , , ~ m n . .~. wv wu i s er

   . _ . .       -.-               .-~       -     -                _ _ - . - - -                    .-         - - -       - _
           'p0WER DISTRIBUTION _ LIMITS BASES ND At power levels below APL , the limits on AFD are defined in the COLR, f'1.e.,thatdefinedbytheRA0Coperatingprocedureandlimits.                                            These limits
                                                                                                                         ^*\    l were calculated in a manner such that expected operational transients, e.g.,

load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of time allowed outside of the limits at reduced power levels will not result in signi-ficant xenon redistribution such that the envelope of peaking factors would change sufficiently to prevent operation in the' vicinity of the APL ND power level. At power levels greater than APLND, two modes of operation are permis-sible; 1) RAOC - the AFD limits of which are defined in the COLR, and 2) Base Load operation, which is defined as the maintenance of the AFD within a COLR specified band about a target value. The RAOC operating procedure above ND is the same as that defined for operation below APLND. ~ APL However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation withqF (z) less than its limiting value. To allow operation at the maximum permissible value, the Base Load operating procedura restricts b

                                  .DeletA i

9 CATAWBA - UNITS 1 & 2 8 3/4 2-2 pg5[g55[.} "g., , [f[j} }} S

                                ,        _              . . , - . .          ...4   _ _        . , ,     ,         - - .

i 44&saw POWER DISTPIBUTION LIMITS l BASES l l l 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SY$ TEM FLOW RATE AND NUCLEAR ENTHALPY RISE m0T CHANNEL FACTOR (0:n;iN::b (bit Q the indicated AFD to relatively small target bsnd and power swings (AFD target AO band as specified in the COLR, APLND < pg,,7 < gpLBL or 100% Rated Thermal Power, I whichever is lower). For Base Load operation 7 it is expected that the Units will operate within the target band. Operation outside of the target band for the I short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above. To assure there is no residual xenon redistribution impact from past operation on the Base Load operation, a 24 hour waiting period at a power level above APL"0 and allowed by RAOC is necessary. During this time period load changes and rod motion are restricted to that allowed by the Base l cad procedure. After the waiting period extended Base Load operation is permissible. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE axcore channels are: 1) outside the allowed A! power operating space (for RA00 operation), or 2) outside the allowed AI-target band (for Base Lead occration). These alarms are active when power is greater than: 1) $0% of RATED THERMAL POWER (for RAOC operation), or

2) APLND (for Base Load operation). Penalty deviation minutes for Base load )

operation are not accumulated based on the short period of time during which j v operation outside of the target band is allowed. f gc The limits on heat flux hot channel factor,- :0:1:nt '1:; r:te, .and nv;1 ear n er., enthalpy rise hot channel factor ensure that: (1) the design limits on r,eak

local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the p :t fe:1 h d t::;:r:tu : $ n:t ;xc;;d tN H E+ W ECCS acceptance criterial W 4 he+e#11mits are specified in the CORE OPERATING LIMITS REPORT (, cod per Specification 6.9.1. . negese<ng 4 4;;h ;f th::: b-measurable but will normally only be determined periodically as specified in Specifi:ations 4.2.2 and 4.2.3. This periodic i

surveillance is sufficient to insure that the limits are maintained provided: t

a. Control rods in a single group move together with no individual red insertion differing by more than t 12 steps, indicated, from the group demand position; 1

b. Control red groups are sequenced with everlapping groups as described in Specification 3.1.3.6; 7%e h a u d, flus het chsrel Steve e,nd nudur en n tm nse. h c charsecJ Actw we each CATAWBA - UNITS 1 & 2 B 3/4 2-2a M, gg} y. y gj} Q

_ ,, -. ~ - .. r

    - _ . --              -    - _ - -          - . .      - - - - . - . - - . -                        - - - - _._                            -~.

i l l p0WER O!STRIBUTION LIMITS BASES i HEAT FLUX HOT CHANNEL FACTOR, d "! ACTC" COOLM !T44SkOW4W-AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR g tinuec) p i et) c. Tu control 3.1.3.6 red insertion are maintained; and limits of Specifications 3.1.3.5wand ~"g@j M u, as s e.s d.- N The axial power cistribution, expressed in terms of Ax1AL FLUX O!FFERENCE, is maintained within the limits. [j.fw}j

                                                                                                                            .& n F        ill be nizintained within its limits provided Conditions a, through d.

above are maintained./ As noted on the figure specified in the CORE OPERAT M I LIMITS REPORT 3 h Reactor Coolant System flow rate and Fh.m6y e " traded off" against one another (i.e., w measured Reac.ter d nt System flow rate N

                ;isacceptableifthemeasuredF                             $, ,y W g g q nsure that the calculated r DWBR will not be belopeJe                                    R value.

function of The rDa%n4( as a RMAf p0WER allows changes in the radial power shape for t4 Q ssible rod insertion limits. F R as calculated in Specification 3.2.3 and used in the figure specified o hneh- I N mcM. in the COLR, accounts for F g less than or equal to the F g RTP limit specified . I in the COLR. This value is used in the various accident analyses where

                'FhinfluencesparametersotherthanDNBR,e.g.,peakcladtemperature,andthus is the maximum "as measured" value allowed. The rod bow penalty as-a function of burnupappliedforFfg is calculated with the methods described in WCAP 8691, Revision 1, " Fuel Rod Bow Evaluation," July 1979, and the maximum rod bow penalty is 2.7% DNBR. Since the safety analysis is performed with plant-specific safety ONBR limits compared to the design DNBR limits, there is sufficient thermal                                             l Qrginavailabletooffsettherodbowpenaltyof2.7%DNBR.                                                           _

[ cycleThe hot channel f actor F (2) is measured periodically and. increased by a and height dependent power factor appropriate to either RAOC or Base Load operation, W(2) or W(z)BL, to provide assurance that the limit on the hot channel factor, F (z), is met. W(z) accounts for the effects of normal oper-9 fationtransientsandwasdeterminedfromexpectedpowercontrolmaneuversover the full range of burnup conditions ie. the core. W(z)gg accounts for the more restrictive operating limits allowed by Base Load operation which result in less severe transient values. The W(z) function for normal operation and the l W(Z)gt function for Base Load Operation are specified in the CORE OPERATING ( LIMITS REPORT per Specification 6.9.1.9. l (2celaer- u th t ytewsmu,t.2 CATAWBA UNITS 1 & 2 B 3/4 2 4 5:n=ent 2. " (Uni t +)- 5:nc;nt 2. 50 (Un+t-Lt

I 1 i for Power Distribution Limits Bases i Attactiment 11 The limits en the nuclear enthalpy rise hot channel facter, FtH(X,Y), are

i specified in the CDIR as Mwinn A11cwabt. Radial Feaking limits, obtained by dividing the Mwi== Allowable 7btal Tsaking (MAP) limit by the axial peak (AXIAL (X,Y)) for location (X,Y) . By definition, the Mwlmn A11cwable Paiial Peakim limits will, for Mark-N fuel, result in a DER for the limiting transient that is equivalent to the DER calculated with a design FAH(XeY) value of 1.55 ard a limitig reference axial pcwar shape. The Mark-N MAP limits may be a@ lied to 0FA fuel, provided an appropriate adjustment factor is applied to pzwide equivalems to a 1.49 design Fdl(X,Y) for the OFA.

This is reflected in the MAP limits specified in the cotR. The telaxation of Fag (X,Y) as a function of 7HEFNAL PDWER allows changes in the radial power for all permissible control bank insertion limits. This relaxation is inplemented by the application of the follcuing factorst k = [1 + (1/RRH) (1 - P) ) whe.tw k = power factor nultiplier applied to the MAP limits P = THERMAL PWER / PATED THERMAL PWER RRH is given in the CDIR l l l l l l l ll

_~ _. _ ... . . . . - - ........ -. ~... - - - -. - __--.- ... - - .. - - . - . - . . 1 1 for Power Distribution Limits Bases o Attadiment 2: F M(X,Y,Z) ard FAHR M (X,Y) are reasured periodically, ard ocrparisons to the aklovable limit are made to provide reascnable assurance that the limitirq critaria vill not be me for cperation within the Twhnical Specification limits of Sections 2.2 (Limitirg Safety Systems Settirgs), 3.1.3 (M:vable Control Assemblies), 3.2.1 (Axial Flux Differen:2) , ard 3. 2. 4 (Quadrant Power Tilt Ratio). A peakirq margin calculation is performs:1 to provide a basis for decreasing the width of the AFD ard f(AI) limits ard for reducinJ 'DE:R%L POWIR. When an Fg M(X,Y,Z) measurement is obtained in accordance with the surveillance requirements of Specification 4.2.1, no uncertainties art

                                                                      ; the requirvd uncertainties are included in the ar. plied to peakirg                the measured limit.           When To f(X,Y,Z) is naasured for reasons other than meetirg the requirements of Specification 4.2.2, the measured peak is irenai by the radial-local peakirg factor to convert it to a 1ccal peak. Allowances of                             .

54 for measurement uncertainty ard 3% for manufacturirg tolerances a s then applied to the measured peak. When an FAHR M (X,Y) measurement is obtained , tigardless of the reason, no uncertainties are applied to the masured peak; the required urcertainties are incitded in the peakirg limit, i l r-rt-w -

1 1 POWER 0!$TRIBUT10N LIMITS %[ p[u.te a se 4 spo Ate BASES uQ HEAT FLUX HOT CHANNEL FACTOR easuk=#tMCM-t^^ ^ ^ " ^"'- ^ ^ ^ ^ ^ ^" AND NUCLEAR IhTHALPYRISEHOTCHANNELFACTORgnti univ 0 hen Reactor Coolant System flow rate and Fh are measured, no addi (a allowanegare necessary prior to comparison with the limits of the gure specified TWe COLR. Measurement errors of 2.1.% for Reactor ant System total flow rateh45 for F N have been allowed for in e ermination of the design DNBR value. , The measurement error for Rea tor Coolant stem total flow rate is based upon performing a precision heat b(aThqe and ing the result to calibrate the Reactor Coolant System flow rate indica , Potential fouling of the feedwater venturi which might not be detected cou) b the result from the precision heat balance in a nonconservative ma er. The ore, a penalty of 0.1% for undetected fouling of the feedwat venturi is in ided in the figure specified in the COLR. Any fouling whi ight bias the React Coolant System flow rate l measurementgreaterthan0.J4canbedetectedbymonito ng and trending various plant performance parame rs. If detected, action shall taken before per-forming subsequent sion heat balance measurements, i.e., either the effect of the fouling s be quantified and compensated for in the e(ctor Coolant System flow rat easurement or the venturi shall be cleaned to eN inate the fouling. ThM2-hour periodic surveillance of indicated Reactor Coolant Syst flow ( sufficient to detect only flow degradation which could lead to oper tio outside the acceptable region of operation specified on the figure spec-Q ed in the COLR. , 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-I tion satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. h The limit of 1.02, at which corrective action is required, provides_0NB AtI ,m - and linear heat oeneration rate _ protection with x-v olane Dower tilts / A ] ! m*

limit of 1.02 was selected.to provide an allowance for the uncertainty associated) 3 g (with the indicated power tilt e The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on s reinstated by reducino y[

the maximum allowed power by 3% tor each percent ' tilt in excess of For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incere detectors are used to confirm that the' normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incere detector monitoring is done with a full incore CATAWBA - UNITS 1 & 2 B 3/4 2-5 grjpgy.[ N H rw. i mms ii w ow. - gvunw (. /

for Pow.r Distribution Limits Bases Attachment 3: A peakirq in:rcase that reflecta a OA'1PRTT Kk'ER TILT PATIO of 102 is included in the generation of tne AFD limits. l l l

POWER DISTRIBUTION LIMITS BASES QUADRANT POWER TILT RATIO (Continued) j (uni t- b k flux sap or two sets of four speetric thimbles. The two sets of four symmetric da/8t thiables is a unioue set of eicht detector locationfihe normal locations areN C 8, E-5. E-11. H-3, H-13 L-5, L-11, N 8. ~ Alternate locations are availablej if any of the _ normal locations are unavailable.f 3/4.2.5 DNB PARAMETERS (Re vicete r

u -^dcded The limits on the DNS-related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the /ud L w+I initial FSAR assumptions and have been analytically demonstrated adtqyate to) maintain a design limit DNBR throughout each analyzed transient @The indicated value and the indicated pressurizer pressure value correspond to analytical T,yfts lie of 594 .

8'F and. 2205 3 psig respectively , with allowance for measurement 4 uncertainty, @ c.dd L3ce a , 4 Had The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their ' limits following load changes and other expected transient operation. Indica-tion instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1, 0 %nser+ l N F will be-ma4nte4eed-within4ts44mit+- rov 4ded-Condttrions-ar-though-4.- Mi F. a vre 12

            -abo #e-+r: ::inta4nedr- As noted on th64+gure-spee4f4ed-in-the-C4RE-OPERAHN9-
                                                                                              . rn e m vo w e n L-MRS REPORT (00t4h Reactor Coolant System flow ra,te and f-- may be " traded a low measured Reactor Coolant Sy* tem flow rate off"againstoneanother(i.e.(y nctMtre is acceptable if the ::::ur:d g is also low) to ensure that the calculated ONBR will not be below the design DNBR value. -The-re4tratica of ih a:                          ;
           -functfon-o f-4HE RMA L-90WE R-a H ows-c h a ng e s-i n-the-redf ei-powe r-e ha pe-f o r -aH-
            -psm4u 4W-red-+nsert4cn-Hsh 5.                         De rela konsh:p W etd en %n W re ma.es w ! et n   .

e ! . ., o: Ha 'e : .:> lac e ek on n. e s v e 'e s e Gn Mal py r; s e Aof' channel be. fee, ;~ ,n hec l.o.Gn gp3 , Cwc. mas i u.ne c) .

               @ k :ert v+                  a c,  ir ke ded               en  m1+ p ..

l l CATAWBA - UNITS 1 & 2 B 3/4 2-6 f :n h nt Ne. 5 (Unii 1)

NW 12 Lse rt af @ en p re cer ch 9 A nje . When Reactor Coolant System flow rate :nd F"u-;kmeasured,c.o. n s n sm addit.i,onal r: allowances geri'ded " th: are COLR necessary_ prior errors' e Measurement to comparison with of 2.1% for Reactor the limits of -t"; 'ip$ystem Coolant I total flow rate-end '% f;r .h den been allowed for in determination of the design DNBR value. The measurement error for Reactor Coolant $ystem total flow rate is based upon performing a precision heat balance and using the result to calibrate the Reactor Coolant System flow rate indicators. Potential fouling of the feed ater venturi which might not be detected could bias the result from the precision. heat balance in a nonconservative manner. Therefore, a penalty of 0.1% forf j m itI) undetected fouling of the feedwater venturi is included in%ne fisure etcif...="

      -in-the-COMr- Any fouling which might bias the Reactor Coolant $ystem flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before per-forming subsequent precision heat balance measurements, i.e. , either the ef fect of the fouling shall be cuantified and compensated for in the Reactor Coolant System flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

l \ l L l i

 -- -      - _. .       __.                --    .~         __   _ . - ~ , . - _ , .

3/4.2 POWER 0!$TRIBUT!0N LIMITS ( Qm + ?) BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core greater than or equal to design limit DNBR during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design critoria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded. The definitions of certain hot channel and peaking f actors as used in these specifications are as follows: Fn(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. 3/4.2.1 AXIAL FLUX DIFFERENCE (u n; t L) The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fq (2) upper bound envelope of the F TP limit specified in the CORE OPERATING LIMITS REPORT (COLR) times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux d'fference is determined at equilibrium xenon conditions. The full 1ength rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations. 7 CATAWBA - UNITS 1 & 2 B 3/A 2 + Amendment-Hor-74-(4n4-W 4eendsnt L. GS (Unit 2)

i POWER 0!$TRIBUT!0N LIMITS BASES l

                                  - At power levels below APLND, the limits on AFD are defined in the COLR,                                                       -l                  l 1.e., that defined by the RAOC operating procedure and limits. These limits                                                                                  ,

were calculated in a manner such that expected operational transients, e.g. , i lead follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of time allowed outside of the limits at reduced power levels will not result in signi- l

                       - ficant xenon redistribution such that the envelope of peaking factors would change sufficiently to prevent operation in the vicinity of the APLND p,,,7                                                                                 j level.                                                                                                                                                  -

At power levels greater than APL HO , two modes of operation are permis-- I sible; 1)LRAOC, the AFD limits of which are defined in tne COLR, and 2) Base Load operation, which-is defined as the maintenance of the AFD within a COLR-

specified band about a target value. The RAOC operating procedure above NO i s the same as that defined for operation below APLND.- However, it is 1

APL possible when following-extended load following maneuvers that the AFO limits may result in. restrictions in the maximum allowed power or AFD in order to

                       -guarantee _ operation with                                                 q F (z) less than its limiting value. To allow operation at the maximum permissible value, the Base Load operating procedure restricts b

i ( L I l S CATAWBA - UNITS 1 & 2 8 3/412-2 f.::nd::nt N:, 'A(Unit 1)

                                                                                                                                   ^;;ad;;at "e. SS (Unit 2) ;

t ,wp w y er>r - p;y-q- -m+ evg3 g y -r-,--,/- ->mm-p,-g--==w- -- m e-mm= -y w - ei- " w p-i-_--- m- _

I l l $ l POWER DISTRIBUTION LIMITS 1 BASES 4 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ontinue (a,y d} the indicated AFD to relatively small target band and power swings (AFD target band as specified in the COLR, APLNU < power < APLOL or 100% Rated Thermal Power, I whichever is lower). ForBaseLoadoperation!itisexpectedthattheUnitswill operate within the target band. Operation outside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of pe6 king factors would change sufficiently to prohibit continued operation in the power region defined above. To assure there is no residual xenon redistribution impact from past oporation on the Base Load operation, a 24 hour waiting period at a power level above APLND and allowed by RAOC is necessary. During this time period load changes and rod motion are restricted to that allowed by the Base Load procedure. After the waiting period extended Base Load operation is permissible. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are: 1) outside the allowed AI power operating space (for_RAOC operation), or 2) outside the allowed al target band (for Base Load operation). These alarms are active when power is greater than: 1) 50% of RATED THERMAL POWER (for RAOC operation), or

2) APLND (for Base Load operation). Penalty deviation minutes for Base Lead operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

, The limits on heat flux hot channel factor, coolant flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak l local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel elad temperature will not exceed the 2200*F ECCS acceptance l criteria limit. These limits are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position;

b. Control-rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; l

CATAWBA - UNITS 1 & 2 B 3/4 2-L. k ^nnhnt-Nc. ?4 (Unit 1) f u nin at h. 50(Unit 2)

I l l I POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL F ACTOR (Continued) (m,, g) c.

                                                               \         #

The control rod insertion TTmTts of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and

d. The axial power distribution, expressed in terms of AXIAL FLUX O!FFERENCE, is maintained within the limits.

FhwillbemaintainedwithinitslimitsprovidedConditionsathroughd. above are maintained. As noted on the figure specified in the CORE OPERATING LIMITSREPORT(COLR),ReactorCoolant$ystemflowrateandFhmaybe" traded off" ~against one another (i.e. , a low measured Reactor Coolant System flow rate isacceptableifthemeasuredF[q is also low) to ensure that the calculated DNBR will not be below the design DNBR value. The relaxation of F g as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. R as calculated.in Specification 3.2.3 and used in the figure specified intheCOLR,accountsforFhlessthanorequaltotheFhP limit specified in the COLR. This value is used in the various accident analyses where N F g influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed. The rod bow penalty as a function of burnup applied for Fh is calculated with the methods described in WCAP-8691, Revision 1 " Fuel Rod Bow Evaluation," July 1979, and the maximum rod bow penalty is 2.7% DNBR. Since the safety analysis is performed with plant-specific safety

            -ONBR limits compared to the design DNBR limits, there is sufficient thermal                              l margin available to offset the rod bow penalty of 2.7% DNBR.

The hot channel factor F (z) is measured periodically ar;d increased by a

          ' cycle and height dependent power facter appropriate to either RA00 or Base Load operation, W(z) or W(2)gt, to provide assurance that the limit on the hot channel factor, F (z), is met. W(2) accounts for the effects of normal oper-9 ation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. W(z)BL accounts for the more

restrictive operating limits allowed by Base Load operation which result in

            'less severe transient values. The W(z) function for normal operation and the W(Z)BL function for Base load Operation are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9.
                                                                            /D CATAWBA - UNITS 1 & 2                                 B-3/4 2-4            A:;nd:at N . " (Unit 1)---
                                                                                        %:nd::nt H: 68(Unit 2)

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR IContinued)(p n. t 2) K J WhenReactorCoolantSystemflowrateandFharemeasured,noadditional allowances are necessary prior to comparison with the limits of the figure , specified in the COLR. Measurement errors of 2.1% for Reactor Coolant System totalflowrateand4%forFhhavebeenallowedforindeterminationofthe design DNBR value. The measurement error for Reactor Coolant System total flow rate is based upon performing a precision heat balance and using the result to calibrate the Reactor Coolant System flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a nonconservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in the figure specified in the COLR. Any fouling which might bias the Reactor Coolant System finw rate measurer.unt greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before per-forming subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the Reactor Coolant System flow rate measurement or the venturi shall be cleaned to eliminate the fouling. The 12-hour periodic surveillance of indicated Reactor Coolant System flow is sufficient to detect only flow degradation which could lead to opera-tion outside the acceptable region of operation specified on the figure spec-ified-in the COLR. 3/4.2.4 OVADRANT POWER TILT RATIO (a nit p) The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-l tion satisfies the design values used in the power capability analysis. l_ Radial power distribution measurements are made during STARTUP testing and periodically'during power operation. The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. 1 The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided te allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing 9 the maximum allowed power by 3% for each percent of tilt in excess of 1. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore CATAWBA - UNITS 1 & 2 B 3/4 2 t

POWER 0!$TRIBUTION LIMITS JASES QUADRANTPOWERTILTRAT,3 ntinuedgui i 2 ) flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. The normal locations are C 8 E-5, E-11, H-3 H-13, L 5, L-11, N-8. Alternate locations are available if any of the normal locations are unavailable. 3/4.2.5 DN8 PARAMETERS (u n; + Q The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to j maintairs a design limit DNBR throughout each analyzed transient. The indicated

T,y value and the indicated pressurizer pressure value correspond to analytical j lie ts of 594,8'F and 2205.3 psig respectively, with allowance for measurement l

uncertainty. ! The 12 hour periodic surveillance of these parameters through instrument , readout is sufficient to ensure that the parameters are restored within their ' 1imits following load changes and other expected transient operation. Indica- 'i tion instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1. 1 I l CATAWBA - UNITS 1 & 2 83/42,h f.:: nt; n t-Her-Nr-(hM-1-)- f.:;nc.nt h a ("Mt4)-

1 1 1 l 1 ECCS TS MARK-UPS 1

4 i. l-l- i .

De eh & z 3/4.5 EMERGENCY CORE COOLING SYSTEMS _ % 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION

     . S.1' 1.1 Each cold leg injection accumulator shall be OPERABLE
          .                                                                                                    ith:           l

a. The discharge isolation valve open,

b. A contained borated water volume of between 785 and 8171 gallons,

c. oron concentration of between 1900 and 21 ppm,

d. A nit gen cover-pressure of between 385 d 481 psig, and

, e. A water i el and pressure channel OPE BLE. APPLICABILITY: ' MODES 1, 2, and 3*. (VH! Sys m Operable) , I ACTION:

a. With one cold leg infection a umulator inoperable, except as a result of a closed isolation alve boron concentration less than 1900 ppm, restore the inoperable cu.ulator to OPERABLE status within 1 hour or be in at least HOT STA BY within the next 6 hours and in HOT SHUT 00WN within the foll 6 hours,

b. With one' cold leg inje ion a umulator inoperable due to the isolation' valve bein closed, e her immediately open the isolation valve or be in at i st HOT STAND within 6 hours and in HOT SHUTDOWN within the followi 6 hours.

c. With one accumul tor inoperable due t boron concentration less than 1900 ppm and:

1) The vol me weighted average boron con ntration of the three limit g accumulators 1900 ppm o- 3reat r, restore the inoperable accu ulator to OPERABLE status n 24 ours of the low boron de rmination or be in at least . STANDB witMn the next 6 ours and reduce pressurizer pressure to ss t.'an 1000 psig ithin the_following 6 hours.

2) The volume weighted average boron concentration f the three limiting accumulators less than 1900 ppm but graa er than 1500 ppe, restore the inoperable accumulator to OP ABLE status-or return the_ volume weighted average boron concentr tion of the three limiting accumulators to greater than 1900 m and

, enter ACTION c.1 within 6 hours of the low boron deter nation

j. or be in HOT STANOBY within the next 6 hours and reduce res-surizer pressure to less than 1000 psig within the follow g 6 hours.

   " Pressurizer pressure above 1000 psig.

CATAWBA UNITS 1 & 2 3/4 5-1 Amendment No. 32 (Unit 1) Amendment No. 23 (Unit 2) q

             -                  .       - , . . . -                                                   -                       =.

OeRh- Ms EMERGENCY CORE COOLING SYSTEMS PT LIMITINGCONDITIO{04 OPERATION (Continued)

3) The volume weighted average boron concentration the three limiting accumulators 1500 ppm or less, return ne volume weighted average boron concentration of the t ree limiting accumulators to greater than 1500 ppm and e er ACTION c.2 within 1 hour of the low boron determinati or be in HOT STANDBY within the next 6 hours and reduc pressurizer pres-ure to less than 1000 psig within the f lowing 6 hours.

SURVEILLANCE REQUIR$MENTS 4.5.1.1.1.1 Each cold le injection accumulat shall be demonstrated OPERABLE:

a. At least once per 12 ours by:

1) Verifying, by the senc of alarms, the etntained borated

water volume and nit o n cover pressurc in the tanks, and

2) Verifying that each o leg injection accumulator isolation l valve is open.

b. At least once per-3 days and wi in 6 hours after each solution volume increase off reater than o equal to 75 gallons by verifying the boron-concen etion of the accu lator solution;

c. At least once er 31 days when the Rea or Coolant System pressure is above 200 psig by verifying that pow r is removed from the isolation y lve operators on Valves NI54A, NI65B, NI76A, and NI888 andthat/herespectivecircuitbreakersar padlocked; and

d. At le t once per 18 conths by verifying that ach cold leg injection accu blator isolation valve opens automatically nder each of the fo owing conditions:**

'han an actual or a simulated Reactor Coolant stem press'ure

           /) 2) signal exceeds the P-11 (Pressurizer Pressure B1 k of Safety Jnjection) Setpoint, and Upon receipt of a Safety Injection test signal.

4 .1.1.1. 2 Each cold leg injection accumulator water level and press e hannel shall be demonstrated OPERABLE:

   ^*This urveillance need not be p.rformed until prior to entering HOT STAND Y

/ following the Unit 1 refueling. CATAWBA - UNITS 1 & 2 3/4 5-la Amendment No. 32 (Unit 1) Amendment No.23 (Unit 2)

I

~'

EMERGENCY CORE COOLING" SYSTEMS

 ,                  . SURVEILLANCE RE0VIREMENTS (Continued)-
                                                                         ?    -

a'. At least once per s by-the ormance of an ANALOG CHANNEL OPERATIONAL TEST,'and a-

At least once months by the per ce of a CHANNEL t CALIBRATIO '

                                          ,,                        6  f 't .                 3 9%E-                                             ;

l es

(

y .. ! l . I o, t L. K If i e

                 - CATAWBA - UNITS 1 & 2                     3/4 5-lb-            Amendment No. 32 (Unit 1)       !

{ Amendment- No. 23 -(Unit -2)

o /

 ' #1    3/4.5 EMERCENCY CORE COOLING SYSTEMS
      ,  3/4,5.1 ACCUMULATORS COLO LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.1.1.2     Each cold leg injection accumulator shall be OPERABLE with:

a. The discharge isolatfor valve open,

b. A contained borated water volume of between 7704 and 8004 gallons,

c. A boron concentration of between 1900 and 2100 ppm,

d. A nitrogen cover pressure of between 585 and 678 psig, and

e. A water level and pressure channel OPERABLE.

, A,,PplICABILITY: MODES 1, 2, and 3*. -(U"! - physical 3.y disconnecte4-Cc144,6cr

                                                                                        %cumu'eters-end-dtschei ge-path.

suitably =edi'ied) ACTION: a. With one cold leg injection accumulator inoperable, except as a result l of a closed isolation valve or boron concentration less than 1900 ppm, restore the inoperable accumulator to 0PERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUT 00WN within the following 6 hours,

b. With one cold leg injection ac::umulator inoperable due to the isolation valve being closed, either irrnediately open the isolation valve or be in at least HOT STANOBY within 6 hours and in HOT SHUTOOWN within the following 6 hours,

c. With one accumulator inoperable due to boron concentration less than 1900 ppm and:

1) The volume weighted average boron cencentration of the three limiting accumulators 1900 ppm or greater, restore the inoperable accumulator to OPERABLE status within 24 hours of the low boron

! determination or be in at least HOT STANDBY within the next 6 hours and reduce-prc: cur!:+ vithinthefollowing6 hours.gpressuretolessthan1000psig& R ec.c -kor C.c e kw. S sbeVW l g

2) The volume weighted average boron concentration of the three limiting accumulators less than 1900 ppm but greater than 1500 ppm, restore the inoperable accumulator to OPERABLE status or return the volume weighted average boron concentration of the three limiting accumulators to greater than 1900 ppm and

         *bes(eurhec pressure above 1000 psig.

R c.c.h CATAWBA er Coo

                    - UNITS   1&2 %r  %s Ye m 3/4 5-2                                                                                                     -Amendmente-Me. 22 (Uni t 1)-
                                                                                                                                                           %endment Mc. 23 (Unit 2)-

m . .

                 , ;j 2$

EMERGENCY CORE-COOLING-SYSTEMS i 4 SURVEILLANCE REQUIREMENTS-(Continued) ; 74.5'.1~,'1.2.2 Eachcoldleginjectionaccumulatorwaterlevelandpressure ! channel shallebe demonstrated OPERABLE: a.. At leastionce per 31 days by the performance of an ANALOG CHANNEL ! 0PERATIONAL TEST; and.

b. At least once per 18' months by the_ performance of a CHANNEL-

                                                                        --CALIBRATION.                                                                   !

c. , j i

NO Ut0L y do j

                                                                                                               %o pacf                       '

I 4 1 u o L- -, 1: ll t i CATAWBA - UNITS 1 & 2 3/4.5-2b Amendment No 32 (Unit 1) Amendment No.23 (Unit 2)

i

i q l EMERGENCY CORE COOLING SYSTEMS- ; LIMITING CON 0! TION FOR OPERATION (Continued) ACTION:'(Continued)' , enter ACTION c.1 within 6 hours of the low boron determination , or be in HOT STANDBY within the next 6 hours and reduce + pew ' mri:= Apressure to less than 1000 psig within the following 6 hours. 8 ecsc.+c( Coo \ M t O g54e m . [ 3)- _The volume weighted average boron concentration of the three l limiting accumulators 1500 ppm or less, return the volume weighted average boron concentration of the three limiting accumulators to greater than 1500 ppm and enter ACTION c.2 within 1 hour of the low boron. determination or be in HOT STANDBY within the-next 6 hours and reduce p ::: W m pres- 3 sure toLless than 1000 psig within the following 6 ho M SURVEILLANCE REQUIREMENTS Aeqc.4or Coek v* 0[M i 4.5.1.1.2.1 Each cold leg injection accumulator shall be demonstrated ' OPERABLE:

a. -At least once per 12 hours by: !

          -1)      Verifying, by the absence of alarms, the contained borated              .
                 ' water volume and nitrogen cover pressure in the tanks, and             )

2) Verifying that each cold. leg injection accumulator isolation valve is open,

b. At least once per 31 days and within 6 hours af ter each solution ;

volume increase of greater than or equal to 75 gallons by verifying 1

          'the boron concentration of the accumulator solution;-
                                                                                 ~

g

c. At least once per-31 days when the Reactor Coolant System pressure-

           ,is above 2000 psig by verifying that power is removed from the isolation valve operators on Valves N154A, NI65B, NI76A, and NIS8B and that the respective circuit breakers are padlocked; and                   ,

d. At 'least once per 18 months by verifying that each cold leg injection accumulator isolation valve opens automatically under s each ,of the ' following conditions:**

1)- When an actual or a simulated Reactor Coolant System pressure signal. exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) Setpoint, and 2). Upon receipt of a Safety Injection test signal.

- ihis surveillance need not be performed until prior to entering HOT STANDBY tellowing the Unit 1 refueling.

CATAWBA - UNITS 1 & 2 3/4 5-2a f :adaent No.02 (Unit 1)- ' f::ndment No.20 (Unit 2)

D e l& +6 l EMERGENCY CORE COOLING SYSTEMS Y - l

   -UPPER HEAD INJECTICH-                               bbb Mcl [ b'[$.  $/q      t-l wi Me rvH o

U '

L MITING CONDITION FOR OPERATION /

3.5 2 Each Upper Head Injection Accumulator System shall be OPERABLE with:

a. The discharge isolation valves open, /

b. ;A minimum' contained borated water volume of,1807 cubic feet,

c. A oron concentration of between 1900 and 2100 ppm, and

d. The itrogen-bearing accumulator pressuriz to between 1185 and 1285 sig.

APPLICABILITY: ES 1, 2, and 3.a ACTION:,

a. With the Upp Head Injection Adcumulator System inoperable, except as a result o closed icolatiofi valve (s), restore the Upper Head.

                 -Injection Accum lator ?,steVto'0PERABLE status within 1 hour or be
                                              ~

in at least HOT ANDBY witfiin the next 6 hours and in HOT SHUTDOWN < within the followi 6ho/rs.

          ^b.      With the Upper-Head      jection Accumulator System inoperable due to the isolation isolation     valve valve (s) (y)b pr eing     closed, in HOT      either within STANOBY    immediately 6 hours open and bethe in HOT    q SHUTDOWN within t      next      hours.                                           *

SURVEILLANCE REQUIREMENTS 4.5.1.2 Each Upper ead Injection Accumu tor System shall-be demonstrated OPERABLE:

a. . Atlea/tonceper12hoursby: s 1)

                        /Verifying the-contained borated w er level ~in the surge tank i

and: nitrogen pressure in the acedm ators, and

2) Verifying +. hat each accumulator disch ge isolation valve is-open.

             . At least once per 31 days and within 6 hours a er each solution volume increase of greater than or equal to 138. gallons by-verify-ing the boron concentration of the solution in th water-filled accumulator;
     " Pressurizer pressure above 1900 psig.

CATAWBA --UNITS 1 & 2 3/4 5-3 32 Unit 1) Amendment No. 23 ((Unit Amendment No.

i l EMERGENCY CG8E COOLING SYSTEMS ! i Ns

       . SURVEILLANCE REDUIREMENTS (Continued)

N / c.- At least once per 8 months by:

1) Verifying that eac ccumul or discharge isolation valve-closes automatically w he_ water level'is 93.2 2.7 inches !

                        -(Unit 1) and 93.1 2 2.7        est(Unit 2) above the working line         ;

on the water-filled a amula ,- and

2) Verifying that the otalLdissolved trogen and air in the water-filled acc ulator'is less than -scf per 1800 cubic feet of water huivalent to 5 x 10 5 po of nitrogen per pound of wate ,

d.- At least once r 5 years and if the requirements of S cifica-tion 4.5.1.2 ) are not met by replacing the membrane-installed I between the ater-filled and nitrogen-bearing accumu~lators, i

                                                         $ 5.) &G I

b k j --),. CATAWBA - UNITS 1 & 2 3/4 5-4 Amendment No. 32 (Unit 1) Amendment No. 23 (Unit 2)

s EMER6ENCLCORE COOLING SYSTEMS -

            -UPPER HEAD INJECT!
           ;(Deleted upon the~ physical;discon3 set                      the-UHI System from the Reactor.

Coolant' System)

                         /               }                                                                                  >

w C)- :; l . ,s

  ..-                                                                                                                      -j l;

i

j i

                                                                                                                        .j f
                                                                                                                         .i 4.

1 1 L -4 L_ - l .. L. c LCATAWBA - UNITS.1 &-2 3/4 5-4a mendment No. 32 (Unit 1) l~ Amendeent No. 23 (Unit- 2) L

EMERGENCY CORE COOLING SYSTEMS r 3/4.5.2 ECCS SUBSYSTEMS T,yg > 350*F LIMITING CON 0! TION FOR OPERATION 3.5.2 Two independent Emergency' Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging-pump,-

b, One OPERABLE Safety Injection pump,

      .c.      One OPERABLE residual heat removal heat exchanger,

c. One-OPERABLE residual beat removal pump, and

e. An OPERABLE flow patn capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump.during the recirculation.

phase of operation. APPLICABILITY: M00ES.1, 2, and 3. ' ACTION: ,

a. With one ECCS suosystem inoperable, restore the-inoperable subsystem to-OPERABLE status witnin 72 hours or be in at least HOT STANDBY within the next 6 nours and in HOT SHUTDOWN within the following 6 hours, b., In tne event the ECCS is actuated-and injects water into the Reactor '

Coolant System, c Special Report shall be prepared and submitted to the Commission oursuant to Specification 6.9.2 within 90- cays describ-- , ing the _ circumstances of the actuation and the total accumulated ~

              . actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special' Report whenever its value exceeds 0.70.

O bM t% P e

 -CATAWBA -~ UNITS 1 & 2                    3/4 5-5

a

                                                              &O NW               &

EMERGENCY CORE COOLING: SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each-ECCS subsystem shall-be demonstrated OPERABLE: a, At least once per 12 hours by verifying that the following valves are 'in the indicated positions with power to the valve operators removed: Valve Number Valve Function Valve position NI-162A Cold Leg Recire. Open NI-121A Hot Leg Recire. Closed NI-152B Hot-Leg.Recire. Closed NI-1838 Hot i.eg Recirc. Closed NI-173A Residual Heat Open Removal Pump Disch. NI-178B Residual Heat Open Removal Pump Disch.

                  -NI-100B             Safety-Injection       Open Pumo Suction from Refue' ling Water Storage Tank NI-147B'           . Safety Injection      Open                 -

Pump Mini-flow ,

b. At least once per 31 days by: s I

1) Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discnarge piping high points, and

             '2)     Verifying that each val;e (manual, power-operated _ or automatic) in the flow path that is not locked, sealed, or otherwise secured-in position, is in its correct position.                    .s

c. By'a-visual inspection which verifies that no loose debris (rags, 1 trash, clothing, etc.) is present-in the containment whien could be '

l transported to the-containment sump and cause' restriction of-the L pump suctions during LOCA conditions. This visual inspection shall

              'be-performed:
                                                                                          ]

L 1) For all accessible areas of the containment prior to establish-

 ;                   ing CONTAINMENT INTEGRITY, and

! 2) Of the areas.affected wit'hin containment at the completion of l each containment entry when CONTAINMENT INTEGRITY is established. At least once per 18 months by:

d.

1) Verifying automatic isolation and interlock action of the residual heat removal-system from the Reactor Coolant System by ensuring'tnat:

a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the interlocks prevent the valves from being opened, and CATAWBA - UNITS 1 & 2 3/4 5-6

hO du O EMERGENCY CORE C00 LINO SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 660 psig the intee. locks will cause the valves to automatically close.

2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion,

e. At least once per 18 months, during shutdown, by "

1) Verifying,that each automatic valve in the flow path actuates to its-correct position on Safety Injection and Containment

 .. .                    Sump Recirculation test signals, and

2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection test signal:

a) Centrifugal charging pump, , b) Safety Injection pump, and , i L c) Residual heat removal pump. -

f. By verifying that each of the following pumos develops the indicated differential pressure when tested pursuant to Specification 4.0.5:

1) Centrifugal charging pump t 2380 psid, 2); Safety Injaction pump 1 1430 psid, and

3) Residual heat removal pump- t 165 psid.

l: ! '- g. By verifying:the correct position of each electrical and/or mechanical

stop for the following ECCS throttle valves:
=- ,1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve wnen the ECCS suosystems are required to_be OPERABLE, and: 2) At least once per 18 months.

Centrifugal Charcing Pume Injection Tnrettle Safety Infection Throttle Valve NumDer Valve Numcer-NI-14 NI-164 NI-16 HI-166 HI-18' NI-168 NI-20 NI-170

       "" inis surveillance .need not be performed until prior to entering HOT SHUT 00W following the Unit One first refueling.

CATAWBA - UNITS 1 AND 2 3/4 5-7

                                                                                                                        ~
      ,                                                                       s                          ,

EMERGENCY CORE-COOLING SYSTEMS-- ., gr, l LSURVEILLANCE REQUIREMENTS 1(Continued)' l

Jh, By performing a . flow balance test, during snutdown, following com-pletion of- umdifications to the ECCS subsystems that alter the-subsystem ficw characteristics and verifying that:

                                     -1)      For centrifuga11 charging pump-lines, with a single pump runningi             l a) ' The sum of the. injection _ line flow rates, excluding the highest flow rate,_ is greater than or equal to 33; som, and                                              :NI b)     The total pump flow rate.is less than or equal-to 565 gpmi-2). For Safety Injection pump lines, with a single' pump running:                j La)-    =The sum of=-the injection line flow rates, excluding the              -l
                         .                          highest flow rate, is greater.than or equalLto         gpm,-               .

and-

                                             'b). LThe totalLpump.. flow rate'is less than or equal to 660 gpm,
                                                                                                                            .y 13 )-  ' For: residual Lheat removal pump lines, with a .siogle pump.                  ,
                                             ; running,7the sum of-the . injection line flow ratesJis greater-      ,
                                             -than.ortequaltto:3648 gpm.                                                    j

L ,

s

-n .

r]

          -                                                                                                                ,i

( l J ' t J L! m k i

                     . CATAWBA?- UNITS 1 &~2                           3/4 5-8                                             -A

a 4 4 - 4'. 4 4>- - . .- ha-,e -w a s. . - _ #< -4u,* c .L+>,s - . -i-, ,2 i 5 6 1 1 ADMINISTRATIVE TS MARK-UPS i

1 5 s i !S' i 1

i. t I ' _t l'

?-

i. -

P 1

4

ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)

The Radioactive Effluent Release Reports shall include a list ano description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active materials in gaseous and liquid effluents made during the reporting period. The Radioactive Effluent Release Reports shall iieclude any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (00CM), n well as a listing of new locations for dose calculations.and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2. MONTHLY OPERATING REPORTS

       ~ 6. 9.1. 8 Routine reports of operating statistics -and shutdown experience, in-ciuding documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555, with a copy to the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report.  -

CORE-OPERATING LIMITS REPORT

       - 6. 9.1. 9 Core = operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1. - Moderator Temperature Coefficient BOL and EOL limits-and 300 ppe surveillance-limit for Specification 3/4.1.1.3,

2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

_3.- Control Bank Insertion Limits for Specification 3/4.1.3.6, T 4 -AxialFluxDifforenceLimits,targetbandfandAPLHD y,p / Specification 3/4.2.1,

       '5.       ! Heat Flux Hot Channel Factor, F RTP     K(Z).W(Z)$PLNM,Mnd W(Z)gg forSpecification3/4.2.9,and            gggpg

6. Nuclear Enthalpy Rise- Hot Channel Factor,dFhP***n# d Power Factor \

Multiplier,MF],MimitsforSpecification3/4.2.3. ) The analytical methods used to determ1ne the core operating limits shall be those previously reviewed and approvod by HRC in:

1. WCAP.-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION HETH000 LOGY,"

            - -July 1985 (W Proprietary).

(Methodology for Specifications 3.1.1.3 - Moderator Temperature . Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, ' jr:sp . 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux ' A rt>c4 w t / CATAWBA - UNITS 1 & 2 6-19

                                                                                 ".;;;;nd.;ent ncent 2.SO2. N (Unit (Unit 2) 1)

for Specification 6.9.1.9 Attachment 1:'

      Reference 5 is not applicable to target band and APL*.

References 5 and 6 are not applicable to V(2), APL*, and W(Z)st, Reference 1 is not applicable to F6HR .L Reference 5 is not applicable to Fj'8 and MFa. t 4 t 1

ADMINISTRATIVE CONTROLS E CORE OPERATING LIMITS REPORT (Continued) Difference, 3.2.2 Heat Flux Hot Channel Factor, and 3.2,3 - s Nuclear Enthalpy Rise Hot Channel Factor.)

2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ /

SQVEILLANCE TECHNICAL SPECIFICATION," June 1983 (W Proprietary). (Methodology for Specifications 3.2.1 - Axial Flux Difference \ (Relaxed Axial Offset Control) and 3.2.2 Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FqMethodology.)

3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING 8 ASH CODE," March 1987, (W Proprietary), s

              -(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)_        k e The core operating limits shall be determir.ed so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS

-[ limits, nuclear limits such as_ shutdown margin, and transient and accident analysis limits) of the safety analysis are met. - L The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and

Re-ident Inspector.- , O m ~ m. .~. -... ~

                 .-               __       .~ . _ ,           ..            . . ,_.. .   .       _ _ _ .     - .~._ . . . _ . _ .

for Specificction 6.9.1.9 ;

        ,           ' Attachment'2

4. - SAW.10152. A " NOODLE . ~ ALMulti Dimensional Two. Croup Reactor Simulator ,"

June 1985.

                                     '(Mathodology for Specification 3.1.1.3             Moderator Temperature Coefficient.)                                                                                      :

i' BAW.10163P.A.. " Core Operating Limit Methodology for Westinghouse.

                     .5._

Designed PVR's ," June .1989. (Methodology. for Specifications 3.1.3.5 . ' Shutdown Rod Insertion

                                     . Limits, 3.1.3.6 .' Control Bank Insertion Limits, 3.2,1             Axial-

3. 0 Flux Difference, 3.2.2 Heat Flux Hot-Channel Factor, and 3.2.3 .

                                      -Nuclear.Enthalpy Rise Hot Channel Factor.)

3 BAW 10168P,- Rev,1, 'B6V Loss of.Ceolant Accident Evaluation Model for ;i

6 '. Recirculating Steam Cenerator Plants,"_ September, '1989.:

(Methodology for Specification 3.2.2 ' Heat Flux Hot Channel-Factor.); A 4 9 I 3'

                                                                                                                                     'l i

1 i l L, 1 h l'. ' i [-

p I L L l l y p

4-t- ^ ^ " - * ~ " " '--'

                                         " " '

C'" * * * ' " - " " - " * " - ' ' ' ' - - - - " * - *

                                                                                     -                                * * * - " " * ' " - * *       ' ' - ~ ' - * * **      "

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                                                                                                                                                -m+-   r        -      ,.-

J .2_a_J - m m; ha A 1 4 SAFETY LIMITS AND POWER DISTRIBUTION TECHNICAL JUSTIFICATION AND NO SIGNIFICANT HAZARDS ANALYSIS . . i o

                                                < i.

i i f 4 i

1 LProposed Technical' $pecification Revision Fioure 2.1-1 a 1

   ~                                                                                 ..

This proposed Technical _ Specification -(TS) revision changes Figure 2.1 to i reflect; use- of the BWCMV2 CHF correlation- and LB&W Fuel Company's Statistical f' Core DesignL(SCO); methodology _with a 1.50 thermal design limit, The new figure . applies -to- Unit 1 - only. The figure numbers and references in T.S. , 2.1.1--have_been changed to indicate the correct Figure 2.1-'1 for each unit. l Technical Justification

                - With: the first- batch implementation of its. Mark-BW fuel design, B&W = Fuel
                   -Compapy-(BWFC) has ' recalculated the Catawba reactor core safety limits using:

its ~ BWCMV CHF correlation = along .with its' Statistical Core Design' (SCO)- I methodology. .Withlthe implementation of:' these design methodologies for Ethe d Catawba icore, it was - possible to increase the.-nuclear enthalpy rise _ hot [ N

                , channel' fdctor, F g, from .1.49 to 1.55 for the Mark-BW fuel to allow- greater
                  . fuel" cycle design- flexibt.11 ty.                                                The -nuclear. enthalpy- rise hot channel factor, iFh,[has been maintained at' 1.49: for. tho' Westinghouse: 0FA's, . The pro' posed changes: to Figure _2.1-1_ reflect the~.use of this new design limit _as well Las theLuse'of BMCMV and SCD,                                                                                                                                   1

i The :ireactor core: safetyL _ limits -

provided on Figure. 2 J1 depict the 4 zcombinations : of : thermal Jpower, _ reactor cool _ ant system pressure,- and : average

                    . temperature :below which the~ calculated DNBRLis no 41ess Lthan the design 111mit                                                               = -

iDNBR E.value, torj the average L enthalpy ~at f ther vessel ext t ~ ls less thanythe ' fenthalpyL of saturated liquide . The. analysis . which defined . these . limi ts'.was: 1 L based on:~ad f ull core ofiMark-BW~ assemblies withf a - thermal design flow ' rate' '

                ' that1 bounded 'the minimum measured flow. at' Catawba. The DNB limited portions'                                                                              l of these: curves wereJ defined usingL the BWCMV -CHF : correlation with- a' thermal                                          -
                ; design : limit of 1.50. This 1.50 -l thermal design -limit provided 10' percent -

cthermalomargin_ 'to the 1.345 BWCMV . statistical design limit which was . defined forfthe Catawba coretusing-the:BWFCLSCD methodology. The safety limits' were: based ton aEdesign' peaking distribution: wi th - 'a nuclear enthalpy rise . hot" channel factor, ' F H, f -155 and a reference cosine axial power shape wIth 'a , peak- of. 1.55. To veri fy that this design peaking = distribution was-conservative on'a cycle-specific basis,. maximum allowable peaking (MAP) limits

                !WNP337/1.
         . c.           ~,,e                         -.---+.ec            ,,.---er    ..,-r,.

my, = w g. ,., r - se +- <r ei y -v-,--- -sy v. 3 4* * - < -

that provided DNB equivalence to the desiga distribution at various safety limit statepoints were defined. To verify that margin was available, these MAP limits were then compared to the cycle-specific peaking distribution. As part of the safety limit / MAP limit analyses, an evaluation was. performed which showed that if power was reduced below 100 percent, peaking could be increased according to the following relationship: k = 1 + 0.3(1 - P) where k = the factor by which the MAP limits are adjusted to define reduced power limits p = the fraction of rated power Comparison of the Mark-BW safety limits to the Westinghouse OFA safety limits th'a t they are replacing shows that - at all points the Mark-BW limits fall outside the OFA limits. As stated in the bases for the OFA safety limits, the 0FA curves were based on a nuclear enthalpy rise hot channel factor of 1.49. Mixed core studies have shown that the new Mark-BW safety limits are also applicable to the Westinghouse 0FA's if the nuclear enthalpy rise hot channel factor fur the 0FA's is maintained at 1.49. To further ensure the applicability of the Mark-BW safety limits to the OFA fuel, the Mark-BW maximum. allowable peaking (MAP) limits were adjusted at all points by the ratio of'the design peaks (1.49/1.55). Mixed core studies verified.that this paaking adjustment is conservative for the OFA. Proposed Technical Specification Revision Table 2,2-1 In footnote to Table 2.2-1, Reactor- Trip System Instrumentation Trip Setpoints, loop minimum measured flow is changed from 96,900 gpm to 96,250 gpm (Unit 1 only). For Power Range Neutron Flux,.High Setpoint and low Setpoint, both Z- values change from "4.56" to "5.92", where Z is the residual channel uncertainty when the calibration and drift components have been removed. The High Setpoint ALLOWABLE VALUE changes from "111.1%" to "110.9%". WNP337/2

J The High - Pressurizer Pressure- Z value changes from "4- 96" to "0.71" and the - _ Sensor Error ( s) . f rom . "O 5" to _ "1. 5". - The- Low Reactor Coolant: Flow TOTAL

             -ALLOWANCE changes from "2.5" to;"2,92", the Z value from "1.41" to "1.48",-and                               i the ALLOWABLE'VALUE from:"88.8%"-to "88.9%". The Overtemperature' AT Z value changes from~"5.41"Lto "3.0", the S value from "2.65" to 2.12", and the-TOTAL ALLOWANCE f rom "8.9"               to "6.98". The ' table has also been updated -to reflect deletion: of the: RTO Bypass System.

i Technical 4 Justification lable.2.2-1 =1

~!

The loop minim'u m--measured flow at Catawba is contained in the footnotes of ' Technical _ Specification Table 2.1-1. Beginning with Catawba 1 Cycle 6, the - value'specified is being reduced from 96,900 gpm-to 96,250 gpm. These values,

             -which -are one-fourth .of the total design _.RCS: flow, reflect a reduction in- the i

nominal thermal--design flow rate-from 387,600-gpm to 385,000 gpm. LFor Catawba 21 Cycle 6, all safety and operating limit thermal-hyoraulic analyses'have been i

Ebased.on-a nominal thermal. design flow rate of 385,000 gpm. .For.SCO analyses, this-value was;used as.is, for non-SCO analyses this value-was reduced:by 2.2 ( percent. t Changesf made to Lthe' high and ~1ow' power range : neutron flux _,f high7 pressurizer;

             ' pressure, 'andJlow RCS" flow trip functions' are being -made toi reflect -both, the-                       !

JB&Wisafety- analysis Lassumptions -as well as revised' instrument _ uncertainties. '

New safety analysis assumptions for _t'he Low Reactor Coolant Flow? and Over- -

f.emperature : AT reactors tripsx necessitate changes to . the TTOTAlc ALLOWANCE for. g these trip: functions'since it was desired.to maintain the plant trip setpoints: l Unchanged..- i Pl ant-speci fic.. instrument uncertainty -icalculations- Tare ^ the -

.]

technical -basisc for the changes to the ALLOWABLE:V^LUE, and Z cterms tor Eall }

           = five-; trip functions. : Thel ALLOWABLE VALUES for- both; the 3 Power LRange Neutron Flux High Setpoint and;the- Low- Reactor < Coolant Flow tripsL are conservativelyL                         l being _made'more restrictive as a' result of this error. calculation.

WNP337/3

l Siand Z are less significant parameters which may be employed to determine the operability of the channel should the trip setpoint drift beyond its ALLOWABLE - VALUE. . The net effect of the changes to the S and Z values for the High Pressurizer Pressure Low Reactor Coolant Flow, and Overtemperature d trips is _ to: permit' a -larger setpoint drift before the channel. must be declared inoperable. -This i s, in effect, increasing the margin between the error adjusted trip setpoint and the safety analysis assumption. Conversely, the l proposed changes to the Z values for the High and Low Power Range Neutron Flux

                  'setpoints; conservatively restrict the permissible rack dri f t.                             Since this action : statement provision has never been taken advantage of at Catawba, no I

past operability determination is invalidated by this change. ;

i Since deletion of the RTD Bypass System has been completed on Catawba Units 1 ' and. 2' Table 2.2-1 has been ' updated to reflect only those ' values which are applicable upon deletion of the RTD Oypass System as indicated by "#". This change-is administrative only. ! Revision to Technical Specifications Table 1-The- f ( AI); function was- set to zero for .the OPAT setpoints. Discussion-o , The OPATfsetpoints provide protection for Centerline Fuel Melt (CFM) as stated

i n . BAW-10163 P-A.- This reference states that either f t (AI) or f ( AI) could be used Lin the, safety setpoints if either f(AI) function protected both DNBR and-l' CFM limits. DIt has been shown for -Catawba - 1 Cycle 6 that -the OTAT fi (AI) functf on and the OPAT1 trip function- without f ( AI) ' provide both DNBR ~and CFM

protection, The OTATL3 f ( AI); function provides imbalance- protection - and the.

l! OPAT -provides -overpower protection.. The .exampleHtechnical specifications in F 20AW-10163P-A are based on . f (AI) setpoints and these setpoints are adjusted ! when negative CFM margins are' calculated. Since. f (AI) providesuthe imbalance i protection instead of f(Al), f (AI) would- replace f ( A1) in this 3 specification. However, changes to the tf (AI) function during operation are difficult and undesirable. Therefore, the reduction of f ( AI) for negative i i WNP337/4 i

 ,                                                                                                                                 4
                                                 ,a -             ,     ,,_      _      -        -                   a

        . ._         .         ._ - _ . . _ _            __    __            . _ _ - _ _ . _ _             ._m.

CFM margins was replaced by La reduction of the : K value in ~ the OTAT trip - function.. " i

                                                                                                                         .)
                        -Technical Justification
                                                                                                                        .+

The' Technical. Specifications presented herein reduce _the Ki value of the OTAT h

setpoints by' an ; amount thatiis equivalent to reducing f(AI) function '- for-~

negative CFM margin calculations.

                                              -                                                                              1
                         ' Revisions . To Power Distribution Specifications - 3/4.2.1,- 3/4.2.2, -3/4.2.3, 314,2-4-, , and 3/4.2.5                                                                        l r

The current: Power Distribution TS have been changed to reflect applicability to Unit 2'only. This has'been done!by placing "(Unit 2)" in the Applicability- ' I

Section: of! the current TS. The Unit .2 Power Distribution TS will.. have "B"- .

placed inifront: of- the page number (Ex: 3/4 B2-76), and the Unit- 1 Power f SDistribution- changes will:have "A" placed in front of the page numbers. The '[

Unit 1-TS will be changed.as marked up and copied on white paper,- The Unit 2-

Power Distribution TS will-be changed as described and copied on yellow paper. '

iThe Power -Distribution TS will b'e ; separated by Unit,- anil placed.-on -dif ferent -

                                                                                                                        ;j fcolored:paperJduringlthe-period of time when the Unit'TS are difforent because-
                       -:of7 the change !toLMark-BW fuel,' tothelp ensure that the :TS lare appl-led to the-
      ,                  correct Unit.:         '

l n }

Revisions' to Techhical: Specification 3/4.2.1 J

i n ~

The target AFD forL Base Load -operatio'nLand the RA0C olimits have~ been replaced

j f with :an; envelope of a11 owed iAFD v~alues at various ~ power . levels. ' The AFD !! b, c setpoints given in. the COLR- replace the RAOC. operating space referred t'o in' the current. Specification; Since'the reactor is-not constrained tonoperate at

a :specified:; target AFD, the . target AFD and' associated band have been eliminated fromLSpecification' 3.2.1. The allowable operating AFD space is-l.

l: L 'WNP337/5 L : L , -

anywhere within AFD setpoint envelope in the COLR. This change applies to Unit 1 only. Technical Specification Justification 3/4.2.1 Specification 3.2.1 was revised to provide an LC0 statement and required actions consistent with BWFC methodology for core power distribution control as discussed in BAW-10163P-A (Reference 1). During the reload ' licensing analyses for Catawba Unit 1, cycle 6, BWFC performed a three-dimensional maneuvering analysis to Jetermine the Axial Flux Difference (AFO) limits based cn the methodology of BAW-10163P-A. Specification 3/4.2.1 was revised for consistency with the new analytical methodology and to reflect the results of the cycle 6 analysis. The resulting. AFD setpoints were.placed in the Core Operating Limits Report (COLR). The AFD limits of Specification 3 2.1 prevent the core power distribution from exceeding. the allowable values based on the LOCA peaking limits and the initial condition DNB maximum allowable peaking (MAP) limits during power operation, The AFD limits are defined by a three-dimensional core maneuvering analysis that determines core peaking dependence on core loading, fuel. depletion, thermal-hydraulic statepoint, control rod position, and xenon distribution. Correlations between peaking margin and axial power offset are developed that allow determination of negative and positive offset limits at selected power levels. The resulting offset limits preclude operation with negative margin, and are translated into corresponding AFD limits. The peaking margins are calculated from augmented nodal peaks calculated as described in BAW-10163P-A. The margin database comprises calculations from the entire range of power distributions generated in-the maneuvering analysis, including control bank insertion to the insertion limit and transient xenon conditions. Separate K(Z) peaking limits for Mark-BW fuel and 0FA fuel were used in the maneuvering analysis to compute LOCA margins. The AFD limits determined for cycle 6 were adjusted for measurement uncertainty. The adjustment was applied -WNP337/6

     .m         -._             ._      _             -    _ ._ ._._ _ . __. _ _._._ __ _                             . _. ._ _

k > at ,each ' power Ilevel between - 100% and . 50% of rated thermal power. The AFD setpoints given in the cycle 6 COLR are these adjusted values of the limits.

                      ;K(Z) limits for the Mark-BW' fuel were determined by- the ECCS analysis performed- for Mark-BW fuel, as documented in BAW-10174 - (Reference 2).                  The
                       ,K(Z) flimits for' the OFA fuel- are the current Westinghouse K(Z) values for a

Catawba,-_as given in_the current Catawba-COLR. Initial conwition 'ONB peaking margins were computed from the augmented peaks ' and L the = MAP limits based on' Statistical Core Design (SCO) methodology, as described in BAW-10170P-A (Reference 3). The MAP limits - are a family- of

                     . peaking' limits for - which -either- the minimum .ONBR is equal to the t hermal
                      -design limit, or_~the coolant quality'at the minimum DNBR_ location is. equal'to-                             !
                      -the CHF correlation. quality limit. The MAP-limits provide linkage between the
                      'DNBR analyses, with their design peaking distributions, and the co'e operating limits..-The operating-limit maps-are' based on the statepoint-thst represents                                i Lthe point of ' minimum: DNBR during the most . limiting . non-0 TAT transieut. 'To

ensure ' applicability .of the _ Mark-BW MAP limits to the OFA fuel, ths "'rk-BW --

iMAP . limits 'were adjusted at all points by the ratio of.- the design peaks 1(1 49/1.55).. Thermal / hydraulic analyses Lof _ mixed _ cores verified that' this peaking: adjustment-is conservative for OFA-fuel' . 4 References-I

                      .1 -     BAW-10163P-A'        ,   . Core Operating Limit Methodology for Westinghouse-Designed:

PWR s., . ' J u n e , 1989.

                    -2.-      .BAW-1017_4, _ Maa BW Reload LOCA Analysis ~for Catawba and McGuire.

l E '3. BAW-10170P-A, Statistical Core Design for Mixing Vane Coros, December, 988, a J . 1 ns to Technical Specification 3/4.2.2

This-c'hange applies to Unit'1 only.

(WNP337/7- -

 -           -       -         .    , ..             . .    - ..     - ~ -        . - -                        . .   - . - . ~ .       . . - . _ .
                                                                                                                                                              ,k 1

1 i

                         . Specification L3/4.2.2--was revised to-_ reflect the power peaking surveillance
                         ; method.describedLin BAW-10163P-A,_              These revisions are summarized as follows:

l1.= ;The statement of 'the LCO was revised to reflect the-nomenclature for. the heat flux hot channel factor [(F (X,Y,Z)] used =in--BAW-10163P-A- and n [

                                  - throughout the' Reload Report. Also, as discussed above,. separate K(Z)
                                 ' curves are provided for the different fuel types-(Mark-BW and 0FA),-                                                     j i

2.- Action.a'in the current l specification ~has been replaced by Actions'a, b,- and _ c inl the new specification'. The thermal power reduction required , when F g (X,Y,Z) exceeds its= limit are the same as the current requirement,- as isL the reduction l required in the OPiT trip setpoints. - Action is a - ! new l requirement, and is provided to limit _the allowable AFD shen i Fg (X,Y,Z) exceeds .its = limit. This reduces :the possibility of operating _ the! core in - excess 'of the F (X Y,Z)" , limit _ when a. margin calculation-9 (discussed in. item 7;:below) indicates _ negative operational margin exists. t 3 4

                         -3.       There is no change to SR 4,2.2.1.                                                                                            [

t c4 ? SR:-4.2.2.2 addresses obtaining ; an incore' flux map and the . requirements ! .;

b'a sad ) on the result's of-the measurement. The reference to RAOC operation- [

has- been Edeleted. . since-.RAOC operation is unique 'to Westinghousei-1 4

y , kethodology.

                                                                                                                                                            .y
                        - 5,:     There is,no. change to.'SR 4.2.2.2,a,

(

                         '6i       SR~ 4,2i2;2.b           in 'the current? surveillance has been deleted.                         -The
                                 - allowances for' meJsurementJuncertainty and'manuf acturitir. tolerances have -                                           j Lbeen g includedJin th 611mit-[FQ,(X,Y',1)]7and therefore the measured peak j,
                      ,          sFq(XiY,Z)isnotincreas6d-bythesefactors.

9-D7. 'SR -4.2.2.2.cn in the current surves1;ance has been deleted. No simple determination is made of only whether or' not 'the limit has been exceeded. Instead,_;ihe: amount by which the 4.2.2.2 measured value is- above or below the limit is' qualified as detailed below. L 1 l . L ' i' L

WNP337/8

                                                                                                ,  - - - . , ,     .,,o          -      -
                                                                                                                                               -w-- r2 .. -

'8. SR 4.2.2,2.d (current surveillance) specifies the frequency for measuring the core power distribution. This is done by part b in the _new l surveillance. Part b.3 has been added to this surveillamce, requiring an Fn (X,Y,Z) measurement when the excore quadrant power tilt ratio is normalized using incore detector measurements. This ensures that the impact of any core tilt on F (X,Y,Z) 9 will be determined and reflected'.in the margin calculasions of part c.

9. SR 4.2.2.2.e -has been replaced by SR 4.2.2.2.d in the new surveillance. l The intent of these requirements is similar in that projections of the measurements are made to determine at what point peaking would exceed allowable limits if the current trend continues. In the new surveillance, an incore flux map is obtained and the margin calculations are performed at the time when zero margin is projected. This

     -requirement ensures the core is monitored at a frequency that considers              r conditions when measured peaks are underpredicted.                                  4

10. The caw SR- 4.2.2.2.c replaces 4.2.2.2. f in the current surveillance. The

_ purpose of = part c.1 is to perform margin calculations based- on the measured peaks- b With the new methodology, the limit ([Fg (X,Y,Z)]) to which- the measurement is compared is the design peak at steady-state conditions, increased by a factor that-' represents the maximum amount that the power:at the given assembly location and axial elevation can increase above --the design value before the measured value may become limiting, Margins to both the LOCA peaking limit- (operationa'l ' margin) and the

     -centerline fuel melt limit- (RPS margin) are calculated. The. operational margin forms the. basis for restrict ing the AFD limits in par.t.'c.2, and the RPS margin f rms th2 be t : 'or reducing the OPAT trip setroint -in part c.3.                                                                            t

11. SR 4.2.2.2.c.? (new) replaces SR 4.2.2.2. f.2 in the current surveillance.

    -The reduced AFL limits determined in part c.2 are based on the amount of negative operatic al margin resulting from the margin calculation of part c.1. The parameters NSLOPE 1 and PSLOPE) are the maximum negative and positive AFD reductio'ns required per percent margin change, and are determined from the maneuvering analysis.        These parameters will be given
                                                                                          'J WNP337/9

In the COLR. The AfD must be controlled to these new limits to reduce "g(X,Y,Z), and to 9nsu" that peaking will be limited for continued power operation.

12. SR 4.2.2.2.c.2.b (new) corresponds to SR 4.2.2.2.f.2 b (current surveillance).

13. part 4.2.2.2.c.3 has been added to the surveillance. This part of the surveillance requires reducing the Ki value of the OTAT trip setpoint if the RPS margin is negative. This requirement ensures that centerline fuel melt protection exists when core peaking may be greater than the design values for the specified time in fuel cycle and operational conditions.

14. SR 4.2.2.2.f.2.c, which addresses Base Load operation, has been deleted from the new surveillance. The new power distribution methodology does not recognize this approach and does not constrain core operation to a target AFD.

15. SR 4.2.2.2,g has been replaced by SR 4.2.2.2.e in the new surveillance; there are no substantive changes to this surveillance.

16. SR 4.2.2.3 addresses Base load operation and has been deleted from the new survet116nce.

17. SR 4.2.2.4 addresses surveillance of peaking in Base load operation and has been deleted fro'n the new surveillance.

18. SR 4.2.2.5 has been replaced by SR 4.2,2,3 in the new surveillance; there are no substantive changes to this surveillance.

19. SR 4.2.2.2.1, SR 4.2.2.2.2, and SR 4.2.2.2.3 address F,y monitoring and have been deleted from SR 4.2.2. The new methodology for F(X,Y,Z) g surveillance utilizes peaking limits based on three-dimensional calculations exclusively, and does not- address F g(X,Y,Z) monitoring against planar peakina factors, i

WNP337/10

d Technical JustificatJon: 3/4.2J 9 Specification 3/4.2.2 was revised to provide required actions and surveillance requirements consistent with BWFC methodology for core power distribution control and surveillance of the heat flux bot channel factor, as discusseo in

,           BAW-10163P-A (Reference 1).

The heat flux hot chnnel factor [F g(X,Y,Z)] is a specified acceptable fuel design limit that preserves the initial conditions for the ECCS analysis. F (X,Y,Z) is defined as the maximum local heat flux on _the surface of a fuel 9 rod at a given core einvation (Z) in an assembly located at (X,Y), divided by the average fuel rod heat flux, allowing for manufacturing tolerances on the fuel pellets and fuel rods. Since Fg (X,Y,Z) is a ratio of local surface heat fluxes, it is related to the total local power density in a fuel rod.

          . 0peration within the Fg (X,Y,Z) limits given in the Ccre Operating Limits Report (Cf1R) prevents power peaking that would exceed tne loss of coolant accident (LOCA) peaking limits derived by the ECCS analysis, The F (X,Y,Z) n limit is stated as the product of the peaking ilmit at rated thermal power (Fg RTP) and the normalized peaking limit as a function of core elevation

, [K(Z)]. Separate K(Z) peaking . limits for Mark-BW fuel and 0FA fuel were used in the maneuvering analysis to determine the operating limits. The K(Z) limits for the Mark-BW fuel were determined by the ECCS analysis performed for Mark-BW fuel, as documented in BW-10174 (Reference 2). The K(Z) limits for the OFA fuel are the current Westinghouse K(Z) values for Catawba, as given in the current Catawba COLR, l The reload maneuvering analysis determines limits on global core parameters that reflect the core power distribution. The primary parameters used to monitor and control the core power distribution are control bank insertion, ; axial flux difference (AFD), and quadrant power tilt ratio. Limits are placed on these parameters to ensure the core pnwer peaking factors remain bounded during power operation. Nuclear destgr model calculational uncertainty, manufacturing tolerances (engineering- hot channel factor), effects of ' fuel densification and. rod bow, and modeling simplifications (such as treatment of spacer grid effects) are accommodated through the use of peaking augmentation factors in the mtneuvering analysis. WNP337/11

1 l Measurement of the core power distribution at steady-state conditions by using the incore detectors to obtain a three-dimensional flux map provides confirmation that the measured heat flux hot channel factor F (X,Y,Z) is 9 within the values of the designed core power distribution. This comparison verifies the applicability of the design power level, control bank insertion, AFD, and excore quadrant power tilt ratio to the measured core conditions to preserve the LOCA peaking criteria. References

1. BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed pWRs, June, 1989.

2. BAW-10174, Mark-BW Reload LOCA Analysis for Catawba and McGuire.

31Ljions i to Teghnicaljpecification 3/4.2J This change applies to Unit 1 only. Specification 3/4.2.3 was revised to reflect the power peaking surveillance method described in BAW-10163P-A. These revisions are summarized as follows:

1. The statement of the LCO was revised to reflect new nomenclature for the nuclear enthalpy rise hot channel factor [FN(X,Y)] g and related e parameters required by the methodology of BAW-10163p-A and used throughout the Reload Report.

2. Those requirements of Actions a, b, and e in the current specification relating to the Reactor Coolant System flow rate have been incorporated in Specification 3.2.5. The Actions have been revised to include the N

reduction of allowable thermal power when FAHR (X,Y) exceeds the limit within 2 hours. The factor (RRH), by which the power level is decreased per percent F3g(X,Y) is above the limit, is defined in the COLR. The inverse of this f actor is the fractional increasc in the Maps allowed when thermal power is decreased by 1% RTP, When a power level decrease WNP337/12

is required because FAH(X,Y) has exceeded its limit, then Action b requires restoration of F3q(X,Y) to within its limit, or a reduction in the high flux trip setpoint. The amount of reduction of the high flux trip setpoint is governed by the same factor (RRH) that determines the thermal power level reduction. This maintains core protection and an operability margin at the reduced power level similar to that at rated thermal power.

3. Action b.3 was replaced by Action d. The portions of the Action requirements related to Reactor Coolant System flow rate have been incorporated in Specification 3.2.5, and do not appear in Action d of the new specification.

4. There is no change to SR 4.2.3.1.

5. SR 4.2.3.2 formerly covered only surveillance frequency. It has been expanded as detailed below to reflect the power peaking surveillance method described in BAW-10163P-A and the format of the revised SR 4,2.2.2. Part a addresses obtaining an incore flux map.

6. SR 4.2.3.2.b (new) replaces the current 4.2.3.2 and addresses the frequency for confirming that FAH(X,Y) is within its limit. In addition to performing the surveillance prior to operation above 75't RTP af ter each fuel loading and at least ance per 31 EFPD, the revised surveillance requires measurement of the peaking f actor whenever the excore quadrant power tilt ratio is normalized using incore detector measurements. This ensures that the impact of any core tilt on F3g(X,Y) will be determined and reflected in the margin calculation. This is comparable to the new SR 4.2.2.2.b on Fg(X,Y,Z).

7. SR 4.2.3.2.c has been added. The purpose of part c.1 is to perform margin calculations based on the measured radial peak ratio. The limit l

[ FAHR (X,Y)] to which the measurement is compared is based on the allowable design MAP limit, increased by a factor that represents the maximum amount that the power at the given assembly location can increase above the design value before the measured value may become limiting. WNP337/13

                             -       .. .-       -     .            -        - _      -  - _ _ = - .

Part c.2 uses the amount of margin determined by this procedure to form the basis for the anount of power level reduction and the reduction in the high flux and OTAT K trip setpoints required in the ACTION 3 statements for the specification. This is comparable to the new SR 4.2.2.2.c on F (X,Y,Z). 9

8. SR 4.2.3.2.d has been added. This surveillance requires projections of the measurements to be made to determine at what point FAH(X,Y) would exceed the allowable limit if the current trend continues. In part d.1, M

a penalty is applied to FAHR (X,Y) if the trend indicates that the measured peak would exceed the limiting peak within the 31 EFPD surveillance period, and the margin calculations are repeated. This provides additional margin, or a buf fer, to help ensure that the peak will not exceed the limit prior to next 31 EFPD measurement interval. In part d.2, the measurement is obtained and the margin calculations are repeated so that appropriate actions can be taken before zero margin is reached. This surveillance ensures the core is monitored at a freemney that considers conditions when measured peaks are underpredicted. This is comparable to the new SR 4.2.2.2.d on Fg (X,Y,Z). 9 SR 4. 2. 3. 3, 4. 2. 3. 4, and 4.2.3.5 in the current specification address measurement of Reactor Coolant System flow rate. These requirements have been incorporated in Specification 3.2.5, and have been deleted from the revised requirements for SR 4.2.2. .TA ghnical Ju6tification: 3/4.2.3 ' Specification 3/4.2.3 was revised to provide required actions and surveillance requirements consistent with BWFC methodology for core power distribution control and. surveillance of the nuclear enthalpy rise hot channel factor, as discussed in BAW-10163P-A (Reference 1). The nuclear enthalpy rise hot channel factor [F3g(X,Y)] is a specified acceptable fuel. design limit that preserves the initial conditions for the most- limiting non-0 TAT DNB transient (i.e., primary protection against DNB is . WNP337/14 _ - .-. _ o

a 1 1 1 not provided by the OTAT trip function). F3g(X,Y) is defined as the ratio of the integral of linear power, along the rod with the highest integrated power, I to the value of this integral along to the average rod. Since F3g(X,Y) ) integrates the power along the length of the rod, it is related to the linear heat generation rate of the fuel rod, averaged over the lengt- of the rod. When power distribution measurements from the incore detectors are obtained, the measured value of the assembly radial peaking factor is labeled F X,Y). 6H The FAH(X,Y) limits are preserved by the licensing design analysis, as l described in BAW-10163P-A. Operation within the FAHR (X,Y) limits defined in the COLR ensures that the measured peaking will be within the design L calculations. The FAHR (X,Y) limits are derived from the Maximum Allowable Peaking (MAP) limits specified in the Core Operating Limits Report (COLR). The MAP limits are a family of maximum allowable total peaking curves, typically plotted as maximum allowable peak versus axial location of peak, parameterized by the axial peaking factor. The family of curves is the locus of points for which the minimum DNBR is equivalent to that calculated for the 4 most limiting non-0 TAT transient (based on the reference design peaking). Therefore, the MAPS in the COLR are based on the statepoint that represents the point of minimum DNBR during this transient. 'The MAP limits provide itnkage between the ONBR analyses, with their design peaking distrioutions, and the core operating limits, separate MAP limits are specified in the COLR for Mark-BW fuel and 0FA fuel. The OFA MAPS were derived from the Mark BW MAPS by adjusting the Mark-BW MAPS at all points by the ratio of the design peaks (1.49/1.55). Thermal-hydraulic analyses of mixed cores have verified that this peaking limit adjustment is conservative for OFA fuel. The reload maneuvering analysis determines limits on global core parameters that can be measured directly. The primary parameters used to monitor and control the core power distribution are control bank insertion, axial flux dif ference (AID), and quadrant power tilt ratio. Limits are placed on these parameters to ensure the core power peaking factors remain bounded during power operation. Uncertainties for the nuclear design model, engineering hot channel factor, assembly spacing, axial- peaking factor, and other WNP337/15

uncertainties in the CHF correlation were statistically combined to produce an overall DNBR uncertainty. This overall uncertainty was used to establish the statistical design limit (SDL), as described in BAW-10170P-A (Reference 2). Since the MAP limits link the peaking limits to the DNBR limit, these uncertainties are accounted for in the MAP limits, and are not applied explicitly in the maneuvering analysis. Comparisons of the measured core power distribution at steady-state conditions to the design power distribution provide confirmation that the measured F3g(X,Y) is within the values of the designed core power distribution. This comparison verifies the applicability of the measured core condition to the designed condition, so that if the control bank insertion, AFD, and excore quadrant power tilt ratio are at their most limiting values, then the initial condition DNB peaking criteria are preserved. When the measurement is obtained, values of measured F3g(X,Y) are not compared directly to the MAP limits. Instead, the ratio of the measured FAH(X,Y) to the maximum allowable radial peak derived from the MAP limits is formed. The maximum allowable radial peak ratio is derived from the MAP limits by dividing the MAP for the assembly under surveillance by the measured axial peak for that assembly, and is denoted by FAHRN (X,Y) in the COLR. The corresponding maximum allowable radial peak limit ratio is derived by dividing the MAP for the assembly by the axial peak from the design power distribution database for the particular conditions at which the measurement is made. This ratio is then increased by a factor that represents the maximum amount that the power at the given assembly location can increase above the design value before the measured value may become limiting. The resulting value is denoted by FAHR l (X,Y) in the COLR. If FAHR"(X,Y) is less than FAHRl (X,Y), then a positive margin will exist, and it is inferred that FaH(X,Y) is within the limit. l l WNP337/16

l} . Reference 1-

1. BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed pWRs, June _1989.

2. BAW-10170-A, Statistical Core Design for Mixing Vane Cores. December, 1988.

Revisions to Technical $pecifigtion 3/4.2d . i Specification 3.2.4 was revised to reflect the requirement to decrease thermal 4 power by at- least.3% for each 1% of indicated quadrant power tilt ratio .in excess' of 1.02. This 'was done because the power distribution analysis ; includes a peaking allowance for quadrant power tilt ratios up to 1.02. When the quadrant' power tilt ratio:-increases above 1.02, reductions- in thermal power are required to limit'the maximum local linear heat rate. The actions required to reduce thermal power are provided'in .the current specification. Therefore, this change in = the- specification reflects a quadrant power tilt ' g tratio of 1.02-as the " reference" value,-above which a thermti power reduction '

                  !is required.        This revision is consistent- with _ the peaking allowance for              j
                  . quadrant : power tilt, -as described above.               This revision applies to Unit 1 only.                                                                                          4 4

The Applicab1_11ty of Specification 3.2.4 is for Mode 1 operation above 50% of -

                  ; rated - thermal power. The - phrase "above 50% of rated thermal powe r -was              l removed fromn the !LC0 statement and written .into _the Applicability $tatement              :i forL clarity ~ and . consistency with the format ~f o    the Standard Technical     !

sSpecifications. 'A statement that "The provisions:of 3.0.4 are not applicable"- was'also addedt to clarify lthat the surveillance requirement would be completed above 50% of, RATED THERMAL POWER. This- change is administrative in nature

                 'because.it does not represent an actual change to' the requirements of' 1 Specification 3.2.4: ..or to its required actions.                     The revision to' the     '

Applicability Section applies'.to Unit 1 and Unit 2.

L

                 -.WNP337/17

A footnote was added to the Applicability statement to indicate that the specification is not applicable until completion of excore detector calibration subsequent to refueling. T e e h n I La.L}M c i f i e a.li,p ndvili_f i s a l i o n ;_J &,L4 Specification 3.2.4 was revised to provide required action $ tonsistent with BWFC methodology for core power distribution control as discussed in BAW-10163P-A (Reference 1). During the reload 1teensing analyses for Catawba Unit 1, cycle 6, a three-dimensional maneuvering analysis was performed to determine the core operating limits for power distribution, based on the methodology of BAW-10163P-A. The analysis addressed the Axial Flux Dif ference (AFO) limits (Specification 3.2.1), control bank insertion limits (Specification 3.1.3.6), and Quadrant power Tilt Ratio (Spect fication 3.2.4). The current control bank insertion limits were verified by the analysis, and revised AFD lhtts were set based on the new power distribution methodology and the cycle 6 core design, These results include an allowance for the limiting quadrant power tilt in the core. When the control bank insertion and AFD limits are determined during the maneuvering analysis, the calculated peaking is increased by an amount corresponding to a 2% quadrant power tilt, equivalent to a quadrant power tilt ratio of 1,02, Therefore, the resulting control bank insertion and AFD limits are valid for excore quadrant power tilt rattos up to the Technical

                $pecification value of                                             1.02. Simulations of quadrant power tilts and corresponding peaking increases have shown that a peaking allowance of 3% in the analysis is suf ficient to bound the increased peaking due to tilt up to a quadrant power tilt ratio of 1.02.                                          This peaking allowance has been used in the maneuvertng analysis to determine both AFD (operational) limits and f(AI)

(safety) limits. WNP337/18

Referencel

1. BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PW s, June, 1989.

Erop_ojed TechniqpLSAecification Revision 3/4.2.5 This change applies to Unit 1 only. This proposed Technical Specification (TS) revision changes TS 3.2.5 to include a new figure that defines power reduction requirements for low flow operation. This new figure, Figure 3.2-1, replaces Figure 3.2-3 which had previously been relocated from Specification 3.2.3 to Figure 8 of the COLR. Technical Justification 3/4.2d Since maximum allowable peaking limits are still defined in Specification 3.2.3, the peaking dependence that was previously included on COLR Figure 8 in the parameter R has been eliminated from the new figure. Specification 3.2.5 has also been revised to include the action statements from Specification 3.2.3 that govern the response to the power and flow combination being in the regions of restricted or prohibited operation, while maintaining the current action statement that governs the response to temperature and pressure exceeding their limits. In addition, all flow rate surveillance requirements previously contained in Specification 4.2.3 have been moved to Specification 4.2.5. Although Figure 3.2-3 had previously been moved to the Core Operating Limits Report (COLR), the new figure, Figure 3.2-1, is being returned to the plant Technical Specifications, since the parameters that are governed by the figure (power and flow) do not change on a cycle by cycle basis. Previously, Figure 3.2-3 in Specification 3.2.3 related minimum flow as a function of both power and R, where R was a function of the Nuclear Enthalpy Rise Hot Channel factor, power, and two cycle specific parameters (the design WNP337/19

Nuclear Enthalpy Rise Hot Channel factor and the low power peaking adjustment factor). Because Figure 3.2-3 contained cycle specific parameters, it was moved from Specification 3.2.3 to the COLR. The new figure, Figure 3.2-1, relates minimum flow only to power. The dependence of minimum flow on R has been eliminated. Therefore, since the dependence of the figure on cycle specific parameters no longer exists, it can be returned to the Technical Specifications. Figure 3.2-1 defines the trade-of f in power and flow that will maintain the bases of the core safety and operating limits for a low flow condition. The analysis that verifies this trade-of f considered a flow reduction down to 95 percent of the thermal design flow rate. To verify the validity of the trade-off, several DNBR evaluations were performed. These evaluations demonstrate that the Overtemperature AT safety limits remain valid as flow is reduced and that the specified reduction in operating power level produces improved DNBR margins for the limiting conditions !! event. All reduced power statepoints considered the increase in allowable core peaking at reduced power consistent with Technical Specification 3.2.3. Therefore, no additional limits on the maximum Nuclear Enthalpy Rise Hot Channel Factor are required due to reduced flow conditions, thereby removing the R dependence. To ensure that the -level of protection - that has been assumed in the plant safety analysis 'is maintained, all action statements that were previously included in _ Specification 3.2.3 have been retained in the new Specification 3.2.5. This assures that non-loss-of-flow transients, like the rod withdrawal at power, are protected at the low flow condition. l l l WNP337/20

f F NO SIGNIFICANT !!AZARDS ANALY5!S FOR SAFETY 1,IMITS AND POWER DISTRIBUTION TECHNICAL SPECIFICATIONS The following analysis, required by 10CFR 50.91, concludes that the proposed amendment will not. involve significant hazards consideraticas as defined by 10 CFR 50.92. 10 CFR 50.92 etates that a proposed amendment-involves no significant' hazards considerations if operation in accordance with the-proposed amendment would noti i l

1) Involve a significant-increase in.the probability or consequences of an accident previously evaluated or

2) _ Create the possibility.of a new cr different kind of accident from any previously evaluatedt or

3) . Involve a significant reduction in the margin'of safety.

The fuel for Catawba Nucicar Station Cycles 1 5, for both Units 1 and 2, is Westinghouse supplied. As a result of Duke Power's , decision to open_ future reload contracts to competitive bidding, the ' fuel for at:least Cycles 6-9 of Catawba Unit 1 and 6 and 7 of. Unit ~2

                                        .will:be supplied by B&W Fuel Company. Ur.it_1 Cycle 6 will be the                                l first cycle for:which BWFC supplies the reload. fuel. The Catasba
                                                                                                                                            ~

Unit 1. Cycle 6 Reload. Safety Evaluation Report (Attachment 3) presents an evaluation which demonstrates that the core relcad using

Mark-BW fuel will not adversely impact the safety.of the plant. ,

During Cycle 6 the core will contain 72 fresh fuel assemblies i' supplied!by B&W and 121~ Westinghouse supplied-Optimized Fuel Assemblies (OFA). Methods and models havu been developed to support .

                                      - Catawba Unit 1 operation __during both normal-and off normal .                         .

operation.- These methods and models ensure safe operation with an-entire core of Mark-BW fuel and with a core of mixed Westinghouse and Mark-BW fuel. The analysis methods:aro documented in Topical

                                     - Reports'.which have been submitted to the NRC,'and are either under
                                                          ~

review or approved.= These' Topical. Reports are listed-in Section 10 i L of Attachment 3. i- . i For:the roload-related Technical Specifications tho' probability or L V consequences of an accident previously evaluated-is not significantly increased- - r I l L A.LOCA evaluation for operation of Catawba' Nuclear' Station with. I Mark-BW fuel has been completed-(BAW 10174,

                                     - Analysis for.the(Catawba and McGuire Units);               .       Hark-BW Operation       of Reload the'   LOCA stat on while in transition from Westinghouse supplied 0FA: fuel to i
                                     ; B&W supplied Mark-BW fuel is also-justified in-this topical.

BAW 10174 demonstrates that Catawba Nuclear Station continues to meet the-criteria-of-10 CFR 50.46 when operated with Mark-BW fuel.- i i,

    . ,                  ; ,, _, - , _ , . . . - , - - - ,               - ,---,... - -;a _ _       .--_.._--..,.-,a..-~,,-.          _.-

Large Break LOCA calculations completed consistent with an approved evaluation model (BAW 10168P and revisions) demonstrate compliance with 10 CFR 50.46 for breaks up to and including the double ended severance of the largest primary coolant pipe. The small break LOCA calculations used to license the plant during previous fuel cycles are shown to be bounding with respect to the new the fuel design. This demonstrates that the plant meets 10 CFR 50.46 criteria when the core is loaded with Mark-BW fuel. During thu transition from Westinghouse OFA fuel to Mark-BW fuel both types of fuel assemblics will reside in the core for several fuel cycles. Appendix A to BAW-10174 demonstrates that results presented above apply to the Mark-BW fuel in the transition core, and that insertion of the Mark-BW fuel will not have an adverse impact on the cooling of the Westinghouse fuel assemblica. BAW 10173P, Mark-BW Reload Safety Analysis for the Catawba and McGuire Units, provides evaluations and analyses for non-LOCA transients which are applicable to Catawba. The scope of BAW 10173P includes all events specified by sections 15.1-15.6 of Regulatory Guide 1.70 (Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants) and presented in the Final Safety Analysis Report for Catawba. The analysis and evaluations performed for BAW 10173P confirm that operation for Catawba Nuclear Station for reload cycles with Mark-BW fuel will continue to be within the previously reviewed and licensed safety limits. One of the primary objectives of the Mark-BW replacement fuel is compatibility with the resident Westinghouse fuel assemblies. The description of the Mark-BW fuel design, and the thermal-hydraulics and coro physics performance evaluation demonstrate the similarity between the reload fuel and the resident fuel. The extensive testing and analysis summarized in BAW 10173P shows that the Mark-BW fuel design performs, from the standpoint of neutronics and i thermal-hydraulics, within the bounds and limiting design criteria applied to resident Westinghouse fuel for the Catawba plant safety analysis. Each FSAR accident has been evaluated to determine the effects of Cycle 6 operation and to ensure that the radiological consequences of hypothetical accidents are within applicable regulatory guidelines, and do not adversely affect the health and safety of the public. The design basis LOC /. evaluatior.s assessed the radiological impact of differences between the Mark-BW fuel and Westinghouse OFA fuel fission product core inventories. Also, the dose calculation effects from non-LOCA transients reanalyzed by BWFC utilizing Cycle 6 characteristics were evaluated. Differences in the current FSAR dose values that are not related to the insertion of Mark-BW fuel reflect the application of the latest revisions to Standard Review Plan dose assessment methodology. The calculated radiological consequences are all within specified regulatory guidelines and contain significant levels of margin, j

The analyses contained in the referenced Topical Reports indicate that the existing design criteria will continue to be met. Therefore, these TS changes will not increase the probability or consequences of an accident previously evaluated. As stated in the above discussion, normal operational conditions and all fuel-related transients have been evaluated for the use of Mark BW fuel at Catawba Nuclear Station. Testing and analysis was also completed to ensure that from the standpoint of neutronics and thermal-hydraulics the Mark-BW fuel would perform within the limiting design criteria. Because the Mark-BW fuci performs within the prnviously licensed safety limits, the possibility of a new or different accident from any previously evaluated is not created. The safety analyses performed in support of any reload necessarily involve the assumption of a number of input parameter values. Because of the differences in methodologies between vendors, and the proprietary nature of the analyses, a side by side comparison of input assumptions is generally neither possible nor useful. Reactor Coolant System flow is an exception, because it is a TS constrained and measurabic value. The B&W analyses referenced in the above discussion assumed an RCS flow of 385,000 gpm. TS have been changed to reficct this new value (in Table 2.2-1, footnote, loop minimum measured flow = 96,250 gpm, and Figure 3.2-1 Rated Thermal power vs. Flow). The change also affects Figure 2.1-1, Reactor Core Safety Limits. Because the new safety limits continue to provide assurance that DNB, and hot leg boiling will not occur, this change does not represent a significant decrease in the margin of safety. The reload related changes to the TS do not involve a significant reduction in the margin of safety. The calculations and evaluations documented in BAW 10174 show that Catawba will continue to meet the criteria of 10 CFR 50.46 when operated with Mark-BW fuel. The evaluation of non-LOCA transients documented in BAW-10173p also confirms that Catawba will continue to operate within previously reviewed and licensed safety limits. Because of this, the TS changes to support the use of Mark-BW fuel will not involve a significant reduction in the margin of safety. Several changes have been made to Table 2.2-1. Theso changes reflect updated plant specific instrument uncertainty calculations. The allowable values for both the power range neutron flux high setpoint and the low RCS flow trips are conservatively being made more restrictive as a result of this error calculation. The S and Z terms are used to determine the operability of a channel if the trip setpoint exceeds its allowable value. The modifications to the S and Z values for high pressurizer pressure permit a larger rack drift before the channel must be declared inoperable. The changes to the S and Z values for the high and low neutron flux trips conservatively restrict the rack drift. The changes to Tabic 2.2-1 will not significantly increase the probability or consequences of an accident previously evaluated. The changes to the allowable values for the power range neutron flux 1

(high setpoint) and low RCS flow, and the S and Z values for power range neutron flux (both setpoints), are conservative. The modification to the Z value permits a larger rack drift for pressurizer pressure, low RCS flow, and overtemperature T before the channel becomes inoperable, however these changes more accurately represents expected values, and are within the safety analysis assumptions. Fot similar reasons it can be concluded that these charges will not create the possibility of any new accide,t from those previously evaluated. It can also be concluded that since all new TS values are bounded by safety analysis assumptions that this change will not significantly decrease the margin of safety. Several of the requested amendments are administrative in nature. The requested change which updatas Table 2.2-1 for deletion of the RTD Bypass System, reflects a change which has been previously approved by the NRC (Amendment No. 40 to Facility Operating License NPF-35 and Amendment No. 33 to Facility Operating License NPF-52) . Since the needed modifications have been completed on both Catawba Units 1 and 2 the TSs which no longer apply are beir.g deleted. Since there is no change in requirements this change does not involve significant hazarde considerations. An administrative change has been requested for TS 3.2.4 to delete "above 50% of RATEL TIIERMAL POWER" f rom the LCO, and add it to the Applicability section. A statement that "The provisions of 3.0.4 are not applicable" was also added which would clarify that the surveillance requirement would be completed above 50% RATED TillM1AL POWER. This change is consistent with both the Westinghouse Standard Technical Specifications and the way the plant is currently operated. Since there is no change to the current requirements this change is administrative in nature, and involves no significant hazards considerations. An administrative change is being made to the TS which apply to Unit 2, and no longer apply to Unit i after the reload. The current Figure 2.1-1 has been relabeled as applicable to Unit 2 only. Table 2.2-1 also notes that the the existing Reactor Coolant flow applies only to Unit 2 and the new Reactor Coolant flow (96,250) applies to Unit 1. The Applicability Section of the Power Distribution TS (3/4.2.1, 3/4.2.2, 3/4.2.3, 3/4.2.4) have also been revised to show that the existing TS still apply to Unit 2. The existing TS will be copied on yellow paper to further distinguish them from the new TS which apply to Unit 1 only. The Power Distribution TS will have an "A" in thc page number for Unit 1 and a "B" for Unit 2, the pages will also be marked " Unit 1" or " Unit 2". This change is administrative only, and is being made to distinguish between the TS for Unit 1, which will be operated with TS revisions which reflect the use of Mark-BW fuel, and Unit 2 which will continue to operate with Westinghouse supplied fuel.

                             ~ - - - . _ _ _ _ - - _ _ _ _ _ _

i ENVIRONMENTAL IMPACT STAT! MENT The proposed TS change has been reviewed against the criteria of 10 CFR 51.22(c)(9) for environmental considerations. As shown above, 1- the proposed change does not involve any significant harards consideration nor significantly increase the types or amounts of effluents that may be released offsite, nor significantly increase the: individual or cumulative occupational radiation exposure. Based on this', the proposed Technical Specification change meets the criteria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement. 4, l i I . l f l i l

TECILNICAL JUSTIFICATION AND NO SIGNIFICANT llA7.ARDS ANALYSIS FOR ECCS TS t

l prffosed y Technical __ Specification _ Revision 3/4.5 This change applies to both Unit I and Unit 2. This proposed TS revision deletes LCO 3.5.1.1.1 and the associated surveillance requirements. The note in the Applicability Section of 3.5.1.1.2, (VHI physically disconnected; Cold leg accumulators and discharge paths suitably modified), is deleted. TS 3.5.1.2 and its associated surveillance requirements is also deleted. Technical Justification 3/4,5 This change to the Technical Specifications is administrative in nature. A previously approved change to the TS allows operation of both Catawba Units 1 and 2 with the Upper Head injection System removed. TS 3. 5.1.1.1 and 3. 5.1. 2 along with the associated surveillance requirenents no longer apply since the Upper Head Injection system has been removed. The note in the Applicability Section of 3.5.1.1.2 is no longer needed. The modification of the Upper Head Injection System has been completed on both Catawba Units. The note was included as an interim measure while the modification of the system was in progress. Since VH1 has been removed, deletion of the Specifications, which call for the system to be operable will add clarity to the TS. Changing " pressurizer" to " Reactor Coolant System" in TS 3.5.1.1.2 ACTION c.1, 2, and 3 is a change to reflect the instrument used by the plant to complete the required ACTIONS. Pressurizer pressure goes off scale low at 1700 psig so it can not be used to measure pressure below 1000 psig as stated in tr.e current TS. P_roposed Technical Specification Revision 4.5.2.h This proposed amendment changes the 4.5.2.h l.a flow rate from "333 gpm" to "345 gpm", and 4.5.2.h.2.a flow rate from "462 gpm" to "450 gpm". This change WNP337/21

applies to Unit 1 and Unit 2. There are also administrative changes to remove items in TS 3/4.5.1 and 3.5.1.2 to reflect the fact that VH1 has been physically disconnected, and to change " pressurizer" pressure to " Reactor Coolant System" pressure in Action C.1, C.2, and C.3 to TS 3.5.1.1.2. Technical Justification 4.5.2.h Technical Specifications 4.5.2.h.1.a and 4.5.2.h.2.a are flow requirements for the centrifugal charging pumps and safety injection pumps, respectively. Each flow value represents the minimum acceptable total injected flow through three injection lines with the Reactor Coolant System at atmospheric pressure. The flow through the fourth injection line is assumed to spill out of a pressure boundary rupture during a LOCA and is not available for core cooling. The 333 gpm value must be increased to 345 gpm in order to match the flow assumed in the new FSAR Chapter 15.6.5 LOCA analysis associated with the Catawba 1 Cycle 6 reload. The flow assumed in the existing LOCA analysis is 308 gpm. The 345 gpm value is only a small increase in the previous value and is not associated with a change in the centrifugal charging pumps or piping system. The present configuration of the system meets the 345 gpm flow requirement. 345 gpm was assumed in the new LOCA analysis since the analysis is valid for both Catawba and McGuire, and the existing McGuire Technical Specification value is 345 gpm. The proposed change from 462 to 450 qpm for the safety injection pumps is nuessary to provide margin for instrument string uncertainty during flow testing, and for a reasonable telerance on the test acceptance criteria for injected flow imbalance between the four injection lines. The flow assumed in the new LOCA analysis associated with the Catawba 1 Cycle 6 reload is 405 gpm. The flow assumed in the existir ,CA analysis is 462 gpm. The impact of decreasing the ficw assumed in the existing LOCA analysis from 462 gpm to 450 gpm has been determined to be acceptable based on the excess flow available from the centrifugal charging pumps (308 gpm assumed and 345 gpm available). The flows delivered by the ECCS sub-systems can be traded of f in this manner as long as the total ECCS flow assumed in the LOCA analysis is not invalidated. The proposed Technical Specifications are consistent with both the existing LOCA analysis, and the new LOCA analysis associated with the Catawba 1 Cycle 6 reload, and are therefore acceptable. WNP337/22

The following changes requested are administrative in nature. The deletion of Specifications in Section 3/4.5.1 which require the UHI System to be operable in the applicability, and 3.5.1.2 which is marked in the T$ to be deleted when VH1 is physically disconnected from the Reactor Coolant Syttem, also reflects a change which was previously approved by the NRC ( Amendent No. 32 to Facility Operating License NPF-32 and No. 23 to Facility Operating License NPFa$2. Since the needed modifications have been completed on both Catawba Units 1 and 2 the TSs which no longer apply ate being deleted. Another administrative change is changing " pressurizer pressure" to " Reactor Coolant System" pressure in ACTION C.1. , C.2, and C.? to TS 3.5.1.1.2. This change is administrative because it reflects the instrument used by the plant to complete the required ACTIONS. Since pressurizer pressure goes of f scale low at 1700 psig, it cannot be used to measure pressure below 1000 psig as stated in the current TS. Since there is no change in requirements, this change does not involve significant hazards considerations. I WNP337/23

NO SIGNIFICANT HA7.ARDS ANALYSIS TS 3/4.5 FNERGENCY CORE COOLING SYSTFMS The proposed changes to TS 3/4.5 (Emergency Core Cooling Systems) consist of an administrative change to remove TS related to the Upper Head Injection System and revisions to the required flowrates for the centrifugal charging pumps (TS 4.5.2.h.l.a. "333 gpm to "345 gpm") and safety injection pumps (TS 4.3.2.h.2.a "462 gpm" to "450 gpm"). The following analysis required by 10CFR 50.91 concludes that the proposed amendment does not involve a significant hazards consideration as defined by 10CFR 50.92. 10 CFR 50.92 states that a proposed ame.ndment involves no significan' hazards considerations if operation in accordance with the proposed amendment would not:

1) Involve a significant increase in the probability or consequences <

an accident previously evaluatedt or

2) Create the possibility of a new or difforent kind of accident from any previously evaluatedt or k s

3) Involve a significant reduction in the inargin of safety, i The proposed rovisions to the required flowrates for the centrifugal charging pumps and safety inject. ion pumps do not involve a significant.

Increase in the probability or consequences of an accident previously evaluated. The new pump ilowratos represent a change in the assumptions made in the LOCA analysis, not a physical change in the plant. The increase in the required centrifugal charging pump flows is small, and i both Catawba 'Jnits currently meet tho new requirement. The required ; flowrate for the safety injection pumps has been lowered to allow for instrument uncertainty and to allow for a reasonabic tolerance in the acceptance criteria for injected flow imbalance between the four injection lines. Lowering the safety injection pump flows will be accepta510 with respect to the existing LOCA analysis, because as discussed above there is excess flow available from the centrifugal charging pumps. Since a total ECCS flow value is assumed in the LOCA analysis, lowering the required flow for the safety injection pumps is acceptable for the current LOCA analysis, as long as the total ECCS flow assumed in the LOCA analysis remains availabic. Because the new TS requirement.s are consistent with both the now and the existing LOCA analyses, neither the probability or the consequences of an accident previously evaluated will be significantly increased. The proposed Technical Specifications meet the criteria of both the new and the existing LOCA analysis. No changes have been made in the plant, and both Catawba units are currently operating within the proposed TS. Since a total ECCS flow is assumed in the LOCA analysis increasing the required centrifugal charging pump flow to account for a decrease in I

j required safety injection flow insures that the existing LOCA analysis remains valid with the new TS requirements. Because the new TS values ensure that both the now and existing LOCA analysis remain valid, this change vill not create the possibility of a new or different accident from any previously evaluated. s The LOCA analysis assumes a minimum ECCS flow. Both the new and the existing LOCA analyses remain valid with the proposed TS changes. Because the LOCA analysis remains valid, this change will not involve a significant reduction in the margin of safety. The following changes are administrative in nature. The deletion of Specifications in Section 3/4.5.1 which require the UH1 System to be . - operable in the applicability, and 3.5.1.2 which is marked in the TS to

                                - be deleted when UHI is physically disconnected from the Reactor Coolant System, also reflects a change which was previously approved by the NRC (Amendment No. 32 to Facility Operating License NPF-32 and No. 23 to Facility Operat'.ng License NPF 52). Since the needed modifications have been completed on both Catawba Units 1 and 2 the TSs which no longer apply are being deleted. Another administrative change is changing
                                -" pressurizer" pressure to " Reactor Coolant System" pressure in ACTION C.1, C.2, and C.3 to TS 3.5.1.1.2. This change is administrative because it reflects the instrument used by the plant to complete the required ACTIONS. Since Pressurizer pressure goes off scale low at 1700 psig, it cannot be used to measure pressure below 1000 psig as stated in the current TS. Since there is no change in requirements, this change does not involve significant hazards considerations.

F.NVIRONMENTAL IMPACT STATFE NT The proposed TS change has beenireviewed-against the criteria of 10 CFR 51.22(c)(9) for environmental considerations. As shown above. the proposed change does not involve any significant hazards consideration, nor significantly increase the types or amounts of effluents that may be released offsite, nor significantly increase the individual or cumulative occupational radiation' exposure. Based on this, the proposed Technical Specification change moots the critoria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement.. i i y - g---, ci e w ,+.y-..g sww-w - --+.- g.- c.-v,. , . -gew'- #g, y --vy - ,r ,i- -i.v- w 3--yg-yyw- - - "gye+-, y

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                                                  . NO SIGNIFICANT llAZARDS ANALYSIS FOR TS 6.9.1.9

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NO SIGNIFICANT HAZARDS ANALYSIS ADMINISTRATIVE CIUsNCES TO TS 6.9.1.9 There has been an administrative change proposed to TS 6.9.1.9 (Core Operating Limits Report) to reflect the use of BWFC methodolo3y and analyses. Note "*", "**", and "****" on Attachment 1 for Specification 6.9.1.9 are added to clarify that only Referencee 1, 2, and 3, which reference Westinghouse methodology and analysis apply. **" clarifies that Westinghouse analyses does not apply to IAHR Notg"*The references added as Attachment 2 to TS 6.9.1.9 (BAW-10152-A " NOODLE-A multi-dimensional Two Group Simulator," BAW-10163P-A, " Core Operating Limit Methodology for Westinghouse Designed PWR's", and BAW-10168P, Rev. I "l%W Loss of-Coolant Accident Evaluation Model for Recirculating Stearn Cencrator Plants") reflect the use of BWFC methodology to determine the cycle specific limits in the COLR. These references reflect methodology which is approved or under review by the NRC staff. 10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated or

2) Create the possibility of a new or different kind of accident from any previously evaluatedt or

3) Involve a significant reduction in the margin of safety.

This change is administrative in nature, reflecting the use of NRC approved methodology to determine the operating limits in the COLR. The use of BWFC methodology and analysis has been previously justified in this submittal. The administrative change to the references for the COLR will not increase the probability or consequences of an accident previously evaluated. For the above reasons it can also be concluded that this change will not create the possibility of a new or different accident from any previously evaluated. The use of BWFC methodology and analyses have been previously determined to be acceptabic by the NRC, and their use to determine the operating limits for Catawba Unit 1. Cycle 6, is previously justified in this submittal. Because this change is administrative, simply listing the methodologies already determined to be acceptable, it does not involve a significant reduction in the margin of safety.

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DAW 2119 October 1990 RELOAD REPORT Catawba Unit 1, Cycle 6

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BAW 2119 ! October 1990 i i i s RELOAD REPORT [ Catawba Unit 1, Cycle 6 j.w .- p y ng p e L gy.e ( y[7,.Q,sA..y.w. .

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