ML20058J348

Proposed Tech Specs to Reduce Required Min Measured RCS Flow for Unit 1 Only from 385,000 Gpm to 382,000 Gpm

ML20058J348
Person / Time
Site:	Catawba
Issue date:	12/06/1993
From:	DUKE POWER CO.
To:
Shared Package
ML20058J345	List: ML20058J342 ML20058J348
References
NUDOCS 9312140071
Download: ML20058J348 (21)

Text

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Attachment I 4 Marked-up Technical Specification Pages i

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I 9312140071 931206 IIT  !

PDR ADOCK 05000413 L3 'l P PDR V-a \

F s .

l RE PLRCE .

TOTAL F1.OW - 385000 GPM '

t 660 : _

i 655 2 l 2 2455 psia '

UNACCEPT* LE 2

650 ;

OPERA CN I f

645 -  !

2400 psia p

s4o 2

635 -

2280 psia i a

e 630 2 I 1

~

c- I r

o 2100 ps g 620 5  :

" d J

S15 : i 4

3 .345 La f 610 -

2 4_

605 -

-t 600 2

595 3 -

4 3 ACCEF ABLE 590 --

ope T10N 2 >

SSS _

]

SSO , .

0 0.2 0.4 0.6 0.8 1 1.2 Fracuan of Rated Thermal Power F! CURE 2.1-1 i REACTOR CORE SAFETY LIMITS - FOUR LOOPS lti OPERATI0ff I. ,

CATAWBA - UillTS 1 & 2 2-2 Amendment tio. 107 (Unit 1) ,

Amendment tio. 101 (Unit 2)

Yhll$ $ ONl Y 670 DO NOT OPERATE IN THIS AREA

~

660 -

• 650 --

2455 psia 640 2400 pslo b630 -

2280 pslo

?

3 0 620 -

2100 psio 610 -

1945 pslo 600 -

590 -

ACCEPTABLE OPERATION 580 O.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1-1a REACTOR CORE SAFETY LIMITS - FOUR LOOPS IN OPERATION 382,000 Opm CATAWB A - UNIT 1 2 - A2

VsWf 2. Ot)LV :

670 '

DO NOT OPERATE IN THIS AREA ,

x . . l l

gg _

2455 pslo  !

640 - 2400 pslo .

i e

t,630 -

2280 psia  !

0)  !

W l v>

h620 2100 psic  !

I i

610 -

1945 psia l

UJO -

590 -

i ACCEPTABLE OPERATION  ;

580 l O0 0.2 0.4 0.6 0.8 1.0 1.2 i Fraction of Roted Thermal Power  !

Figure 2.1-1b i REACTOR CORE SAFETY LIMITS - FOUR LOOPS IN OPERATION '

l 385,000 gpm l CATAWB A - UNIT 2 2-D2

v y v 3 TABLE 2.2.-l f ,._

9 REACTOR TRIP SYSTEM _ INSTRUMENTATION TRIP SETPOINTS 6

> FUNCTIONAL UtilT TRIP SETPOINT ALLOWABLE VALUE i

E 1. Manual Reactor Trip ,

N.A. N.A.

2. Power Range, fleutron Flux d'

a. High Setpoint s109% of RTP* 5110.9% of RTP*

7

b. Lcw Setpcint $25% of RTP* $27.1% of RTP*

3. Power Range, Neutron Flux, $5% of RTP* with a s6.3% of RTP* with High Positive Rate time constant a time constant .

-2 2 seconds 2 2 seconds

4. Intermediate Range, Neutron Flux s25% of RTP* s31% of RTP*

r0 5 fg 5. Source Range, Heutron Flux s105 cps sl.4 x 10 cps

'6. Overtemperature AT See Note 1 See Note 2

7. Overpower AT See Note 3 See Note 4

8. Pressurizer Pressure-Low.

21945 psig 21938 psig***

b j [j 9. Pressurizer Pressure-High $2385 psig $2399 psig 3.E ll$" 10. Pressurizer Water Level-High 592% of instrument span s93.8% of instrument span

?

Y 11. ' Reactor. Coolant flow-Low 290% of loop minimum 288.9% of loop minimum 2E F measured flow ** - measured flow **

$55

~- *RTP - RATED THERMAL POWER 9fsoo F SF " Loop. minimum measured flow -4frr260 gpm' _

'j 72; - *** Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for. lead g;; and 1 second for lag. Channel. calibration shall ensure that these time constants are adjusted to

-these values.

v N ih 2 onl -

TABLE 2.2.-l '

Q I E REACTOR TRIP SYSTEM IfiSTRUMEtiTATI0ti TRIP SETPOIffTS -

f FUNCTIONAL UtilT TRIP SETPolliT ALLOWABLE VALUE 5 1. Manual Reactor Trip 3 ' N.A. ff. A.

} 2. Power Range, ticutron Flux

{ a. High Setpoint

$109% of RTP* s;10.9% of RTP*

b. Low Setpoint

$25% of RTP*

s27.1% of RTP*

3. Power Range, fleutron Flux, High Positive Rate s5% of RTP* with a s6.3% of RTP* with time constant a time constant 2 2 seconds 2 2 seconds

4. Intermediate Range, fleutron Flux s25% of RTP* 531% of RTP* '

? '5. Source Range, tieutron Flux eg s105 cps $1.4 x 105 cps 4 6. Overtemperature AT See flote 1 See Note 2

7. Overpower AT See liote 3 See flote 4

8. Pressurizer Pressure-Low

>b 21945 psig 21938 psig*** '

{g 9. Pressurizer Pressure-High s2385 psig

c. t $2399 psig i ;! 10. Pressurizer Water level-High 592% of instrument span s93.8% of instrument span

d. 11. Reactor Coolant Flcw-Low 290% of loop minimum

.} } measured flow **

288.9% of loop minimum MS measured flow **

~~

f" *RTP = RATED THERMAL POWER

? ** Loop minimum measured flow - 96,250 gpm h ***andTime constants I second for lag. utilized in the lead-lag controller for Pressurizer Pressure-Low are 2l seconds for lead these values. Channel calibration shall ensure that these time constants are adjusted to l

TABLE 2.2-1 (Continued) (1,44 } o d y . _

TABLE NOTATIONS -

n h NOTE 1: OVERTEMPERATURE AT c

5? 1 + rd) ( 1 ) s AT, (X, - K2 (0 + rJ) [T(I I - T'] + K3 (P - P') - f (al)) 3

, AT ((1 5) + r2 (1 + bS) 1 + r sS) 1 + r eS)

. t c

35 Where: AT - Measured AT by Loop Narrow Range RTDs; N

i ' - 1 + rd = lead-lag compensator en measured AT; ,

@ l+r52 t3 Time constants utilized in lead-lag compensator for AT, r3 = 12 s, r, , r2 =

r2 - 3 s; I - Lag compensator on measured AT; 1 + r35 r3

- Time constants utilized in the lag compensator for AT,3 r - 0;  ;

AT, - Indicated AT at RATED THERMAL POWER;

{Yb  ;

4 Ki - -1.190 1.19 F4 4

K, =

A03163/*F o. oM71 1 + r,S

= The function generated by the lead-lag compensator for T,,,

l+rSs dynamic compensation;.

j. {

.. 3 ff a

r. , rs = Time constants utilized in the lead-lag compensator for T,,, r, = 22 s, rs - 4 s;-

II Average temperature, 'F; ef T - . . ,

l 1 - Lag compensator on measured T,,,;

n

• r, = Time constant utilized in the measured T,,, lag compensator, r, = 0; .

3:.

. enummmme o* W e ur'p w tow a tr eeT+'wre-ww e'rww w y-wet we een w +- e w e ese%- N'w- w er e e e we, mres'wwr*,r- 'P9' e vW"6-**1P-W-w'*-at- t' w-w--e t"Mwe- rTT %vr'&W n-*-WP^STt eM' * ' * *W-"4-W-^-M W**er W

• we evnwe e w e'w NW*1r'v-i-tm#W -u e w = w'-wrvre e +,w.w we in wh-r ,, ww gyin in, w p, - we ,meav.,,y,w,we w .*e

N .

TABLE 2.2-1 (Continued) desh 2. on l TABLE NOTATIONS ,.

n h tiOTE 1: OVERTEMPERATURE AT c

-5 1 + riS) ( 1 )

s AT, (X, - X 4

2 ((1 + r S)1 + 7 5) [T(I I ) - T') + K3 (P - P') - f, (AI))

, . oT ((1 + r 5) (1 3 + r 5) s 1 + r.5) c.

fi Where: AT = Measured AT by Loop Narrow Range RTDs; 4

L 1 + ril - Lead-lag compensator on measured AT;

$ 1 + r2S N

r3 , r2 - Time constants utilized in lead-lag compensator for AT, ri = 12 s, r2 - 3 s; l I = Lag compensator on measured AT;'

1 + r3S r3 = Time constants utilized in the lag compensator for AT, r3 - 0;

• Indicated AT at-RATED THERMAL POWER;

}T

-Q, AT, =

M X, = 1.1953 .

K2 =

0.03163/'F 1 ' + r, S - The function generated by the lead-lag compensator for T,,,

1+rSs dynamic compensation;

?e n o 4

y' j[ r. , rs = -Time constants utilized in the lead-lag compensator for T,,,, r - 22 s,

.a 3 rs - 4 s; &

4 1.!.' = Average temperature, *F; j

,c '[ T -

I Lag compensator'on measured T,,,; .i N5! -

1- 1+rSe iC:

Time constant utilized in the measured T,,, lag compensator, sr = 0;

[.$ r, =

n..-

me,s m

esuseno r-~ , - .-,-..-,..m->. .4-.-. .4v, -s -,---~,...--o-..-+%,mm-.--. --......--,,4,<,--,#-,v4-..,-.~.,....-~--,..-,..w-...,~e,.,.. --.,---.....4.- .-..,,.--.~~~...~~.,,.-.-..--r,4

TABLE 2.2-1 (Continued)

~ TABLE NOTATIONS (Continued) bal4 '6 o nIy J l *'

n B tio E i: (Continued)

T' s 590.8'F (flominal T,y allowed by Safety Analysis);

g K3 - -0.001111; o. col 629 P = Pressurizer pressure, psig;

-- P' = 2235 psig (flominal RCS operating pressure);

r, Laplace transform operator, s";

S =

and f (o!) is a function of the indicated difference between top and bottom detectors of the power-3 ,

range neutron loi, chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that: , '

-42.o + 8.o (i) For q,- q3 between 30.3 and +hW.,

• N. f,(AI) = 0, where q, and g3 are percent RATED THERMAL POWER in the top and bottom halves of b the core respectively, and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER:

M

- 42.0 (ii) For each percent Al that the magnitude of q, - q, is more negative than -394';, the AT Trip Setpoint shall be automatically reduced by a-9ff/. of AT,; and

3. fe72. 9 9,o

L > (iii) For each. percent AI that the magnitude of q, - q, is more positive than +he%, the oT Trip Setpoint shall be automatically reduced by 2:3M% of AT,.

ll f.(A o l ag

!9 5- NOTE 2: The. channel's maximum Tri) Setpoint shall not exceed its computed Trip Setpoint by more pg than 34% 4. 5 */, gc h M c M R er G,$

[

B. ,

f

~_ v [

TABLE 2.2-1 (Continuedl l TABLE NOTATIONS (Continuedi (_, laid 2orh}. .-

3 NoiE 1: (Continued)

T' s ' 590.8'F (fiominal T., allowed by Safety Analysis);

5 g K3 - 0.001414; i

$ P = Pressurizer pressure, psig; n . P' - 2235 psig (fiominal RCS operating pressure);

~

S - Laplace transform operator, s"; ]

and f,(o!) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTVP tests such that: ,

(i) . For q, - q, between -39.9% and +3.0%,  !

f,(AI) = 0, where q, and q, are percent RATED TiiERMAL POWER in the top and bottom halves of G the core respectively, and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER; ,

a c4 For each percent AI that the magnitude of q, - q, is more negative than -39.9%, the oT Trip (ii)

Setpoint shall be autcmatically. reduced by 3.910% of AT,; and yp (iii)

For each. percent AI that the magnitude of q, - q, is more positive than +3.0%, t.he AT Trip gg Setpoint shall be automatically reduced by 2.316% of AT .

EE EEP The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more NOTE 2:

gg than -hek 45%gQ g%(p%.

J -m as

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~=

iw T wtC4w- %ewp*-N W W e e 91w-MM- P wimme#rb 7'W'-IW'riwe'1-W-WW--%W'*W- Wewr"W M-**v'** Wer Mvw -9wik eir w *w-hW*Y' M'4"m'-wh-TM*Fmmh='9'*et'--@e- Wu*twv'v-'Nwww er im- M f+ tMiiww'"e TNu#ime*-* &se-w,-u'e**w arv - 1 TTD'w'-N et N wM'*w w My'1=ww+ww w w-=new W7'Ta-& e wvvd wew -s'r'vvw F'TN'W-%'y

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TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

Un,4 Iod '

l g

NOTE 3: OVERPOWER AT c

$ I I 1 )

AT 1(II+ r,5)

IItE)( 1 +I r3S) s AT, (K. - K3 (I-_r,S_

1 + r,5) _) (1 +( r 5) )z,y ( 7 ((1 + r,5) } - I"] - fdal))

E T Where: AT - As defined in Note 1, A 1 + riS) -

As defined in Note 1

" l + r,S ,

ri , r, - As defined in Note 1, 1 - As defined in Note 1, I+r53 r3

- As defined in Note 1,

[ _o AT, - NsdefinedinNote1, h -

4-0&& l.0969 K5 -

0.02/*F for increasing average temperature aqd 0 for decreasing average temperature, ,

h _r,S_ -

The function generated by the rate-lag controller for T , dynamic

@g 3 1 + r,S compensation,

o. a

$; R2=R r, - Time constant utilized in the rate-lag controller for T,, r, - 10 s, m

" [7

" I

@.k' As defined in Hote 1, 1+rS ..

w c o

].ee

r. - As defined in Note 1,

3. 3.

~ +

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n n n g T g n n n n i n -

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i i n i i

f a ca n i n g f f i f ns o i

f f 9

• r f i e e f e e e 1

/e un c e f d d e d d a 8 2 p f e d e d p e d ,

) s s s s s 0 0 m e em w s

) A A s A A 'A 1 0t h o '

A s ,

5 3

A T c i A -

r - -

1

+ - - - -

1 )

(( S S, 5 T ri r r 3

S.

A )) 55, + +

r, I L,S, r 1

r R r E

r T A

i

+

r3 T

A hg r_ + + .

W O

+ +

1 l r I K K _.

1 r

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1 R ( ( r -

E e .

V T h O A W 3 .

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7 t" C 9sc5 , E. C

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"i'c w .

Usk \ oJy - -

1 TABLE 2.2-1 (Continued) 1 - -

i TABLE NOTATIONS (Continued) n .

$ liOTE 3: (Continued) -

$5 o co l2 62.

K3 - 0,001291/*F for T > 590.8'F and K 3 = 0 for T s 590.8'F, 7 ,

T - As defined in. Note 1, y~

• T" Indicated T a l ~

fnstrumentaIYon,t f 590.8'F), RATED THERMAL POWER (Calibration temperature for 6 ,

' " S - As defined in Note 1, i

and f2 (AI) is a function of the indicated differences between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured .

instrument response during plant startup tests such .that:

(i) for POWEk q - q3 between -35% and +35% AI; f2 (AI) = 0, where q and q3 are percent is total RATED TilERMAL in the tcp and bottom halves of the core respective,ly, and q, + q3 M THERMAL POWER in percent of RATED THERMAL POWER; f J # ,

.2 -q is more negative than -35% Al, 1

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more 12 fiOTE 4: iR than 2-6H 35 3.o /o Eded Thered Power SE if .if 122 bh n~ Y

• I e-_ _ - _ - -_._ea -

ie-a * ---w,= _ereve we i a- v e a- % **a-r-o 7-rw a a +-e e'e =s t ews3-e-w at e-e' -*in-swat =--=-i-+ +ee wwww =&ew ww'te- ep s = - ar *- w en--em-a se-~%--- e - u t+ e '=*-s- ++ue--"4W -et--s te t

• et-r---% v m-=,4<---<:-ec. e- e-

ma s- . ~ ~ Un; !r Z o Iy - . I TABLE 2.2-1 (Continued) TABLE NOTATIONS (Continued) f n j NOTE 3: (Continued) l cz K3 - 0.001291/*F for T > 590.8'F and 3K - O for T s 590.8'F, 7 E T - As defined in. Note I,

• T" - Indicated T a

- E instrumentaf. Yon,t s 590.8'F), RATED THERMAL POWER (Calibration temperature for AT ' o .S - As defined in Note 1, and fg (AI) is a function of the indicated differences between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured - . instrument response during plant startup tests such that: (i) for q - q between -35% and +35% AI; f3 (AI) - 0, where qt and 9 are 3 percent RATED 4 lHERMAL POWEk in The top and bottom halves of the core respectively, and q, + q3 is total mp THERMAL-POWER in percent of RATED THERMAL POWER; is more negative than -35% AI, (ii) for each percent AI that the magnitude of q - the AT Trip Setpoint shall be automatically reduced,by [.0% of AT,; and (iii) - for each percent AI thf.t .aagnitude of q - q is more positive than +35% AI, the 1 AT Trip Setpoint shall be automatically reduced by 7.h% of AT,. , M3 NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more EE than -2,8k

' 3 37,,d ROcc\ The.cmd %cr E. E.

22

5. 3.

r c- ______.._____________m_.. . , _ , _ _ . _ _ ___m_ ___.. _ , - - _m - _. e ,, ,6-., ,%, - 3--..m-a ,_%-..,, n., .,%, ,,,_,.%-, ., -.-wy , ,_ ,,-%,.,w., , , .._.,_.y,_,, , ,vg., gm%#, ,..,,,,,,_.,,,.,_,,3.,,,_,g, _ , , , . , _ . . , , t . . i i Un;4 I or If - i 3 G 5, k? o ( ' Gue6% a s .n ut y a t o. s t o, ,, o. .e . g ' e x - e e u . .ntu,, e xan; ina . Permissiblo e a su om.ni .re nianty 2: 2.- 5. tc' CCeration 's u . nc e n :n e ng u .. Regscn - 3 & 2, c00 3G2000 22 m _................................................. 'sa.a secci [ I - i i _ m, - Restricted Operation 3%, Iso (8 8 M"50) E 24uso-. , 5 Regson o  ! g - . t ~ 37-4,360 3: 3'N o p g, b - , (94.ml i u. E - o is - f.n - 3')o,5vo MO,W ' ye SHM i (92..W$tto) (. o , e ' O o Frchibited 4 ~ 34,72o Op erat. ion a Mr720 (so.:tteoc) Region I 3 M9tCO-t I }  ! _%2,creo i 00575-0 ~ 3 Y'so Eol - 464900 I 85 85 90 92 94 96 95 10 0 10 2 l i Fraction of Rated Thermal Power  ! figure' 3.2-1 Reactor Coolant System Total Flow Rate Versus

Rated Thermal Power - Four Loops in Operation l A 2.-l b l i.ATAWBA - UNITf 1 -fr-E- 3/4 2- M h admont t{nJ W44 -t)  !

Amendmettt ::c. t J (Ur ; t +)._. l  ! T ,*.,- 2 01 $ f /- t 388850- . 4 3,o u17 o f 0_ M t er un oe t ect ec _ <e ,a w a t e< v entur1 r x >,g soc a f~ 9 N 3 3 d.0 s a su e wt u c e< t anty o f 2.*% f or Co ar aliOD i

o. at e muoes n uus rig.s e.

R 891C0 , 2 8 5 0 0 0 - - -- -- ---- ~ --~ - ~ -----" ~ ~ --- ~ ~ ~- ~ -"- ~ ~ ~ ~-- --"------- 8.3 8 5 0 00 ) , Restricted Operation (S838

• 1 E 281150 Regica ,

.55 . C ' a . , = . 3 ;94.3773001 a 377300- ,

u.  ;

E o g . GO . (92.373450) 1 $ 373450- ( , o

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m Prohibited . Operation o c !so.359600) Region 3 365600- , m 1 385750 , s ~ f l l 381900 56 85 90 90 94 96 95 10 0 10 2 r Fraction of Rated Thermal Power j Figure 3.2-1 Reactor Coolant System T6tal Flow Rate Versus k , Rated Thermal Power - Four Loops in Operation l CATAWBA - UNIT [-1 2 3/482-16 -Amendmertt--No . 10 7 (Un 44.).-- l , Amefrdment- 14sr LO1 (lin ' t 7 )- Attachment II  ! Justification and Safety Analysis  ! l Over time, degraded steam generator tubes have been plugged or sleeved, resulting in a reduction of reactor coolant system flow. , In addition to this, the hot leg streaming phenomenon affects the - accurate measurement of flow. As a result of these effects, it~  ! will become difficult to ensure meeting the minimum flow requirement (Table 2.2-1 Item 11, as annotated) required by , Technical Specifications to maintain 100% power operation. The proposed reduction in minimum measured flow is applicable to Catawba Unit 1 only. To alleviate this concern, analyses have been performed to justify reduction in the minimum RCS flow to 382,000 gpm. These analyses show that the reduced flow rate will not_have a significant impact on any accident analyses presented in Chapters  ! i 3, 4, 6, or 15 of the Final Safety Analysis Report (FSAR). The overtemperature delta T (OTAT) and overpower delta T (OPAT) setpoint equation constants have been revised to support the reduction in minimum measured flow. The methodology used to generate the constants is described in the April 26, 1993' letter from T. C. McMeckin, Duke Power Company, to USNRC Document Control Desk, Supplement to Technical Specification Amendment  ! Relocation of Cycle-Specific Limits to the Core Operating Limits Report. The proposed revision to the OTAT and OPAT constants is applicable to Catawba Unit 1. The change is not applicable to , Catawba Unit 2, because the steam generators in Unit 2 have not degraded and have not required tube plugging or sleeving to the , extent of the other three units. This is consistent with Duke's I current plans to replace the steam generators in both McGuire l l units, and Unit 1 only at Catawba. The higher minimum flow in  ;

Unit 2 is being retained to provide increased flexibility in fuel  !

cycle design work.  ! l The changes to the OPAT setpoints for Catawba Unit 1 also l necessitated recalculation of the Technical Specification - allowable values of the trip functions. The revised OPAT , allowable values are more restrictive than the existing values. In the course of these calculations, a minor error was discovered-that affected the existing allowable values for all four units. This resulted in a recalculation of the allowable value for , Catawba Unit 2, as well as the three units affected by the flow reduction. Since the setpoint is, by administrative controls, I reset whenever it is found to be different from the correct value  ; by about 1%, past operability is not considered to be a concern. This item is discussed in more detail in Duke's response to a ] request for additional information (Reference letter, M. S. l Tuckman to Document Control Desk, December 3, 1993). Also, to I improve clarity, the allowable values of OPAT and OTAT are now l expressed in units of % Rated Thermal Power. l 4 - - ,a , i l Effect of Reduced Flow on FSAR Analyses LOCA Blowdown Forces, FSAR Chapter 3 The primary factors which affect the blowdown forces resulting l from a LOCA are RCS pressure, vessel inlet and outlet fluid  ! temperatures, and to a smaller degree, the loop and vessel  ! flowrates. The LOCA analyses have been performed with a flow which corresponds to a minimum measured flow (MMF) less than 4 382000 gpm, and therefore a reduction in MMF to 382000 gpm will not affect the assumptions in the blowdown forces analysis. i Thermal Hydraulic Design, FSAR Section 4.4 l The thermal hydraulic design for Catawba Unit 1 was analyzed with the reduction in RCS minimum measured flow (MMF) to 382,000 gpm. The reduced flow rate resulted in a slight reduction of the margin in the core DNB limits. Technical Specification Figure ' 3.2-1, Reactor Coolant System Total Flow Rate Versus Rated - Thermal Power - Four Loops In Operation, was revised to reflect , the lower allowable ficw rate. The Axial Flux Difference limits, i Technical Specification Section 3.2.1, are unchanged and all of the current thermal hydraulic design criteria are satisfied at i the reduced flow conditions. As previously noted, re. vised core thermal limits were generated to reflect the reduced minimum measured RCS flow of 382,000 gpm. Based on these new protection limits, the overtemperature delta T (OTAT) setpoint equation constants (Note 1 of Table 2.2-1), and . the overpower delta T (OPAT) setpoint equation constants (Note 3 of Table 2.2-1 for Catawba) were revised to reflect the necessary changes. The impact of the reduced flow on the coefficients was partially offset by a reduction in the margin assumed in the calculation of the coefficients. Mass and Energy Releases for Containment Analyses, FSAR Chapter 6 l l The reduction in MMF flow can affect the mass and energy releases t for containment analysis only through a change in the NC system l temperature input assumption. RCS average temperature will y remain unchanged with the change in MMF. Therefore, the RCS initial fluid and metal stored energy will remain unchanged. Further, a constant PCS average temperature implies that the driving temperature disference for primary to secondary heat ' transfer will remain unchanged. These two parameters, initial energy content and rate of energy transfer, are the means by. which mass and energy releases influence containment response for i the transients analyzed in Chapter 6 of the FSAR. Since the l reduction in MMF is being made with a negligible change in'RCS temperature, the mass and energy releases calculated in FSAR Chapter 6 will not be affected. l i Accident Analyses, FSAR Chapter 15 All of the FSAR Chapter 15 accident analyses which are applicable i to Catawba Nuclear Station have baen explicitly analyzed with an l initial RCS flow assumption which corresponds to a MMF of 382000 gpm, or have been evaluated to determine the impact of a reduction in MMF of 3000 gpm. As shown in the updated FSAR, the following analyses have been . analyzed with an initial RCS flow assumption which is less than or equal to a MMF flow of 382000 gpm. The results of the analyses demonstrate that all acceptance criteria are met, and therefore a MMF of 382000 gpm is acceptable: 15 .1. 5"' Steam System Piping Failure 15.2.3b Turbine Trip - Peak Primary Pressure 15.2.6 Loss of Non-emergency AC Power 15.2.7 Loss of Normal Feedwater Flow 15.2.8 Feedwater System Pipe Break , 15.3.1 Partial Loss of Reactor Coolant System Flow ' 15.3.2 Complete Less of Reactor Coolant System Flow 15.3.3 Locked Rotor ' 15.4.1 Uncontrolled Bank Withdrawal from Subcritical 15.4.2 C' Uncontrolled Bank Withdrawal at Power , 15.4.3C8 Rod Assembly Misoperation ' 15.4.8"' Rod Ejection 15.6.3C8 Steam Generator Tube Rupture  ; 15.6.5 Loss of Coolant Accident t l Notes: 1) The updated FSAR Table 15-4 is incorrect for  : these events. The steam system piping' failure, FSAR 15.1.5, and the rod ejection accident, FSAR 15.4.8, analyses have been submitted in Duke  ! Power topical report DPC-NE-3001-PA. Table 15-4 for each station will be corrected in the next FSAR update.

2) The uncontrolled bank withdrawal at power, FSAR 15.4.2, and rod assembly misoperation, FSAR 15.4.3, events rely on cycle-specific reload analyses. Since the cycle specific analyses will  :

be performed with a flow assumption.of 382,000

• gpm, FSAR Table 15-4 will be revised.in the.next FSAR update.

3) The steam generator tube rupture'(SGTR), FSAR 15.6.3, event was inadvertently omitted from Table 15-4 of the updated FSAR.- Table 15-4 of the Catawba Oct 91 FSAR update presented the correct input assumptions for the Catawba SGTR analysis. ,

Table 15-4 will be corrected in the next FSAR > update. . 1 e -nn - r As stated in Duke Power Topical Report DPC-NE-3G02-A, certain events are. bounded by other more limiting events, and therefore are not analyzed and the results of these events are not affected by a change in MMF. The events which are bounded by other more limiting events are: 15.1.1 Reduction in Feedwater Temperature 15.1.4 Inadvertent Opening of a Steam Generator Relief Valve 15.2.2 Loss of External Load 15.2.4 Inadvertent Closure of Main Steam Isolation Valves 15.2.5 Loss of Condenser Vacuum and Events Causing Turbine Trip 15.3.4 Reactor Coolant Pump Shaft Break -15.5.1 Inadvertent Opcration of ECCS 15.5.2 Increase in Reactor Coolant Inventory The remaining Chapter 15 events which apply to Catawba Nuclear Station are events which are analyzed with the acceptance criterion of no DNB. These transients are non-limiting with respect to DNB, and DNB is not seriously challenged in any of these events. Therefore, a reduction in MMF of 3000 gpm is not significant to the results of the following analyses: 15.1.2 Increase in Feedwater Flow 15.1.~ Excessive Increase in Secondary Steam Flow 15.4.4 Startup of a Reactor Coolant Pump at an Incorrect Temperature 15.6.1 Inadvertent Opening of a Pressurizer Relief Valve Conclusions As shown above, all of the applicable FSAR analyses have been explicitly analyzed with an initial assumption which corresponds to a MMF of 382,000 gpm, or have been evaluated to determine the impact of a reduction in MMF of 3,000 gp.a. ' Therefore, a decrease from 385000 gpm to 382000_gpm in the Catawba Technical Specification minimum measured flow will not adversely affect the steady state or transient analyses documented in Chapters 3, 4, 6, and 15 of the FSARs. [ _ _ _ . . . - ) l

• I ATTACHMENT III Analysis to Support the Conclusion of No Sionificant Hazard The following analysis, performed pursuant to 10 CFR 50.91, shows that the proposed amendment will not create a significant hazards  !

consideration as defined by the criteria of 10 CFR 50.92. " I

1. This amendment will not significantly increase the probability  !

or consequence of any accident previously evaluated. { i No component modification, system realignment, or change in ( operating procedure will occur which could affect the i probability of any accident or transient. The reduction in , flow will not change the probability of actuation of any i Engineered Safeguard Feature or other device. The consequences of previously-analyzed accidents have been found to be insignificantly different when the reduced flow rate is assumed. The system transient response is not affected by j the initial RCS flow assumption, unless the initial i assumption is so low as to impair the steady-state core cooling capebility or the steam generator heat transfer i capability. This is clearly not the case with a <1% reduction ; in RCS flow  !

2. This amendment will not create the possibility of any new or different accidents not previously evaluated. 3 No component modification, system realignment, or change in operating procedure will occur which could create the possibility of a new event not previously considered. The ,

reduction in flow will not initiate any new events. t

3. This amendment will not involve a significant reduction in a i margin of safety, f As described in Attachment II, the decrease in'RCS flow has been analyzed and found to have an insignificant effect on the  !

applicable transient analyses found in the FSAR. In order  ! to support the reduced flow rate, the OTAT and OPAT setpoint  ! equation constants have been revised. There is no significant ' reduction in a margin of safety. , l i l i l --- ,