ML20059C390

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Proposed Tech Specs for Reduction of Required Min Measured RCS Flow from 385,000 Gpm to 382,000 Gpm
ML20059C390
Person / Time
Site: Catawba, McGuire, Mcguire  Duke Energy icon.png
Issue date: 10/25/1993
From:
DUKE POWER CO.
To:
Shared Package
ML20059C387 List:
References
NUDOCS 9311010142
Download: ML20059C390 (47)
v • d • e


Text

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Attachment Ia Marked-up Technical SDecification Paces Catawba 1

1 9311010142 931025 #9 .,

p DR. ADOCK 05000369 $d PDR 95b

~ . . . . . . .. ..

~' - - - _ _ _ . _ _ _ __ - - - - ~ -- ~ _ _ _ _ _ _ _

E t 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 S IT - FOUR LOOPS IN OPERATION.-d/J(('I

/'RO S T70t/

FIGURE 2.1-lb REACTOR CORE SAMU LIMIT - FOUR LOOPS IN OPERATICN.7MM 2 Na s-62 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES SECTION 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE (FOR UNIT 1).................................... B 2-1 4:2.1Jf.EIRCOR E ~ EOF U NI-T-4. . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... B 2-2 2.1.2 REACTOR COOLANT SYSTEM PRESSURE................ .............

B2-2f 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... B 2-3 CATAWBA - UNITS 1 & 2 III Amendment No. 8'E (Unit 1)

Amendment No. 40 (Unit 2)

A LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4._ 1 BORATION' CONTROL -

Shutdown Margin - T,yg > 200 F........................... 3/4 1-1 Shutdown Margin - T,yg < 200 F............................ 3/4 1-3 Moderator Temperature Coefficient... ....... ........... 3/4 1-4 Minimum Temperature for Criticali ty. . . . . . . . . . .......... 3/4 1-6 3/4.1.2 B0 RATION 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............................... 3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 Borated Water Sources - Operating........................ '3/4 1-12 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 Position Indication Systems - Operating.................. 3/4 1-17.

Position Indication System - Shutdown. . . . . . . . .. . . . . . . . . : ~. 3/4 1-1B Rod Drop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Bank Insertion Limits............................ 3/4 1-21 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE CUnit-1)........................... 3/4f@-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (X,Y,Z) (Un41). . . . . . .

q 3/4fi2-3' CATAWBA - UNITS 1 & 2 IV Amendment No. (Unit 1)

Amendment No. (Unit 2)

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (W it 1)........ 3/4 %2-7 3/4.2.4 QUADRANT POWER TILT RATIO EUM t 1)-. . . . . . . . . . . . . . . . . . . . . . . 3/4 K2-10 3/4.2.5 DNB PARAMETERS (4fn * +).................................. 3/4\2-13

~

TABLE.3.2-1 DNB PARAMETERS (Unit 19............................... 3/4 q-15 FIGURE 3.2-1 REACTOR COOLANT SYSTEM TOTAL FLOW RATE VERSUS RATED THERMAL POWER-FOUR LOOPS IN OPERATION (Unit 1)........... 3/4 A2-16

^ -

3/4.2.1 AX I A L F LUX D I F F E RE fic t Smit =e. . . . . . . .i . . . . . . . . . . . . . . . . 3/4 t5W 3/4.2.2 AT FLUX HOT CHANN L FACTOR - FQ(Z) (4fni ............ 3/4 B2-3

)4.2.3 R TOR COOLANT SYSTE. FLOW RATE AND NUCLEAR "NTHALPY RIS HOT CHANNEL FACTO 2)............... ......... 3 B2-9 3/4. 4 N(Unit QUADRA T POWER TILT RATIO (. Unit 2)................ ...... 3/4 -12 3/4.2. DNB PARA "TERS (Unit 2)....... ..................... .... 3/4B2-g TABLE 3. -1 DNB PAR ETERS (Unit 2)......s.......... A/4-B M 6 FI6012L 7 2.- I l'EA noR CcClpLW Sf5TQf TOTA L lio. .W(Wf ........ g gg g 3/4.3 INSTRUMENTATION VetRhy ttIT6L TH6fddA L foldCld- fR/C LWJr REACTOR TRIP SYSTEM INSTRUMENTATION. IP . . . . .uNif

. . . .1. .epEfcftad . '

3/4.3.1 ..........

TABLE 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 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS........................... 3/4 3-27 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-37 i TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-42 CATAWBA - UNITS 1 & 2 Va Amendment No' $ (Unit 1) .

Amendment-No.f0 (Unit 2)

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ,

)

2.1 SAFETY LIMITS REACTOR CORE hERM MN @

h 2.1.1 The comb 4 nation-of-TMRMf1-POWERr-pressur42er-pressur+,-and-the-highest operating icep coolant temper-ature (Tm) shall not exceed the 'imits shown i-n rigure 2J-la fer four leep cperation. l APPLICABILITY: MODES 1 and 2. [jp.g ACTION: Z'pg (foA UA>F

.SH6&A IN I'b(l henever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer -

pressure iine be in HOT STANDBY within I hour,.and-comply with-the-require- 1

- ments of Spec, fir.ation 6.7.1. b6T6Rhl% IF A SAf67Y Letrr hrs BE15f] Via> TIE 6 '

PRICR To STN([UQa REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

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

l ACTION ,

/ l MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in H0T STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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.

)

CATAWBA - UNITS 1 & 2 2-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

. 4 Insert for Technical Specification 2.1.1:

The maximum local fuel pin centerline temperature shall be less than 5080 - ( 6. 5 x 10-3) x (Burnup, MWD /MTU) *F. Operation-within  :

this limit is primarily assured by compliance with the  :

overtemperature AT and overpower AT trip functions, including the axial imbalance limits.

The DNBR shall be maintained greater than the statistical limit for the BWCMV correlation. Operation within this limit is primarily assured by compliance with the overtemperature AT and overpower AT trip functions, including the axial imbalance limits.

1 l

l l

i l

'l I

l l

i 4

RE PLRCE 6~

{

.OTAL F OW - 385000 GPM 660 -

~

655 2 ~

- 2455 psia UNACCEPT LE g

OPERA CN 645 5*

2400 psia 2

640 7 635 f 2230 psia

~

O

- d c 625 :

?o

- 2100 ps e 620 2 h

J 615 - i A

1 .945 La 610 -

h 605 2 ~

)

600 3 2 ,

i l Srt5 {

3 ACCEP ABLE )

590 -

OP TlON i 2 I Ses 2~ l l

] i 580 . .

0.2 0.4 0.6 0.8 1 1.2 0

Fracuan of Rated Thermal Power FIGURE 2.1-1 REACTOR CORE SAFETY LIMITS - FOUR LOOPS 1N OPERATION ,

I.

CATAWBA - UNITS 1 & 2 2-2 Amendment No. 107 (Unit 1)

Amendment No. 101 (Unit 2)

Uhlrf 1 ONLY 670 DO NOT OPERATE IN THis AREA 660 -

650 --

2455 pslo 640 2400 psic h630 - 2280 pslo 0) b b620 -

" 2100 pslo 610 -

DNB Potometers 1945 pslo Technical Specification 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 PROTECTION LIMITS - FOUR LOOPS IN OPERATION CATAWB A - UNIT 1 2 - A2 i

Utarf 2. OMLV i

670 l DO NOT OPERATE IN THIS AREA 660 -

2455 psia 650 -

640 - 2400 pslo e 2280 pslo o 630 -

h 3

6M -

2100 psia 610 -

DNB Parameters 1945 psio Technical Specification 600 -

T 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-Ib REACTOR CORE PROTECflON LIMITS - FOUR LOOPS IN OPERATION CATAWBA - UNIT 2 2-B2

v y , .

Onh I onb

^

v - / _- .

TABLE 2.2.-l .I -

9 REACTOR TRIP SYSTEM IflSTRUMENTATION TRIP SETPOINTS E"

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE t

E 1. Manual Reactor Trip ,

H.A. N.A.

2. Power Range, Neutron Flux 6 a. High Setpoint $109% of RTP* $110.9% of RTP-

't

b. Low Setpoint s25% of RTP* s27.1% of RTP*
3. Power Range, Neutron Flux, s5% 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, fleutron Flux s25% of RTP* s31% of RTP*

N 5

{g 5. Source Range, Neutron 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***

(( 9. Pressurizer Pressure-High s2385 psig. s2399 psig iLL 593.8% of instrument span

10. Pressurizer Water Level-High s92% of instrument span

[! [

!A
: 11. Reactor Coolant Flow-Low 190% of loop minimum '

188.9% of loop minimum f F measured flow ** measured flow **

!!5!

7- *RTP - RATED THERMAL POWER 9fsoo

TE ** Loop minimum measured flow = -00,250 gpm l' OT - * ** Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead m and 1 second for lag. Channel calibration shall- ensure that these time constants are adjusted to these values.

_ _ _ _ _ _ _ _ _ ____._.__.m_ . - . . - , - . _ w- . , _ _ * ~ ~ - - . -;,. m . _. . . . .,

o u - -

v - Oonik 2 o N  !

TABLE 2.2.-l I.

9 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS A

s> TRIP SETPOINT All0WABLE VALUE FUNCTIONAL UNIT t

si 1. Manual Reactor Trip ,

N.A. N.A.

Z y 2. Power Range, Neutron Flux

" a. High Setpoint s109% of RTP* $110.9% of RTP*

~

b. Low Setpoint s25% of RTP* s27.1% of RTP*
3. Power Range, Neutron Flux, s5% 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*

5

'5. Source Range,' Neutron Flux . s105 cps sl.4 x 10 cps

' Gd '

4 6. Overtemperature AT See Note 1 See Note 2

7. Overpower Ai See Note 3 See Note 4 ,
8. Pressurizer Pressure-Low 21945 psig 21938 psig***

2 m

&  !! 9. Pressurizer Pressure-High $2385 psig s2399 psig E 1

! j! 10. Pressurizer Water level-High s92% of instrument span s93.8% of instrument span L%

2h 11. Reactor Coolant Flow-Low 190% of loop minimum measured flow ** -

188.9% of loop minimum measured flow **

.c P NN

~~ *RTP - RATED THERMAL POWER

_Wf *

  • Loop minimum measured flow - 96,250 gpm l M * ** Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead .

-m - and I second for lag. Channel calibration shall ensure that these time constants are adjusted to

  • these values.
  • =____m_._ms_______m___._.__a_ = -a &_ m-=w -_s- - _ _ _ m -_w, ,_- . _ e,- _ i _s es e -e m e-- s
  • _-_ . _ em-m a_-__ _

TABLE 2.2-1 (Continued) U ,4 4 l o d y i TABLE f(OTATI0f15

l. ,

y fl0TE 1: OVERTEMPERATURE AT c

5 1 + ril) ( 1 ) s AT, (K I ) - T'] + K3 (P - P') - f, (AI))

, AT ((1 + r S) (1 2 5) + r3 i - K (I 2

1 ++ * ) [T(I r S) s 1 + r S)e b Where: AT = lieasured AT by Loop flarrow Range RTDs; M

- 1 + riS = lead-lag compensator on measured AT; n - 1+rS 2 9 = Time constants utilized in lead-lag compensator for AT, ri = 12 s, ri , r2 r2 = 3 s; 1 - Lag compensator on measured AT; 3

r3

- Time constants utilized in the lag compensator for AT, r3 = 0; AT, = Indicated AT at RATED THERMAL POWER;

{Yb 4 K, = -1.13D l.l9 54 K2 =

043M3/*f 0 OM7I 1 + r45 - The function generated by the lead-lag compensator for T,y 1+ras dynamic. compensation; jj' i,7

.3 a

- Time constants utilized in the lead-lag compensator for T,y, r4 - 22 s, E' r4 , r-s '

k {L  :

. rs = 4 s; E,5I T - Average temperature, *F; ..

= lag Compensator on measured T,y;-

22

$f; r, = Time constant utilized in the measured T,y lag compensator, re = 0; g-T

. - ~ ._a N

__ n a a-- - - __.a__. --_.,-.--._.,w__.,__n.=.x__,x---_. _-----__._._._-----_.-_.---_.,_,___-.------____.__n. , , .,..w,.e-nn-,,m- , n ,m u--mm -msw w a

N .

TABLE 2.2-1 (Continued) dnN 2. o .

TABLE NOTATIONS UOTE 1: OVERTEMPERATURE AT 1 + r,1) ( I 5 1 )

s AT, (K - K 2 ((1 + r S)1 + r,5) [T(I) - T'] + K3 (P - P') - fi (AI))

. AT ((1 2+ r S) (1 3 + r S) i s 1 + r oS) b Where: AT - Measured AT by Loop Harrow Range RTDs; g -

. L 1 + riS = lead-lag compensator on measured AT;

d. 1 + r2S ri , r2

= Time constants utilized -in lead-lag compensator for AT, ri - 12 s, '

r2 - 3 3; l

I - Lag compensator on measured AT; ,

j, 3 Time constants utilized in the lag compensator for AT, r3 - 0; t

, r3 AT, - Indicated AT at. RATED THERMAL POWER; i

}TQ -

< K, - 1.1953 l K.2 0.03163/*F 1 + r,S - The function generated by the lead-lag compensator for T.,

  • SFif. 1 + r3S dynamic compensation; E3
r. , rs Time constants utilized in. the lead-lag compensator for T,,,, r, - 22 s, 3d -

rs - 4.s;

$l  ;

i

!? !? T - Average temperature, 'F; .

I = Lag compensator on measured T,,,;

1+rS o g.

7;f,. re

- - Time constant utilized in the measured T,,, lag compen'sator, r, = 0; LiC

.g

-_.._..____.__________..___________m_m_

___.i__.,_____.__-.2. .__ m._. . - .>. . .~ - . _ . . . _ . - , ~ . . , . . m ,--

v TACLE 2.2-1 (Continued) UMI4 '6 o nI l-TABtE NOTATIONS (Continued) f ,

n 3 NOTE 1: (Continued)

E 5 T' s 590.8*F (Nominal T., allowed by Safety Analysis);

c K3 - 0.001414; o.oo ts23 5

E P - Pressurizer pressure, psig; P' - 2235 psig (Hominal RCS operating pressure); j r,

S = Laplace transform operator, s;

and_ f,(AI) 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 STARTUP tests such that: ,

-42.o + 8.o (i) For q, - q, between --39-9% and +3-e%,

Ni f,(AI) = 0, where q, and q, 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;

- 42.0 (ii) For each percent AI that the magnitude of q, - q, is more negative than -iMhM, the AT Trip Setpoint shall be automatically reduced by 3-910% of AT,; and

3. M 2. 9 g,o For each- percent AI that the magnitude of q, - q, is more positive than +h0%, the AT Trip pp (iii)

Setpoint shall be automatically reduced by ih-3%% of AT,.

ge 1 64 o g

3E NOTE 2:

The channel's maximum Tri) Setpoint shall not exceed its computed Trip Setpoint by more -

{g Er than 3dL 4.5'/, $4m % e d h er m o

!)

as f

2:

- _ _ _ _ _ _ - - _ _ _ - - _ _ . _ - _ - _ _ _ = _ , - _ _ . . ._- - . . , .

.2 v .

v TABLE 2.2-1 (Continued) (.)x;d 2 on l-TABLE NOTATIONS (Continued)

E NOTE 1: (Continued)

E 5 T' s 590.8'F (Hominal T., allowed by Safety Analysis);

c K3 - 0.001414; 5

A P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure);

N 5 - Laplace transform operator, s;

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

(i) For q, - q3 between -39.9% and +3.0%,

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

a c4 (ii) For each percent AI that the magnitude of q, - q, is more negative than -39.9%, the AT Trip Setpoint shall be automatically reduced by 3.910% of AT,; and For each. percent AI that the magnitude of q, - q, is more positive than +3.0%, the AT Trip

, pp (iii)

Setpoint shall be automatically reduced by 2c316% of AT .

gg Sir 8E~

r NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more g?g? than 4-05. 4.5 % gg gm{ p C5 5b :

T:

r-___ _ __a t- --2 -a ______--_ rat _ -

c-l_. __ u +t---m ta e 4 -- m - -d. w ..._.e - p ga r e+ _

~

"\ .

TABLE 2.2-1 (Continued) Unth Ony TABLE NOTATIONS (Continued) l-g NOTE 3: OVERPOWER AT c

5 1 + riS ) { l ) I I I I - T") - f (al))

. AT ((1 + r:5) (1 + sr 5) 3 AT. (K. - K1 +(Id'E-)

5 r,5) (1 +I r 5) T - K* [ T 1 +(Ir 5) 2 E

k -

Where: AT - As defined in Hote 1, i- I + riS) - As defined in Note 1, r3 1 + r25 ,

r, , r3 - As defined in Note 1, 1 - As defined in' Note 1, 1 + r3S r3

- As defined in Note 1, N .

jg AT, - As defined in Note 1, h -

-1.0013-l.0969 Ks -

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

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

@3 1 + r,5 compensation, "k a= k

$FJ r, - Time constant utilized in the rate-lag controller for T,, r, - 10 s,

" {T

" I As defined in Note 1

@.b' I+rS w ,,;o o I2 r. - As defined in Note 1, We

$ 3.

bb

-a- - __

---___----___--__------.,--___-__-----___x*-a---_._-- - _ -__ _ ._- n v

TABLE 2.2-1 (Continued) On,h 2. e, fy TABLE NOTATIONS (Continued) l-9 '

g NOTE 3: OVERPOWER AT c

$ 1 + r,5) { l ) 1 I

- I - T"] - f,(AI))

AT ((1 + r,5) (1 + r3 5) s AT, {K. - 3K1(I+ r,L) r,5) (1 ( + r 5) ) T - K [I (I1 + r,5)

.E.

C Where: AT - As defined in Note 1, N 1 + ril) - As defined in Hote 1,

~ l + r,S ,

ri , r, - As defined in Note 1, 1 - As defined in' Note 1, I + r35 r3

- As defined in Note 1, 7 AT, - As defined in Note 1, Cb d

h - 1.0819 Ks - 0.02/*F for increasing average temperature aqd 0 for decreasing average tersperature, t

yif __r,S__ - The function generated by the rate-lag controller for I, dynamic

?g 1 + r,5 compensation, i;e@ E  !" r, - Time constant utilized in the rate-lag controller for T,, r, - 10 s,

,1 .

d::e I

.c F - As defined in Note 1

, _1 I+rS .

g$$ r, - As defined in Note 1, c: c:-

~ ..

~

m

~

" - Unk \; o ,\q ,.

TABLE 2.2-1 (Continued) l

I TABLE NOTATION U Continued) 9 5 NOTE 3: (Continued)

Ei o. co a 2 h2.

K3 - 0.001291/*F for T > 590.8'F and K 6 - O for T s 590.8*F, 7

E T - As defined in.flote 1, 3

  • T" Indicated T a insi.romentaf. s Yon,t RATED THERMAL POWER (Calibration temperature fo 590.8'F),

o

" S - As defined in tiote 1, 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 q -q between -35% and +35% AI; f2 (AI) - 0, where q and 93 are percentisRATED total THERMAL POWEk in Ihe top cnd bottom halves of the core respectively, and q, + 9 3 THERMAL POWER in percent of RATED THERMAL POWER;

,, y ,

3 for each percent AI that the magnitude of q - o is more negative than -35% AI, 6O the(11)

AT Trip Setpoint shall be automatically reduced,by [.0% of AT,; and (iii) for each percent AI that magnitude of q - q is more positive than +35%-AI, the AT Trip Setpoint shall be automatically reduced by 7.D% of oAT .

>> The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more 38 NOTE 4:

RE than 4-eh 5 3.o */o 2ded Tkered Power iE

. h 38 22 b$

n= ,

aummam

+-._____.m____.a________m.______._____________-_________:_.__ +- _ _ _ . _ _ . _ 2 _ _- u__ _ -.m ., = rm-.,. -. v c.

TABLE 2.2-1 (Continued) l

- l' TABLE NOTATIONS (Continued) 9 y NOTE 3: (Continued) en K3 - 0.001291/*F for T > 590.8'F and K 6 - O for T s 590.8'F, f

E T - As defined in. Note 1,

% T" - Indicated T a instrumentaI. Ton,t RATED THERMAL POWER (Calibration temperature for AT f

s 590.8'F),

As defined in Note 1, S

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 q -q between -35% and +35% AI; f2 (AI) = 0, where q and q3 are percent RATED THERMAL POWE[1 in The top and bottom halves of the core respectively, and qt + q3 is total THERMAL POWER in percent of RATED THERMAL POWER; np for each percent AI that the magnitude of q - is more negative than -35% AI, (ii) the AT Trip Setpoint shall be automatically reduced,by [.0% of AT,; and (iii) for each percent AI that magnitude of q - qs is more positive than +35% AI, the AT Trip Setpoint shall be automatically reduced by 7.0% of AT,.

"Mm3I NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more

&R than-A S%.

3 37pf ROcb Thermd %er we 22

5. 5.

ne

Unit } on If i-3G5,92o [

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a,2,w eestse i 3Y,0f50l amse SS 88 90 92 94 96 98 10 0 10 2 Fraction of Rated Thermal Power Figure' 3.2-1 Reactor Coolant System Total Flow Rate Versus Rated Thermal Power - Four Loops in Operation l

$2.-l b CATAWBA - UNIT / l 2- 3/4 4-46 -Amendmon Aman h:nt "c. ' t fin J[] (Uri ty(Unit 2)

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a,gien 3 8 5 0 0 0 - ~ - - - - - - - - - -- -- ---- ---- - - - - - - - - - -- - - - - - - - - - ---- - 1.'.*A .'.'.*.W.'.

Restricted Operation (s e.3anso)

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@ Region

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c (90.3596001 j

S 365600-l l

1 1

365750 i

I 1 361900 86 88 90 92 94 96 98 10 0 10 2 Fraction of Rated Thermal Power Figure 3 2-1 Reactor Coolant System T6tal Flow Rate Versus (

Rated Thermal Power - Four Loops in Operation l CATAWBA - UNIT / 4-4-2 3/482-16 - A-:endmen t "c.107 (Ur;i t ! }-_ l Amettdment 14c . ! n1 (tin i t y

, 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE l

The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission 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 coefficient. DNB is not a directly measurable parameter during operation and _ IpQ therefore THERMAL POWER and Reactor Coolant Temperature /pnd Pressure have Deen related to DNB through the BWCMV correlation. The BWCMV DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio, (DNBR), is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95%

probability that the minimum DNBR of the limiting rod during Condition I and

) 11 events is greater than or equal to the DNBR limit of the DNB correlation being used (the BWCMV correlation in this application). The correlation DNBR 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 DNBR is at the DNBR limit.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters, and the BWCMV DNB correlation are considered statistically such that there is at least a 95% confidence that the minimum DNBR for the limited rod is greater than or equal to the DNBR limit. The uncertainties in the above parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty is used to establish a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

g21-14 -fw Un;4 ) ad 2.l-l b -br Wd 2 p The curves of Figure 1-1 rhe': the leci cf peint cf mgmi mgggg, Rc2cter Coolant Syste-' presur+and average temperature belce dich the calculated DNBR is ac les than the dc;ign DNBR value, or the average-enealpy-at the verse' crit is le:: thar the enthalpy of ::turated, liquid.

CL CC MoC E- resketchlve % hc. ack& Se[e lk;{ c.ar VCS,

,I i l

l CATAWBA - UNITS 1 & 2 B 2-1 Amendment No. /(Unit 1) ]

Amendment No.' (Unit 2)

=. .

2.1 SAFETY LIMITS gg gg, g BASES /

There cur"m 're ased on a nuclear enthalpy rise hot channel factor, FL, of 1.50 and a reference cosine with a peak of 1.55. An allowance is included for an increase in F", at reduced power based on the expression:

FL = 1.50 [1 + 1/RRH (1-P)]

Where P is the fraction of RATED THERMAL POWER.

RRH is given in the COLR.

These limiting heet flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (AI) function of the Overtemperature AT trip.

3 When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits. gj g g d pg.

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,  ;

valves, and 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 design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements. 1 The entire Reactor Coolant System is hydrotested at 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation, j I

l l

l CATAWBA - UNITS 1 & 2 8 2-2 Amendment No. HW(Unit 1)

Amendment No. M1(Unit 2)

-1 4- ,

Insert for Technical Specification 2.1.1 Bases Since pin power peaking is not directly measurable, fuel melt limited power peaks am separately .)

correlated to measured reactor power and imbalance. When the combination of mactor power and axial power imbalance is not within tolerance, the OPDT trip function will provide the necessary ,

fuel pin centerline temperature pmtection. -

b b

-l i

1 h

i l

')

l 1

l e

l

-I

& 4

'j Attachment Ib Marked-up Technical-Specification Paces McGuire f

i P

3 k

1 9

a T..-

i INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS l

2.1.1 REACTOR C0RE................................................. 2-1 l 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.........y ... ............... 2-1 l FIGURE 2.1-1 UNITS 1 and 2 REACTOR CORE $AFETY LIMIT - FOUR LOOPS .

l

__ IN OPERATION... ...................................... 2-2 l

mpfMA%tasea.#.wowCoa r .w.#e i

2.2 LIMITING SAFETY SYSTEM SETTINGS I

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ............... 2-4 l TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-5 l BASES

/

I

- ) SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE .................................... ... B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE........................ .. B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS................ B 2-3

) c McGUIRE - UNITS 1 and 2 III Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)  !

4

  • 2.0 SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS l 2.1 SAFETY LIMITS pgg uj' p)19J1GWA & ATThu%b REACTOR CORE F4gg 2.1.1 The combination of Tl4ERMA[ DOUEQ, presggpjzer pregggre, end the highest operatingJ oop coolant temper 4ture (Tavg) shall not exceed the H=its shown in Sigures 2.1-1 and-24-2-for-four-and-three4 cop-operation, respectively. l APPLICABILITY: MODES 1 and 2 l

^

"hgovJfJ//J []GURE L' henever the point defined by the combination of the highest operating loop average _ temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure 1Tnirp be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, .an&-comply "ith the r -

sents ec ca 5 set &Rf1/ME g a sygry tuurNAS B U VI*MO-REACTOR COOLANT SYSTEM PRESSURE j 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. I 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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, l reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

i l

1 i

McGUIRE - UNITS 1 and 2 2-1 Amendment No. (Unit 1) ]

Amendment No. (Unit 2) t-

-Insert for Technical Specification 2.1.1:

The maximum local fuelipin centerline temperature shall be less.

than 5080 - (6. 5 x 10'3) . x (Burnup, MWD /MTU) 'F.. Operation within this limit is primarily assured by compliance' with the overtemperature AT and overpower AT trip functions, including the axial imbalance limits.

The DNBR shall be maintained greater than the statistical limit for ~

the BWCMV correlation. Operation within this limit is primarily assured by compliance with the overtemperature AT and overpower AT' trip functions, including the axial imbalance limits.

1 l

l l

l i

l

1 8.EPLACE  :

Figure 2.1-1 Reactor Core Safety' Limits -

Four Loons in Operation- ,

665 -

FLOW PER LOOP 796250 GPM  ;

i- 660 5 655 2 -

2455 paia UNAC EPTABLE ,

650 O RATION 645 2 ~

1 2400 psia 640 5 i

. I 635 5 .

2280 psia 630 5-O .

5>-625 5 -

3 -

2100 sia 31 620 --

E  : , I EIS 2 610 1945 La 605 -

_I 600 5 l 595 5 .

g

ACCEP BLE 590 . OPEF TION 535 2 550 - ,

0.00 0.20 0.40 0.60 0.80 1.00 1.20 j Fraction of Rated Thermal Power .

McGUIRE - UNITS 1 and 2 '2-2 AmendmentNo.1 (Unit 1)

Amendment No. .(Unit 2)

1 670 DO NOT OPERATE IN THIS AREA 660 -

650 -

2455 pslo 640 2400 pslo b630 -

2280 pslo e

3 0 620 -

2100 psia 610 -

DNB Porameters 1945 pslo TechnicalSpecification 600 -

4 590 -

ACCEPTABLE OPERATION 580 O.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1-1 REACTOR CORE PROTECI' ION LIMITS - FOUR LOOPS IN OPERATION McGUIRE - UNITS 1 and 2 2-2 1

= - _ ~ . _ . . .

b 6 /. 6 T $ ~

1 i

i 4

1

__. , i 1

Figure 2.1-2 left b ank pending NRC i

) . approval of re loop operation I

i i

) -

1 1

McGUIRE - UNITS 1 and 2 2-3

x TABLE 2.2-1 S

' Si REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l 5 i FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES C

23 1. Manual Reactor Trip N.A. N.A.

Y

~ 2. Power Range, Neutron Flux Low Setpoint 125% of RATED Low Setpoint 5 26% of RATED m THERMAL POWER THERMAL POWER R

m High Setpoint 5 109% of RATED High Setpoint 1 110% of RATED THERMAL POWER THERMAL POWER

3. Power Range, Neutron Flux, < 5% of RATED THERMAL POWER with < 5.5% of RATED THERMAL POWER High Positive Rate a time constant 2 2 seconds with a time constant 1 2 seconds
4. Intermediate Range, Neutron $ 25% of RATED THERMAL POWER 5 30% of RATED THERMAL POWER i Flux 7

w

5. Source Range, Neutron Flux i 105 counts per second -

1 1.3~x 105 counts per second

6. Overtemperature AT See Note 1 See Note 3
7. Overpower AT See Note 2 See Note 4
8. Pressurizer Pressure--Low 1 1945 psig 1 1935 psig H *b 9. Pressurizer Pressure--High 5 2385 psig 5 2395 psig K

F

=, ,m

10. Pressurizer Water Level--High 1 92% of instrument span 1 93% of instrument span  ;

rF i+

-. , > 90% of minimum measured > 88.8% of minimum measured '

{ { 11. Low Reactor Coolant Flow T10w per loop * . flow per loop

  • FG$ 95~Sbo y
  • Minimum measured flow is 05,250 gpm per loop.

vs

$?**

d!aj ^

MMM

=m w- c --w a v e a* v r-- = +<>m--+- ~+- 4

g TABLE 2.2-1 (Continued) -

cn

5. REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 2

i NOTATION E

h NOTE 1: OVERTEMPERATURE AT s

= 1+TS 1 E

m (AT/AT g ) (1 {j)(1+taS)IE l l - E2 (1 + r S)[T(1 , ,c3)-T 9 + K3 (P-P') - f1(AI)

Where: AT = Measured AT by Loop Narrow Range RTD ATg = Indicated AT at RATED THERMAL POWER, I

j = Lead-lag compensator on measured AT, 7

t i , r2 = Time constants utilized in the lead-lag controller for AT , 12 > 8 sec., T2 5 3 sec., ,

= L g c mpensator on measured AT, 1 Ta

=

Ta Time constants utilized in the lag compensator for AT, Ta i 2 sec.*

- B" P K < 1. 1 9 5 0 , l. R 8 8 me 1 -

K 2

  • 0'03143 #'O T4

- a :!

"T 1 + T45

= Th f nction generated b'y the lead-lag controller for Tavg dynamic compensation, 1 + IsS

'$ld 14 Is = Time constants utilized in the lead-lag coritroller for T i

  • 14 > 28 sec. Is < 4 sec., av9' c2 -

%u 3' T = Average temperature, F, N

" 1 = Lag compensator on measured T y ,7s3 avg'

~

. e

~

~.

2 TABLE 2.2-1 (Continued) 1 0

5 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS A

, t10TATION'(Continued)

C 5 NOTE 1: (Continued) ~

d g ts = Time constant utilized in the measured T avg 1 g compensator, to 5 2 sec g T' =

5 588.2 F Reference T avg t RATED THERMAL POWER, K O' 3 '

P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure),

S = Laplace transform operator, sec 2, 7 and f (AI) is a function of the indicated difference between top and bottom detectors i

e of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

--44.o 4120 (i) for qt ~9b between -93% and 4-h0%AI; f (AI) y = 0, where qt and gb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and g *9 t b gy is total THERMAL POWER in percent of RATED THERMAL POWER; gg -44.o gg (ii) for each percent imbalance that the magnitude of qt 9 b is m re negative than --39%AT, kk

~

the AT Trip Setpoint shall be automatically reduced by G:M3% of ATo, and 3.4 %

k[

(iii) for each percent imbalance that the magnitude of q- q is more positive than

+h0%ol, the AT Trip Setpoint shall be automaticalky rebuced by 1,4-14% of ATo

+i2.o 1.bt9 22 -

S. S.

, or Mr _ _ -

e + - 4

g TABLE 2.2-1 (Continued) -

m

5. ftEACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 2

NOTATION (Continued)

C i'i w

NOTE 2: OVERPOWER AT (AT/AT g ) (1 ) (1 + T3 5) I E4 -E5 (1 17 5) (1 c.

tcS) h0(1 + Ta s)- T"] - f (OI) 2 m Where: AT = As defined in Note 1, AT, = As defined in Note 1, l 1 S

= As defined in Note 1 ti, T2 = As defined in Note 1 N

1

= As defined in Note 1, h 1+T53 K

4 5 1080s, Lo8O K

5

= 0.02/ F for increasing average temperature and 0 for decreasing avei 3e temperature, FF 15 7

@@ = The function generated by the rate-lag controller for T avg dynamic 1 + 175

&R oo compensation, aa

[

r 17 =

Time constant utilized in the rate-lag controller for Tavg, 1 7 > 5 sec, 1

y,73

= As defined in Note 1, 6 ,

- ts = As defined in Note 1, EE 3;; K 6

=

0.0012q0/FfrT>T"andK6 = 0 for T 5 T",

og o.oon.o 7

g

~

TABLE 2.2-1 (Continued)' l a

S REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS A

. NOTATION-(Continued)

E d T = As defined in Note 1, T" =

-< 588.2 F Reference T avg at RATED THERMAL POWER, g ,

m S = As defined in Note 1, and f 2(AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for q - a between -35% and +35% AI; f 2(AI) = 0, where q and q are percent RATEDTHERbALPbWERinthetopandbottomhalvesofthecorerbspectihely,and N qt*9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent imbalance that the magnitude of q is more negative than

-35% AI, the AT Trip Setpoint shall be automatically redbc ed qby 7.0% of ATo; and (iii) for each percent imbalance that the magnitude of q - q is more positive than

+35% AI, the AT Trip Setpoint shall be automatically redbced by 7.0% of ATo.

3 2,. Note 3: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 22 -3A% of Rated Therm;l Power.

aa 44 3.o 2g Note 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than-4-2%

55 of Rated Thermal Power.

EE NO G2

5. E.

rr N03

  • __L.___ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ _ - - _ - _ .- . . _

P 2.1 SAFETY LIMITS l- ,

)

BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission 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 coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95% proba-

) bility that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the BWCMV correlation in this application). The correlation DNBR set such that there is a 95% probability with 95% confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and the CHF cor-relation are considered statistically such that there is at least a 95% con-fidence that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The combined DNBR uncertainty is used to establish a design .

DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties, are More redricAlve Nn 4e sich ThecurvesofFigure2.1-1showthelocicfpointsofT"ERiiALI'0WER,h c .

Rccctor Gee 4 ant System pressure, and overage temparcture balcw which the calculated DNOR i; ne les; then-the de:ign 0"SR value er the average enthcipy at the vc;;cl exit ,

ts-Jecc th'r theDTAT enthalpy ad of cat" -frip oPhT Pated lig"Iriid<a5

-fbr ThecurvcMarebasedonanucTearenIhalpyrisehotchannelfactor,FN I AH' 1.50 and a reference cosine axial power shape with a peak of 1.55. An allow-N ance is included for an increase in F AH at reduced power based on the expression:

N F H = 1.50 [1 + (1/RRH) (1-P)]

) Where P is the fraction of RATED THERMAL POWER, and RRH is given in the COLR.

McGUIRE - UNITS 1 AND 2 B 2-1 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

SAFETY LIMITS .

BASES ( 4 These limiting heat' flux conditions are higher than those calculated for the range of all control. rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f 2 (AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance,- the axial power imbalance effect on the Overtemperature- AT trips will' reduce the setpoints to provide protection consistent with core safety limits. .

% sed ara enpk b M 2.1.2 REACTOR COOLANT SYSTEM' PRESSURE Wade je The restriction of this Safety Limit protects the integrity of the Reactor i Coolant System from overpressurization and thereby prevents the release of radio-nuclides contained in the reactor coolant from reaching the containment atmosphere.  ;

The reactor vessel,and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110%

(2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

(-

l

)

l 1

l McGUIRE - UNITS 1 and 2 B 2-2 Amendment No. MS(Unit 1)

. Amendment No.HE3(Unit 2)

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

'l

.q Insert for Tecimical Specification 2.1.1 Bases

-)

- Since pin power peaking is not diectly measurable, fuel melt limited power peaks am separately l cormlated to measured reactor power and imbalance. When the combination of reactor power and '

axial power imbalance is not within tolerance, the OPDT trip function will provide tie necessary fuel pin centerline temperature protection.

I t

i j

-I 1

I

\

4

_ , . . - . . . - , , . - -. + , . . .

i s * .

l l l

j l

POWER DISTRIBUTION LIMITS

/ \

Figure 3.2 - 1. l l

i Reactor Coolant System Total Flow Rate Versus Rated Thermal Power - Four Loops in Operation 365,92o 4&&%5;

' A penaity of 0.1% for undetected feeewater Permissible

, ventun feuhng anc a measurement uncenainty Opersnen gg

, cf 1.7% for flew are incruceo in this figure. .

%n '

3k2,ooo (g3.4,44 ,,}

m . ______ __ _ _____ ________

P

_.  ? ?S,l &b

[ 3')C, t to (96.39tero) ,

Restncted Operation l

= , Region 374,3(,o, j 3'N,%o II#' I M -

E 2

3')o,59 o Prohibned

.: 370,9to (92.37-3469) Opersoon (

g. gen

, s7sese -

o u ,

i l

I g 3 %,'72 0

( 90.3699to)

-369669 -

P 2(,2,900 4667M P

3 59,o to .

, Wee . . . . ,

86 88 90 92 94 96 96 100 102 f

Fraction of Rated Thermal Power McGUIRE - UNITS 1 AND 2 3/4 2-24 Amendment No.1. (Unit 1) j Amendment No. (Unit 2)

i 6 I J Attachment II Justification and Safety Analysis 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 12, as annotated) required by Technical Specifications to maintain 100% power operation. The proposed reduction in minimum measured flow is applicable to McGuire Units 1 and 2, and to Catawba Unit 1.

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 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. McMeekin, 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 McGuire Units 1 and 2, and 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 current plans to replace the steam generators in both McGuire 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.

The changes to the OPAT setpoints for McGuire Units 1 and 2 and Catawba Unit 1 also 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, reset whenever it is found to be different from the correct value by about 1%, it is unlikely that past operability will be a concern. Investigation into the situation is continuing, and a Licensee Event Report will be submitted, if appropriate. The revised McGuire OTAT allowable value is less restrictive than the existing value and the Catawba allowable value is unchanged by the reduction in flow. Also, to 4

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. = ..

improve clarity, the allowable values of OPAT and OTAT are now expressed in units of % Rated Thermal Power.

DNBR and centerline fuel temperature (CFT) limits have been added to Technical Specification 2.1.1 as Limiting Conditions for Operation (LCO), replacing a reference to Figure 2.1-1, Reactor Core Safety Limit - Four Loops in Operation. This is considered to more accurately reflect the requirements of 10 CFR 50.36. A curve was added to the figure to show acceptable operation as determined by DNBR limits; this curve is more restrictive than the existing figure. This change is applicable to both units at each of the two stations. Figure 2.1-1 was redrawn to reflect the reduced flow rate for McGuire Units 1 and 2, and Catawba Unit

1. The change to DNBR and CFT in the LCO of Specification 2.1.1 is consistent with Babcock and Wilcox Improved Standard Technical Specification presented in NUREG-1430.

Effect of Reduced Flow on FSAR Analyses LOCA Blowdown Forces, FSAR Chapter 3 The primary factors which affect the blowdown forces resulting 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 382000 gpm, and therefore a reduction in MMF to 382000 gpm will not affect the assumptions in the blowdown forces analysis.

Thermal Hydraulic Design, FSAR Section 4.4 The thermal hydraulic design for the McGuire and Catawba units 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 flow rate. The Axial Flux Difference limits, Technical Specification Section 3.2.1, are unchanged and all of the current thermal hydraulic design criteria are satisfied at the reduced flow conditions.

Ar previously noted, revised core thermal limits were generated l 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 l the overpower delta T (OPAT) setpoint equation constants.(Notes 2 and 3 of Table 2.2-1 for McGuire and Catawba, respectively) were revised to reflect the necessary changes. The impact of the .

reduced flow on the coefficients was partially offset by a  !

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- reduction in the margin assumed in the calculation of the coefficients.

Mass and Energy Releases for Containment Analyses, FSAR Chapter 6 The reduction in MMF flow can affect the mass and energy releases for containment analysis only through a change in the NC system temperature-input assumption. RCS average temperature will remain unchanged with the change in MMF. Therefore, the RCS initial fluid and metal stored energy will remain unchanged.

Further, a constant RCS average temperature implies that the driving temperature difference 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 the transients analyzed in Chapter 6 of the FSAR. Since the 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.

Accident Analyses, FSAR Chapter 15 All of the FSAR Chapter 15 accident analyses which are applicable to McGuire Nuclear Station and Catawba Nuclear Station have been explicitly analyzed with an 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-cmergency 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 Loss of Reactor Coolant System Flow 15.3.3 Locked Rotor 15.4.1 Uncontrolled Bank Withdrawal from Subcritical 15.4.2") Uncontrolled Bank Withdrawal at Power 15.4.3 0) Rod Assembly Misoperation 15.4.8"' Rod Ejection 15.6.3* Steam Generator Tube Rupture 15.6.5 Loss of Coolant Accident Notes: 1) The updated FSAR Table 15-4 is incorrect for these events. The steam system piping failure, 1 FSAR 15.1.5, and the rod ejection accident, FSAR l

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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. +

The McGuire SGTR analysis has no explicit flow assumption. Table 15-4 for each station will be corrected in the next FSAR update.

As stated in Duke Power Topical Report DPC-NE-3002-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 i Turbine Trip  :

15.3.4 Reactor Coolant Pump Shaft Break 1 I

15.5.1 Inadvertent Operation of ECCS 15.5.2 Increase in Reactor Coolant Inventory 1

The remaining Chapter 15 events which apply to McGuire and Catawba Nuclear Stations 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 15.1.3 Excessive Increase in Secondary Steam Flow 15.4.4 Startup of a Reactor Coolant Pump at an Incorrect Temperature

.e e-1 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 i to a MMF of 382,000 gpm, or have been evaluated to determine the i impact of a reduction in MMF of 3,000 gpm. Therefore, a decrease from 385000 gpm to 382000 gpm in the Catawba and McGuire Technical Specification minimum measured flow will not adversely 1 affect the steady state or transient analyses documented in l Chapters 3, 4, 6, and 15 of the Catawba and McGuire FSARs. )

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i ATTACHMENT III -l Analysis to Support the Conclusion of No Sicnificant Hazard j l

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.

1. This amendment will not significantly increase the probability or consequence of any accident previously evaluated.

No component modification, system realignment, or change in operating procedure will occur which could affect the probability of any accident or transient. The reduction in flow will not change the probability of actuation of any 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 "

the initial RCS flow assumption, unless the initial assumption is so low as to impair the steady-state core cooling capability or the steam generator heat transfer capability. This is clearly not the case with a <1% reduction in RCS flow.

The change to Technical Specification 2.1.1 to refer to DNB and CFT limits rather than Figure 2.1-1-will not cause the consequences of a previously analyzed accident to increase.

No new mechanisms are introduced which could exacerbate a previously analyzed accident.

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2. This amendment will not create the possibility of any new or different accidents not prestously evaluated.

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.

The change to Specification 2.1.1 will not initiate any new events. The introduced Ty and thermal power limits define a  ;

more restrictive operating range than could be inferred from the existing figure. There are no new mechanisms introduced which could create the possibility of a different accident not previously analyzed.

3. This amendment will not involve a significant reduction in a margin of safety.

As described in Attachment II, the decrease in RCS flow has been analyzed and found to have an insignificant effect on the

.- o 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.

The change to Technical Specification 2.1.1 will not reduce a margin of safety. The limits on T,y and thermal power will ,

provide the reactor operator with meaningful and identifiable indications in the event that normal operating conditions are exceeded.

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