Text

Proposed TS Table 3.3-1 Re Reactor Trip Sys Instrumentation & Table 4.3-1 Re Reactor Trip Sys Instrumentation Surveillance Requirements
	ML20114B676
Person / Time
Site:	Catawba
Issue date:	08/26/1992
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20114B675	List: ML20114B673; ML20114B676;
References
	NUDOCS 9208310084
	Download: ML20114B676 (21)
	v • d • e

. .. .-

4

, \

l l

I s

I ATTACIIMENT II Revised Technical Specification Mark-ups for Catawha Unit 1 Cycle 7 Reload 1

f 9200310084 920826 I PDR ADOCK 05000413 i P PDR .j

l l 2 1 -

N O ,

90 3

'T I 8 8 8 8 8 8 3 8 90 1 1 T 1 C

I A

S M

O ^

4 E *.

L S S ,

BS

AE , ,

3 CD ^

I O 4 4 ,

LM 2 2 2 2 P # , ,

P # , ,* ,^ - ,

A 2 1 1 1 1 1 1 1 13 1 3 1 SE MLL UEB .

MNA A I NR .

NAE 2 3 2 3 3 3 4 2 22 22 N N I HP O

I MCO T

A T

) N d E S e M LP -

u U EI A

n R NR .

i T NT 1 2 1 2 2 2 3 1 11 11 N t S A n N HO

_ o I CT

_. C -

( M -

E 1 T S

_ 3 Y

S S 3 .L P OE L I Nf r .

L R h A B T LA 2 4 2 4 4 A Ai l 4 4 2 22 22 N 1 R .l C O 0 I 1 f

_ C O A

E

_ R

- s t t k u u

. c p p r o nrn k s l

I oI e c r r b o e e 0 n n m l k t e 3 n n a r a

._ n g6 r 1 1 o- o o o h e e

- - r r r r C t r I n- o7P P t t aP t - u u t

u t s n B mR cP e e u e . r I e . , a e e0 s3 e s

_. N N N N1 l 1 k d s t ex e ,

- u- a n a st u Rk e e e yal c g g eP pP e a p

_ Sif d

ro n8 n9 g0 n1 g

nt I m ,

B r

p B y

pen el wB

a. a- a- ao e i RP RP RP RN er p r p imo o nu T

I rrr T et P s p

r ex r

ex

'rex ,

r ex is i

r I i r

N t u wi wu wu bs T c I U rne or ol ol wu wu re i oI N LT Pf PF ol ol ur r t r L t~ Pf Pf TP o a o

_ A c t mc t N a c oi c O e . . a t g a I

Ra b c d

e f

g

. e uo e _

T R At R _

C i

. t U '8 7 '98 r.

t9 F 11 J1 J' Xoz -

n> " ' Ea"*"m "

R$ ".'

5g P u(J 2 ?l. + C ggk$ :P u,I

O ALITT- l TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 4 - With the number of OPERABLE channels one less tnan the Minimum Channels OPERABLE requirement, suspend all operations invoiving

+ positive reactivity changes.

ACTION 5 - Delete ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Speci fication 4. 3.1.1.

ACTION 7 - Delete ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive status light (s) that the interlock is in its required state for the exi. sting plant condition,. or apply Specification 3.0.3.

ACTION 9 - With the number of 0FERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel aay be bypassed for up to '

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the other channel. is OPERABLE.

ACTION 10 -'With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor trip breakers within the- next hour.

ACTION 11 - With the number of OPERABLE rhannels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 12- With one of the diverse trip features (Undervoltage or shunt j trip attachment) inoperable, restore it to OPERABLE status within :

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 9. i The breaker shall not be bypassed while one of the diverse trio features is inoperable except-fcr the time required for perform-ing maintenance to restore the breaker to OPERABLE status. With !

the breaker bypassed, apply ACTION 9. I ACTION 13- With any reactor trip bypass breaker inocerable, restore the bypass breaker to OPERABLE status prior to placing it in service.

CATAWBA - UNITS 1 & 2 3/443-6 -Amendment-No4HUnit 1)

-Amendment-NesMUn+t 2)

up.r T l I

i 1ABLE 4.3-1 9

REACIOR TRIP SYSitM INSIRUMENIATION SURVElitANCE REQUIREMENTS

" TRIP

ANALOG ACIUATING MODES FOR E CilANNEL DEVICE WilICH .

Z CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

[

I. Manuai Reactor Trip fl . A . N.A. N. A. R(14) N.A. 1, 2, 3 * , 4 ^ , 5

l

~

, 2. Power Range, Neutron Flux !

a. NigF Setpoint S D(2, 4), M N.A. N.A. 1, 2 M(3, 4), r Q(4, 6),

R(4, 5)

b. Low Setpoint S R(4) M N.A. N.A. l###, 2 i w

5 3. Power Range, Neutron Flux,- N.A. R(4) M N.A. N.A. 1, 2

'O lig"h Positive Rate - --

a --

^

Power Range,. Neutron Flux, N.A. N.A. N.A.

4. R(4) M 1, 2 liigh Negative Rate

4. Intermediate Range, 'S R(4, 5) S/U(1),M N.A. N.A. l###, 2 j,7 4 Neutron flux

-@ 6. Source Range, Neutron flux 5 R(4, 5) S/U(1),M(9) N.A. N.A. 2##, 3, 4, 5 3a 5

' [?

o8

~'

,{. Overtemperature AT S R M. N.A. N.A. 1, 2

)". Overpower AT S R H N.A. H.A. 1, 2 un m "f !

/g Pressurizer Pressure-Low 5 R H N.A. N.A. I 22 }d.

9 Pressarizer Pressure-liigh 5 R M N.A N.A. 1, 2

- 3. 5.

(( sf. Pressurizer Water Level-liigh 5 R M N.A. N.A. I ve

Id. Reactor Coolant iIow-low $ R H N.A. N.A. 1 It i

i r ., -- - - _ _ - - _ - . _ - _ _ _ -

varT I I'ABLE 4.3-1 (Continued)

Y REACTOR IRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP

' MODES FOR ANALDG ACIUATING E CNANNEL DEVICE WillCil _

Z Ci!Ai4NEL CilANNEL OPERAI10NAL OPERA 110NAL ACTUAI10N SURVEltiANCE IUNCTIONAL UNIT CllECK CALIBRATION EST TEST LOGIC TEST IS REQUIRED

[

[ . Uf. Reactor Ir ip System Interlocks (Continued) r7

e. Power Range Neutron Flux, P-10 N.A. R(4) M(8) N.A. N.A. I

f. Power' Range Neutron Flux, Not P-10 N.A. R(4) M(8) N.A. N.A. 1, 2

.w D g. Turbine Impulse Chamber i w Pressure, P-13 ft A. R M(8) N.A. N.A. 1 J".f Reactor Trip' Breaker N.A. N.A. N.A. M(7, 11) N.A. 1, 2, 3*, 4*, 5^

18 '

26. Automatic Irip and Interlock N.A. .N.A. N.A. N.A. M(7) 1, 2, 3^, 4*, 5*

19 togic hg ?g. Reactor Trip' Bypass H. A. N.A. N.A M(7,15)R(16) N.A 1,2.,3*,4*,$*l gg 20 Breakers E O-83 TE a.

,1

4.

s.

t'+

1 i.

~

U pJ.rT L i

TABLE 4.3-1 (Continued)

TABLE NOTATIONS <

Only if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.

Above P-9 (Reactor Trip on Turoine Trip Interlock) Setpoint.

1. Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous 7 days.

l (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel. gains consistent with calorimetric power if absolute difference is greater than 2%. .The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(3) Single point comparison of incere to excore axial flux difference above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are

not applicable for entry into MODE 2 or 1.

1

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector plateau curves shall be obtained, evaluated and compared to manufacturer's data. For the Intermediate Range and-Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable j for entry into MODE 2 or 1.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into

- MODE 2 or 1.

l (7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

1 (8) With power greater than or equal to the interlock setpoint the required ANALOG CiiANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state'by observing the permissive status light, i (9) Monthly surveillance in MODES 3*, 4*, and 5* shall also include verifi-cation tnat permissives P-6 and P-10 are in their required state for

{

existing plant conditions by observation of the permissive statu light, i (10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verifi-cation of the operability of' the Undervoltage and Shunt trips.

(12) Deleted (13) For Unit 1, CHANNEL CALIBRATION shall ensure that the filter time constant associated with Steam Generator Water Level Low-Low is adjusted to a value less than or equal to 1.5 seconds.

(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY o' the Bypass Breaker trip circuit (s).

I CATAWBA - UNITS 1 & 2 3/4A3-12 Amendment -No-63-(-Unt 1)

-Amendment-No. 57 (UnM-2+--

U d.I T Z.

I Allt t 3. 3- 1 (Cont inued) .

$ RlACIOR IRIP SYSIEH lilSIRtiMENIA110N

.c m ;

" MINIMtiM

' CilANNELS CilANNELS APPLICABLE l' TAL NO.

.,f Cl!ANNEtS TO TRIP OPERABLE MODES ACTION E filNCfl0f1AL UNII

18. Reactor Trip System Interlocks

a. ' Intermediate Range 8 2 2##

-" Neutron flux, P-6 2 1

~

b. Low Power Reactor Irips Block, P-7 3 8 P-10. Input 4 2 1 or 8 2 1 P-13 Input 2 1

c. Power Range Neutron 8 4 2 3 1 R flux, P-8 di 2 3 1 8 Power Range Neutron 4 Y d.

Flux, P-9

e. Power Range Neutron 8' 4 2 3 1 flux, P-10 L.> f. Power Range Neutron 1, 2 8 4 3 4 B!! flux, Not P-10

$eh g. 'lurbine Impulse Chamber 2 1 8

$y Pressure, P-13 .2 1 kx 2 1, 2 9, 12

19. Reactor Trip Breakers 2 1 l f} 2 1 2 3 ^ , 4 * , $^ 10

. Lg 2 1, 2 9

'h 20. ' Automatic Irip and Interlock 2 1 3^,4*,5* 10 2

11 no logic 2 1 N.A. ii . A . 1, 2 , 3 ^ , 4 ^ , 5

13

!3ip 21. . Reactor Irip flypass Breakers ti . A .

v M.I'r 2.

TABLE 3.3-1 (Continued)

ACTICN STATEMENTS (Continueo)

ACTION 4 - With the number of OPERABLE channels one less tnan the Minimum i Channels OPERABLE requireinent, suspend all operations involving positive reactivity changes.

ACTION 5 - Delete ACTION G - With the number of OPERABLE channels one less than the Total Numoer of Channels, STARTUP and/or POWER OPERATION may proceeo orovided the following conditions are satisfied:

a, The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 4

b. The Minimum Channels OPERABLE requirement is met; however.

the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other cnannels per l

Speci fication 4. 3.1.1.

F ACTION 7 - Delete ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive status light (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the otner cnannel is OPERABLE.

i ACTION 10 - With the numoer of OPERABLE channels one less than the Minimum Channels OPERABLE reouire.nent, restore tne inoperaole channel to GPERABLE status within 48 hou.s or ocen the Reactor trip

. breakers within the next hour.

ACTION 11 - With tt.e numoer of OPERABLE channels less than the Total Numoer of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ,

ACTION 12- With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperaole and apply ACTION 9. ,

The breaker snall not be bypassed while one of the diverse trio j features is inoperable except for the time recaired for perform- l ing maintenar.ce co restore the breaker to OPERABLE status. Witn i the breaker bypassed, apply ACTION 9. ! ;

I ACTION 13- With any reactor trip bypass breater inoperable, restore the l

bypass breaker to OPERABLE status prior to placing it in service.

CATAWBA - UNITS 1 &_2 3/4B3-6 Amencment-Nm634Unitr 1) --

-4enement-Nos?-(hit- 2 )

-~- -- - - _ - - -

- !' l ! l' l

' Y .

E

5 _

CD 4 -

NE , -

R AR , 4 O ^

z F i I l

l U

Q 3 2 2 3

SHEE , , ,

T ECVR 2 2 # 2 2 # ,

2 2 2 -

Dl R # # #

I J

l OiUS , ,

e MWSI 1 1 l 1 1 l 2 1 1 I 1 1 1 v I

_ NS

_ uE . . . .

l I . . . . . . . .

l A A A A A A. A A A A AC . A. . . . . A. A. . . .

. ll I N N N l f N N. N N. N N N N N IG .

. CO AL L

_ S A )

I G N 4 . . . . . . . . . . . .

N N 0 1 A A A A A A A A A A A A E l l ( . . . . . . .

M LEI R N N. N N. N N N. N N N N N '

E ACA R Pl IRT l

l I iVES i

t RCEPE Q I ADOT E

R )

9 L (

Ci L M, M, f A A N ) )

_ l l t I O

1

(

1

(

l GET A U U

. E ONA .

/ /

V L NRT N M M M M S S H H H H H H R AAES U NiPEl 1

S ACOT

_ - N 3 0

. 1 , ' , ,

4 T N )))) ) )

A O 4465 5 5 E T I

_ L N T . , . , ,) ) ) , ,

B E LA A 23444 4 4 4 4 A M ER . ((((( ( ( ( (

1 U NB N DMQRR R R R R R R R R R R R NI

_ I AL S l iA

_ N CC I

M t i i

L . . .

S NK A A A Y . NC . . .

_. S AL l 1 U S S N H S S S S 5 S 'a S i 1 P CC

. I

. R e I

lg R , ,

x i 0 x x x u h i l

1 u u u l w g - w

. C l l l f o i l o

A F f F L i l e l E n - - v -

R p

n n n o e e e w o o o , r r r l o

. i r re re e t l u u l r tt tt tt g u a s s r I T un t ua ua n e s s e ei n eR eR a N e e e i t r No i N N R r r r a n o p o ,ev e ,

u P P W a t ,t p ,v ex e t l l c ee t ei ei t u g a A r r r o T a gS e gt gt al n r e e e o I e n S ni na if a e r z z t C N R ah as ag d R p e i i i

. I L R g w R o R e en m w r r r r L

l a rl i

i L

o r

P r

N mo rr e

c t e o p

u s

u s t o

A u e eh eh et r r r s s s c N n w wg wg t u u e e e e e a e

O a o oi ne o v v r r r Poii I

M P a b Pi l l I N S O O P P P R I

C N . .

u . . . . . . 0 1 2

8.

f !

2 3 4 5 6 7 9 1 1 1 n>{ 5'EZ"' [, w2yw ,sg@${n i ' >" 21 g.

f.h${e ~

n" 2 1 3 {_

, ; i ' J .:I [! , .

varT L lABLE 4.3-1 (Continued) .

> i

> REACIOR 1 RIP SYSIEM INSIRUMENIAfl0ft SURVElllAriCE REQUIREMENTS

" IRIP ANA10G ACIUAllt4G MollES 10R

[

e CilANNEL DEVICE WillCal ^

CilAtit1El CilANilEL OPERATI0tlAL OPERAllONAL ACTUAI10N S!!RVElll ANCE 3 I UtlCII0tlAL UtlII ClifCK CAtlBRAll0N TEST TEST LOGIC TESI (S REQUIRED g

[ 10. Reactor Irip System Interlocks (Continued) ,

e. Power Range Neutron flux, P-10 N.A. R(4) M(8) N.A. ff A. I

! Power Range fleutron flux,tiot P-10 ti. A. R(4) M(8) ti . A. N.A. 1, 2 :

w Turbine Impulse Chamber hy 9 Pressure, P-13 ft. A. R M(8) ti. A. N . A .. I !

Reactor Irip Breaker N.A. fl. A. N.A. M(7, 11) N.A. 1, 2, 3^, 4^, 5^

19.

it. A. N.A. H.A. N.A. M(7) 1, 2, 3*, 4^, 5*

20. Automatic Irip and Interlock Logic

- fg h 21. Reactor Irip Bypass 11. A. fl. A. N.A M(7,lS)R(16) N.A 1,.2,3*,4^,5^l ,

(; 3reakers 4

=$

L,- l F

- .C .C

'32 Sh cu

v u.I T 2.

t TABLE 4.3-1 (Continueo)

TABLE NOTATIONS Only if the Reactor Trip System breakers happen to be cloud and the i Con;rol Rod Drive System is capable of rod withdrawal.

Above P-9 (Reactor Trip on Turoine Trip Interlock) Setpoint.

1. Pelow P-6 (Interm L iate Range Neutron Flux Interlock) Setpoint.

1. 1. Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous 7 days.

(2) Comparison of calorimetric to excore power iadication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicaole for entry into MODE 2 or 1.

(3) Single point comparison of incore to excore axial flux difference above 15% of RATED THERMAL POWER. Recalibrata if the absolute difference is l greater than or eaual to 3%. The provisions of Specification 4.0.4 are not acolicar le for entry into . MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector p!ateau curves shall be obtained, evaluated and compared to manufacturer's data. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicaole

for entry into MODE 2 or 1.

I (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. Tne

. provisions of Specification 4.0.4 are not applicable for entry into

MODE 2 or 1.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(8) Witn power greater than or equal to the interlock setpoint the requireo ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive status lignt.

! (9) Monthly surveillance in MODES 3*, 4*, and 5* shall also include ver1ri-cation that permissives P-6 and P-10 are in their required state for

existing plant conditions by observation of the permissive status light.

(10) Setpoint verification is not applicable. l (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verifi-cation of the operability of the Undervoltage- and Shunt trips.

(12) Deleted (13) For Unit 1, CHANNEL CALIBRATION shall ensure that the filter time constant associated with Stea;n Generator Water Level Low-Low is adjusted to a value less than or equal to 1.5 seconds.

(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall:inoependently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual _ ;

- Reactor Trio Function. The test shall also verify the OPERABILITY of tne Bypass Breaker trip circuit (s).

. I 1 CATAWBA - UNITS 1 & 2 3/4 83-12 -Amendment-No-61-(-Unit-&

-Amenoment-No. m tun +t--2)

05'ZO-1 ?M Cl i l SP*n r Pon DP'C ff4LCFv; EPClf CCRif0 TO 55313151 p,cy UNIT ; l

., 1 20VER DISTRIEUTI:'i L:u: ;

)

LIMITING CONDIT!!N 09 CRE:aTION ACTION (Continueo' e" Identi. ace correct the cause of the Out-c'*!imi: ::ncition prior 359AL PCWER aoove tne recuca: THEFAL POWER limit

+o i-- ear:yICT:CN require: a. and/or c.2. , aseve; su=secuent POWER OPERA-TION may :r:cesc :revided that F (X,Y) is cemonstr2ted thr:U:P j ,,::r ; ' , .n.,...,, ; ce witnin ne :Imit 3:ecifieC la 188 EE prior :: excetoi G the following THEEFAL POWER levels:

1) 50% of RATED THERMAL POWER,

2) ~5% of RATED THERMAL F0VER, and .

I

3) Vitnin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% 1 of RATED THERMAL POWER.

SURVEILLANCE REOU!RE.uENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 Fd(X,Y) snall be evaluated to catemine wnether F '

its limit by: g (X,Y) is within

a. Using the covaole incere detectors to obtain a power distribution j sap at any THERMAL POWER greater than E: of RATED THERMAL POWER. !

I

b. Measurin; Ftu[(X,Y) according to the following schedule: !

gE st3 1. Prior :: operation above 75% of RATED THERMAL POWER at the, t AnnM3 "i ectnning cf eaca fuel cycle. anc the earlier of:

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

3. At eacn time the QUADRANT POWER TILT RATIO inoicated by the ;

excore catacurs is normalizac using incere detector measurements. j

c. Derforming the following calculations: '

1. For each location, calculate the % margin en the maximum l allowanle casign as follows: i I

%F Margin = 1 - ( ' )I *1 C ff%(X,Y$

' / p *4 .-

> t_. a ,

i

. A % b , ,' No acdtt onal uncitta itias ara required for F X,Y).

'N J because ag(X,Y)M cludes uncenainties

- ~

e,.M Ab 4 3

\ _ , TAVBa - UNITS 1 & 2 3/4 A2-8 3'33 henWM-(Ud U i i

^ tr. = n- W . 00 (UM M b

05-20-1***J2 01315f*f1 P"ROr1 DP"C NUCLEAR. ENDifEEMIND TO 55313151 P.03 t.

for Specification 4.2.3.1 Attachment 1:

1. Upon reaching ecIuilibrium ccnditions af ter exceeding by 10%

or more of PATED THEPXAL POWER, the THEPRAL POWER at which FL (X, Y) was; last determinedl J) , or Attachment 2:

where (FAHb (.v., Y) ] SURV is the design peaking limit defined in the COLR.

Attachment 3:

(3) During power escalation at the beginning of each cycle, "J/ Erit \L POWER may be increased until a power level for extended operativn has been achieved and a power distribution map obtained.

_____.__________-_____.-_-_m_. -

. nti

- . i 4

4

DOWER DISTRIBUTI*.N LIMITS LIMITING CCNDIT:CN FOR OPERATION ACTION (Continueci

2. Fino the minimum margin of all locations examined in 4.2.3.2.c.;

a:cve. f any margin is less than zero, comoly.with the ACTICN ;

s e $< , ,

l

[d. ExtTacciating the two most. recenT. measurements to 31 Effective _ full ,

I Power Cays Oeyond the most recent measurement and if: ~

Fahr" (extracolated) 3 FaHRb (extrapolated) eitner' tf the followi.ng actions shall be taken:

u ,,

1. FAHR'V.V) shall be increased by 2 percent over that specifiea in 4.2.3.2.a anc the calculations of 4.2.3.2.c receatec, or ,

2. A mova01w incere detector power distribution mao shall be cotained, and the calculations of 4.2.3.2.c shall be .cerformed no later than the time at which the margin in 4.2.3.2.c is :

extrapolated to be equal-to zero.

l ACF A-Nachsend l CATAWBA - UNITS 1 & 2 3/4 A2-9 -A***d**"t D 0 5 (Unit II~

8 35 I

?::nt.:nt Mc. 80 (Unit 2)

05-25-1572 0131 ?PT1 f"Pon Dr"C NUCLCm EfolfCER!to TO 55313151 P.04 for Specification 4.2.3.2 Attachment 1:

d. Extrapolating (4) at least two measurements to 31 Effective Full Power Days beyond the m0st recent measurement and if:

FL (X, Y) (extrapolated) 2 (FL (X,Y) )"" (extrapolated), and F".(X.Y) (extrapolated) #

F",, ( X , Y ) '

( FL (X, Y ) ) " (extrapolated) ( FL (X,Y) ) ***

either of the following actions shall be taken

1. FL (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 detector power distribution nap shall be obtained, and the calculations of 4.2.3.2.c shall be performed no later than the time at which the margin in 4.2.3.2.c is extrapolated to be equal to zero.

(4) Extrapolation of FL f or the initial flux map taken af ter reaching equilibrium conditions is not-required since the initial flux map establishes the baseline measurement *or ' ,ure trending.

i

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

uurT \

TABLE 3.3-4 (Continuedl ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS r

hMe 10TAL SENSOR ERROR

}

si; FUNCTIONAL UNIT I

ALLOWANCE (TA) Z_ (5) TRIP SETPOINT ALLOWABLE VALUE

= .

d 18. Engineered Safety Features

- Actuation System Interlocks e

m a. Pressur Ner Pressure, P-11 N.A. N.A. N.A. 1955 psig >1944 psig

b. Pressurizer Pressure, not P-11 N.A. N.A. N.A. 1955 psig <1966 psig

c. Low-Low T , P-12 N.A. N.A. N.A. >SS3 F ESSIF(SSOT\

d. Reactor Trip, P-4 A. N.A. _

N.A. N.A. N.A.

e. Steam Generator Level, P-14 See Item S. above for all Steam Generator Water Level Trip Setpoints w and Allowable Values.

N B

$s*

=a ia

Lg3 3 3II.

b i

'TY2

,.A

4. *.

O j

l> AJ ~T Z .

s TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION IRIP SETPOINTS 2-h\ek-e. TOTAL SENSOR EkR0R e

(S) TRIP SETPOINT ALLOWABLE VALUE ALLOWANCE (TA) Z g FUNCTIONAL UNIT .

bI 18. Engineered Safety features j ra Actuation System Interlocks e ( N.A. 1955 psig 11944 psig Pressurizer Pressure, P-11 N.A. N.A.

na a. ! .

N.A. N.A. 1955 psig $1966 psig

b. Pressurizer Pressure, not P-11 N.A.

c. N.A. N.A. N.A. 3553 F 1654Aff550 F Ks Low-Low Tavg, P-12 N.A. N.A. N.A. N.A. N.A.

d. Reactor Trip, P-4 Steam Generator Level, P-14 See Item 5. above for all Steam Generator Water Level Trip Setpoints e.

us and Allowable Values.

s.

T l U 66

$$a I

22 w.

)

U cI

1912 l 3. S.

! +[

1

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION

~

3.5.1 Eacn cold leg injection accumulator shall be OPERABLE with: l

a. The discharge isolation valve open,

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

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.

APPLICABILITY: MODES 1, 2, and 3*

ACTION:

a. With one cold leg injection accumulator inoperable, except as a result of a closed isolation valve or boron concentration less than 1900 ppm, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the following 6 bours,

b. With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

1) Jha ,yol e weighted average boron concentration of the .-

4-imiting accumulators 1900 ppm or greater, restore the n perable a tor to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the low boron determination or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System pressura to less than l 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e weighted average boron concentration of.th three -

2) ,.Jfwmrtry=jaccumulators l i,..i ti o less than 1900 ppm but greater t 16og1500'p . , restore the inoperable accumulator to OPERABLE status n' ' e weighted average boron concentration of th three limiting)ccumulators to greater than 1900 ppm and

Reactor Coolant System pressure above 1000 psig.

CATAWBA - UNITS 1 & 2 3/4 5-1 -Amendment-Nor-864 Uni-t-1F

-Amendment "c. G0 (Uni t 2'

EMERGENCY CORE COOLING SYSTEMS l

LIMITING CONDITICN FOR OPERATION (Continued)

ACTION: (Continued) enter ACTION c.1 witnin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of the low boron determination or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System pressure to 1 sf D . 00 psig within the fol-lowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 180 0

.3 ) ,T W 4e weighted a rage o on ?nce tration of t tt :

'i-iting ccumulator. 1:^ m, pmation concent or l esfs the

, re f tunn t"c: mo,i

'S ti n-weigrytad average boro Mc'5mulators to greater than ': ^

m and e. 'e A within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the low om n-edermination or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant j System pressure to less than 1000 psig within the following

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMENTS 4.5.1 Each cold leg injection accum. ator shall be demonstrated OPERABLE: l

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution

! volume increase of greater than or equal to 75 gallons by verifying the boron concentration of the accumulator solution;

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

is above 2000 psig by verifying that power-is removed from the

, isolation valve operators on Valves NI54A, NI653, NI76A',' and NISSB and that the respective circuit breakers are padlocked; and f d. At least once per 18 months by verifying that each cold leg

injection accumulator isolation valve opens automatically under 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.

I

- This surveillance need not be performed until prior to entering HOT STAN08Y following the Unit I refueling, r .

CATAWBA - UNITS 1 & 2 3/4 5-2 -Amendment-No,-86-(-Uni t- }-

-Amendment-Nor80ittnTt4P

t 3/4.5 EMERGENCYCORECOOLINGSYSTF3 j BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System accumulator ensures that a sufficient volume of borated water will be immediately forced into the

reactor core through each of the cold legs from the cold leg injection

! accumulators and directly into the reactor vessel from the upper head injection

, accumulators in the event the Reactor Coolant System pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism durir.g large pipe ruptures.

, The limits on accumulator vplurqep otonh to %entration and 9ressure ensure i that the assumptions used for ytcumufator njection in the safety analysis are met. f IPDO Thealloweddowntim/fo e accumt ators are variable based upon boron

. concentration to ensure ha / h reactpf is shutdown following a LOCA and that l any problems are corre ed n 4 tim 1 manner. Subcriticality is assured when J

bor3n concentration is abo e.1600 m, so additional down time is allowed when l con:entration is abov J.5 ppm. concentration of less than 1900 ppm in any single accumulator or als a v e weighted average may be indicative of a pro-blem, such as valve leakaga, but since reactor shutdown is assured, additional

time is allowed to restore boron concentration in the accumulators.

4 The accumulator power operated isolation valves are considered to be a ' operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever o m issive 4

conditions are not met. In addition, as these accumulator isolation *,ai"es fail to meet single failure criteria, removal of pcwer to the vsives is required.

I The limits for operation with an accumulator inoperable for any reason I

except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concur ent with failure of an additional accumulator 1

which may result it unacceptaale peak cladding temperatures. If a closed isolation valve cumot be irediately opened, the full capability of one accumulator is not available and prompt action is required to placs the reactor in a mode where this capability is not . required.

~

T \ f Theor"*nallicgnsingbasisfort Catawba Nuclear Station assumes b th the i UHI Sy nem and Co(d Leg Accumulators function to miti te postulated ac idents, t

~ bsequent analyseh nave demonstrated that the UHI Sys tm is not require

! pr vided minor chan s to the Cold leg ccumulator paranl ters and dischar e pat are implemente . Accordingly, Spe ification 3/4.5. has been modifi d to a ress two 'possibi plant configurati ns:

(1 UHI Operable (0 iginal plant conf uration) i (2) HI Disconnected HI penetrations the Reactor Ve el are cut and l 5 i

pped, " ld Leg Ac umulator level a d cover pres-su ct 'd, Cold L Acenulewr fl w restricting J

j

[f ori ce laced)

\ s EMTE kv 4 TAWBA /

- UNITS 1 & 2 B 3/4 5-1 Amendment-No.-324 Unit-it

- Amendment-Nor2 MUni t-2t-

4 , .

CG dt".cd F % rs Catanba Nuclear Station Appendis 15. Chapter 15 Tsbla and 11gures J

) 0 Table 15 26 (Page I of 2). Parameten for Pmtulated Rod Ejection Offsite Dose Analpis - _ - _

Conscrsathe Healistic j

l 1. Data and assumptions used te estimwe s radioactive source from postulated accidents i a. Power level (MWt) 3565. 3565.

t l

i

.h.---peecent ofw rs u2 <

1

c. Steam generator tube leak rate prior to end- X o,G - 4Het---

hmag :'=&mp fr,r.* Aceipyrg

d. Failed fuel )4fpercent of fuel same i rods in core

~

A, t,.t, te p yrN iN-FVt 4 f'lt4

e. Activity released to reactor coolant from failed M r'(A p u TV 6 i D *l* OF '

4 fuel and available for release ptra, friN g NVE.N it f CV- 5'O'*[=

Noble gavs 40 percent-oft 5te- s.une CF M ' '

i .-gapinventory. - @ d6 i lodines 40parcent of core same

-tap inventory-4--Mehed fuel -

- - - - - -- -0.25 percent-of -tr~

l 1

core-j .gm _. Activity reicased to rer.ctorcoolant from melted .--

fuel and available for releasero.contamment-

! ---Noble Fases

- - 0.25. percent.of L-

! Core inventory I

lodines -- - _ M25 percent of -0.--

' -- core inventorys

h. lodirr Fractions (organic, elemental, and Regulatory Guide same particulate) 1.4

2. Data and assumptions used to estiraate acthity released Containment Free volume (ft$ ) 1.015E + 06 :.ame l u g

b. Consdnrnent leak rate 0.3 percent of Ogpercent of l contatnment containment l

! volume per day, volume per day, Osts24 hr Osts24 hr O.15 percent of 0.025 pc cent of containment contairment volume per day, volume per day, I t > 24 hr t ) 24 hr i

c. Bypass leakage fraction 0.07 0.07 0 dr-lodine partition factor for steam Telene .01 -- - - - - - - < - -

c. Offsite power Lost - >

l i

(01 OCT 1991)

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

ML20114B676

Text

Navigation menu

Search