ML20065B836

Proposed Tech Specs Pages for Missing Page Numbers of TS Pages Submitted in

ML20065B836
Person / Time
Site:	Summer
Issue date:	03/25/1994
From:	SOUTH CAROLINA ELECTRIC & GAS CO.
To:
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ML20065B834	List: ML20065B836 RC-94-0089, Forwards Revised Pages for Purpose of Having Final TS Pages Included on Docket,Due to Discovery of Missing Page Numbers on TS Pages Submitted in Re TS Change TSP 930002 Concerning Limiting Safety Sys Settings
References
NUDOCS 9404040153
Download: ML20065B836 (34)

Text

.

l SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEN INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation and interlocks setooints shall be consistent with the Trip setooint values snown in Table 2.2-1.

APPLICABILITY: As snown for eacn enannel in Table 3.3-1.

ACTION:

With a reactor trip system instrumentation or interlock setooint less a.

conservative than the value snown in the Trip Setooint column of Table 2.2-1 aajust the setpoint consistent with the Trip Setpoint value.

b.

With the reactor trio system instrumentation or interlock setooint less conservative tnan the value emnwn in the Allowanle Salues column c+

Table 2.2-1,r7p ace tne cnannei in tne trippeo conoition witnin 1 nour,]

r ano witnin Lne following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> either:

1.f mw n..

t Eouation 2.2-1 was satisfied for the affect annel and adjust tn tpoint consistent with the Trip Setoo alue of I

Table 2.2-1, or 2.

Declare tne cnannel ino aole and apply aplicaole ACTION statement reouirement of ification

.1 until the cnannel is restoreo to OPERABLE status h it tooint adjustea consistent

]j with the Trip Setooint value.

l' l

EQUATION 2.2-1

+ R + 5 < TA I wnere:

I Z=

the value for col

• Z of Table 2.2-1 for *he affectea cnannel, i

R=

the "as meas d" value (in percent span) of ck error for the affectea c

nel, i

S=

eith, he "as measured" value (in percent span) o he sensor er

, or the value is column 5 of Table 2.2-1 for t affected i

nnel, ana.

I T

the value from column TA of Table 2.2-1 for the affected ch el.

Y g

w

-u-u n

c b f,g,; ag,, n sta%utnequiume.,6f & rah wa it a e;g4'fgL -

j is nsforq #sans s wdh & 9 stps Que.

ss g q SUPNER - UNIT 1 2-4 9404040153 940325 PDR ADOCK 05000395 p

PDR

s SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation and interlocks setpoints shall be consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

a.

With a reactor trip system instrumentation or interlock setpoint less conservative than the value shown in the Trip Setpoint column of Table 2.2-1 adjust the setpoint consistent with the Trip Setpoint value.

b.

With the reactor trip system instrumentation or interlock setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirements of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint Value.

1 i

SUMMER - UNIT 1 2-4 Amendment No.

L 4

I Ai!!! 2.2- !

!!! A! H@ !hlf hMI[H lil5fHilitiftIAllgH lith' hil[plH!$

y, c-m R

functional linit I

5 isIp 5etpoint Allowalele Value A losi

- (IA)

_b l

1.

Hanual Heactor Trip

(

Hot}pplicable N[

HA HA NA a

2.

Power Range, Heutron flux

~,

111 h Sepoint

/.S 4.$6 u

lo'.ax of HIP 111.2% of HIP 9

tow Letpoint

.3 4.56j 0

t

$25% of itTP

$21.2% of f(TP 3.

Power Range, Heutron F' =

j 1.

D.r 0

) :5% of RTP with a time 6

2 3% of BIP with a time i

liigh Posit Ive Rate 5

( con ant 22 seconds const. int 32 seconils 4.

Power Range, Neutron flux 1.6

.5 0

55% of RIP with-a time

$.3% of RIP with a time 6

y liigh Negative Rate constant 1 2 seconds constant 32 seconls m

'5.

Intermediate Range, 11.0 8.4 0

l $25% of RIP

$31% of RIP Neutron flux

\\

)

6.

Source Range. Heutron flux {

17.0/- kl0.0

[

<105 0

5 cps

$1.4 x 10 cps 7.

Overtemperature al

/

10.3

.8 1.6 l See note I See note 2 l.

& l.2'(

8.

Overpower ai

/

5/

1.k6 1.6 See note 3 See note 4 g'

g 9.

Pressurizer Pressure-Low

(,

jf. I

0. 7h 1.5 1

3 810 psig 31859 psig j

10.

Pressurizer Pressure-High f

/3.1

0. 7 I '

l.5

$2380 psig

$2391 psig g- -

11. Pressurizer Water Level-liigh 5.0 2.18 1.5

$92% of instrument

$93.8% of instrument 7.

span span

12. - Loss of flow

{

~4, 30%ofloopdesign 188.9% of loop design 2.5 1.48 flow a-flo.*

u m

• toop design flow - 94,870 gpm ATP - A4IED filEA l

1

-+4. = :;=. = : =HAL POWEA O

g

p ::,;.,_ l.a co, r...... ; m,

, cm. a -

i

_,m

~

TABLE 2.2-1 g

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

o i

Functional Unit Trip Setpoint Allowable Value E

4 1.

Manual Reactor Trip NA NA 2.

Power Range, Neutron Flux High Setpoint

$109% of RTP

$111.2% of RTP Low Setpoint

$25% of RTP

$27.2% of RTP 3.

Power Range, Neutron Flux 55% of RTP with a time 56.3% of RTP with a time High Positive Rate constant 22 seconds constant 22 seconds 4.

Power Range, Neutron Flux 55% of RTP with a time 5 3% of RTP with a time 6

High Negative Rate constant 12 seconds constant 22 seconds 5.

Intermediate Range, Neutron Flux 525% of RTP

$31% of RTP 6.

Source Range, Neutron Flux 5105 cps

$1.4 x 10 cps 5

7.

Overtemperature AT See note 1 See note 2 8.

Overpower AT See note 3 See note 4 9.

Pressurizer Pressure-Low 21870 psig 21859 psig 10.

Pressurizer Pressure-High 52380 psig 52391 psig 11.

Pressurizer Water Level-High 592% of instrument span 593.8% of instrument span yy 12.

Loss of Flow 190% of loop design flow

• 188.9% of loop design flow *

'I @

$7 1a

• Loop design flow = 94,870 gpm yg RTP - RATED THERMAL POWER l

.i

c

~

IABLE 2.2-1 (continued) 50 REACIOR 1 RIP 5YSTEM INSTRIMENTATION TRIP SETPOINIS Functional Unit A

ance M p Setpoint Allowable Value 13.

Steam Generator Water 1.0 tevel low-low M'

6-

>l2% of span free

>l9.2% of span from

11. 2 y

7'7/

5 to 30% RTP B to 30% RIP increasing Iln-locreasing linearly early to >30.0% of to >23rE; of span i

span froe 301 to j

100% RTP train 301]to100%

14.

Steam /feedwater flow His-RIP p ggg qi 16.0 Match Coincident With 13 24 1.5/

<40% of full

<42.5% of full

1. 5 iteam flow at RIP iteam flow at NIP Z

Steam Generator Water tevel tow-tow 12.0 st.2

>l2% of span from

>lDr2% of span from g

j 7 ) 5 to 30% tr>

8 to 30% RIP lacreasing 11n-fucreasing linearly early to 130.0E of to 13Seft of span span from 30% to from 3 o 100%

1901 AIP

15. Undervoltage - Reactor RIP 2.1 gg Coelant Pamp

.28 0.23

>4830 volts 16.

Underfrequency - Reactor

~>4760 4

~

i

7. 5 Coolant Pumps 0

0.1 157.5 lit 157.1 liz

[

17. Turbine Irlp 5

A.

Low Irlp System Pressure

' g 8.

Turbine stop Valve MA NA 1800 psig 1750 psig M

(A Closure i

NA NA 11% open 11% open g-sa e e a t.e m Puuq 8 m

e e

TABLE 2.2-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 2

9 Functional Unit Trip Setpoint Allowable Value U

13.

Steam Generator Water 212% of span from 0 to 30% RTP 211.2% of span from 0 to 30% RTP

~

Level Low-Low increasing linearly to 230.0% of increasing linearly to 229.2% of span from 30% to 100% RTP span from 30% to 100% RTP 14.

Steam /Feedwater Flow Mis-540% of full steam flow at RTP 542.5% of full steam flow at RTP Match Coincident With Steam Generator Water Level 212% of span from 0 to 30% RTP 211.2% of span from 0 to 30% RTP Low-Low increasing linearly to 230.0% of increasing linearly to 229.2% of span from 30% to 100% RTP span from 30% to 100% RTP 15.

Undervoltage - Reactor 24830 volts 24760 Coolant Pump m

i" 16.

Underfrequency - Reactor

>57.5 Hz

>57.1 Hz Coolant Pumps 17.

Turbine Trip A.

Low Trip System Pressure 1800 psig 2750 psig B.

Turbine Stop Valve Closure 21% open 21% open E

e i

8" RTP - RATED THERMAL POWER

?

.i,

r-i j'

TABIE 2.2-l_(continiteit) pyC10R 1 RIP SYSIffl lilSTHilflfillAlllll! IRIP SEIP0lHTS Total L

Functional linit 13 llonance { M}

hjgt31[gint Allowable Value Q

18.

Safety Injection Input il flA j lA flA fl4 from ESF 4

19.

Reactor Trip System.

Interlocks A.

Intermediate Range L NA liA flA

>/.S x 10 6%

>4.5 x 111 '%

Neutron Flux, P-6 i

Indication Indication l

D.

. low Power Reactor Irlps Block, P-7 P-ID input

7. 5 a.

> ~

.56 0

$10% of RIP 512.2% of HIP l

m b.

P-13 loput L

7. 5 4.56 0

<10% turbine

<l2.2% of lustilne i

w l

Impulse pressure Impulse pressure j

equivalent equivalent C.

Power Range Neutron 7.5 4.56 0

<30% of RTP

<10.2% at HIP Flux P-8 4

~

D.

tow Setpoint Power

7. 5 '

4.56 0

>10% of RTP

'21.8% of HIP Range Neutron Flux, P-10

~

L Turbine le l

E.

Pressure, pulse Chamber

<10% turbine W12.2% tuihine.

7.5 P-13 4 56 0

lepulse pressure jiressure equivalent -

equivalent

. 3(

F.

Power Range Neutron Flux..P-9

[

7.5 4.56 0

i

$50% of HIP

$52.2% of HIP il g" -

l20.

Reactor Trip Breakers HA~

NA na

. Hg

.21.-' Automatic-ActualIon logic

- HA g4 t

.Ng ng ng ggg k N D'lllERHAL POWER'.

.g

.l.

92

-- L-- i 2L.--

c

l TABLE 2.2-1 (continued) l FU CTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0INTS

$n j

Functional Unit Trip Setpoint Allowable Value I

5

]

18.

Safety Injection Input from ESF NA NA l

19.

Reactor Trip SyrJam Interlocks A. Intermediate Hege Neutron Flux, P-6 27.5 X 10 6% indication 24.5 X 10 6% indication B. Low Power Reertor Trips Block, P-7

a. P-10 input

$ 0% of RTP

$12.2% of RTP 1

b. P-13 input

$10% turbine

$12.2% of turbine impulse pressure equivalent impulse pressure equivalent

[

C. Power Range Neutron Flux P-8 538% of RTP

$40.2% of RTP D. Low Setpoint Power Range Neutron Flux, P-10 210% of RTP 27.8% of RTP E. Turbine Impulse Chamber

$10% turbine

$12.2% turbine Pressure, P-13 impulse pressure equivalent pressure equivalent 8

F. Power Range Neutron Flux, P-9 550% of RTP 552.2% of RTP I

g 20.

Reactor Trip Breakers NA NA n

21.

Automatic Actuation Logic NA NA 1

?

RTP - RATED THEP, MAL POWER

~

l A.l}l I. 2 : L! !!'!"I!!im."!l Bto!:19B 1B1LSYSIEt!.I!! SIB!! 4!!!S!!9!! !!!!t hl!!t!!!!>

t~.^q

!2 fj0] Al10!! (:: int inueil)

)

f40lE 1:

(Continued)

M a n.1 i, (al) is a tuottnon of Ihe inniltated ditte n e hetwero I op anil but t um ile t et. t us s of ti.e powei e amic nuclear ton chambers; with gains to be selected bas 2d on mea >iired leistruitent response dur ing plant starttip tests such that:

(1) for qt - qb between

.' I pei i ent a nil i 1 per. nt I,

t J. ! )

o t.hr e e qt

""I 4 " P " ' ' ' " I III' II'1'uI4' 1

POWiR in the top anil bot tom halves ni the nr-r es pri t i + c l >, anal qt is total till fillAl l'Old H tre

' 9h percent of RAlf D TilfidlAl P0ut R.

(ll) for each percent t isa t t he riiaijoi tude: ut qt it, ext ee.13 ci lien tent, the al til.

itpoisit s h a l l lo.-

i automatitally reduced by 2.27 percent of las salise at itAllti lilf RflAt POWtR.

'b (ill) for each percent that the magnitude of qt - 'l exceeds e-1 percent, the al trip setpoint shall be l

automat ically reduced by 2.34 percent of it s salue at RAilli liiffff4A1 P0HfH.

I 110I0 2:

lhe channel's maximum trip setpoint shall not exces.1 its compaleil trip point by more than 2.2 pen ent Span.

l 140lf 3

OVERPOWLH al

{ T.s S)

-V F-a AT s bT K-K T-K T-T g

6 (I i t,Si d

E Where: al as detined in riote I h i,,

as def ined in flote i

=

K.

1.0875 re

. e.

K 1

0.02/*t yi for increasing ave age temperature and 0 for decreasing average temperatust 5

• aS w3r lhe function generated by a

3 compensation

.he rate-laij controller for I" 9 dynamic 3+LS y

S2

TABLE 2.2-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

NOTATION (Continued) s C

gNOTE1:

(Continued) and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range 3

nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) for gt - 9b between -24 percent and +4 percent f, (AI) = 0 where gt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and gt + 4b is total THERMAL POWER in percent of RATED THERMAL POWER.

(ii) for each percent that the magnitude of qt - Ab exceeds -24 percent, the AT trip setpoint shall be automatically reduced by 2.27 percent of its value at RATED THERMAL POWER.

1 (iii) for each percent that the magnitude of qt - 4b exceeds +4 percent, the AT trip setpoint shall be automatically reduced by 2.34 percent of its value at RATED THERMAL POWER.

NOTE 2:

The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.2 percent AT Span.

NOTE 3:

OVERPOWER AT-k

~

(t S) 3 E.

ATsAT K-K T-K T-T o

4 5(1 + t 8) 6 g

a g

y Wnere: AT as defined in Note 1

=

g AT, as defined in Note 1

=

'{

K 5

1.0875 4

-y K

2 0.02/*F for increasing average temperature and 0 for decreasing average temperature s

8 zs 3

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

=

1+T S avg

.7 3

compensation

'l

.i L

I sc es 0

1 n

a h

t t

9 e

v r

a nm I

r y

o b

f t

r n

e i

l op l

s-o f

i-i p

i l-t i

f l-t u

r n

I o

t u

o P

P-i-

d t-l e

g I

A S-a 1

tu

)

t P-d l

r n

o p

I-e e

f l

m i

R-u i

o t

)

l

(-

n d

i.

0 c

i r

e 1-D s

t

=

u 0-n e

t

[

n 1-o m

l i

i I-C i

A t

K H

d A-(

tn i-e n

d t

o t-l l

e i

t i

n a

C 0

a c

l 1

(

u-d x

l y

e R-I e

v 1

A z

I-i I

1 a

I t

i 2

S-I 0

o l

e e

n l

l i

f i

2 f-t e

t t

i o

t u

l u

l f-f n

l l

L l

f i

r e

t a

t o

n i

n h

l Al l.

n f

i S-e i

a s

I Y

f S

I d

e d

ts e

R e

n t

n

/

n n

P-o 6

i i

f i

o c

5 t

H-t p

e 0

le 1

1 le I-t m

0 i

L e

R~

i s

l s

O s

O~

0 a

L a

p I-C i

A r

E t

H

=

2

=

(-

=

mu mn iaxp aS m

l fega

'sa t

+

K T

I 5

l ten

)

ne d

nc e

ar he u

n cp i

t e4 n

h o

T 2 c

e

(

3 4

E i

E 0

iu 4

1 lf

,,CImf c2w" 4

,rO 33. 3 *r 7

U %6.

.es s

s d

l

!i:

l l

l:I',

TABLE 2.2-1 (Continued) v3 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS j'

NOTATION-(Continued) e

[l NOTE 3: (continued)

Time constant utilized in rate-lag controller for Tavg* 53 2 10 secs.

l 1

=

3 K

2 0.00156/*F for T > T~ and K = 0 for T < T~

g g

T as defined in Note 1

=

T~

587.4*F Reference T at RATED THERMAL POWER avg

'f 5

S-as defined in Note 1

=

i NOTE 4:

The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.4 percent AT Span.

t e

i en 1

1 8-4

. ~..

..o

2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTE?4 INSTRUMENTATION SETPOINTS The Reactor Trio Setooint Limits specified in Table 2.2-1 are the nominal values at wnich the Reactor Trips are set for eacn functional unit.

Setooints have caen selected to ensure that the reactor core and reactorThe Trip coolant system are prevented from exceecing their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety features Actuation System in mitigating the consequences of accidents.

is considered to be adjusted consistent with the nominal val measureo" setpoint is within the bano allowed for calibration accuracy.

.To accommodate the instrument drift assumed to occur between operational tests and the accuracy to wnich setooints can te measured and calibratea Allowaole Values for tna reactor trip setooints have ceen specifiec in Table 2.2-1.

Coeration with setooints less conservative than the Trip Setooint but within the Allowaole Value is acceptable since an allowance nas been made in the safety analysis to accomodate this error, r-:h'

~

c,rning.

G ERA 821- % f : u b,.

"4-i m.

veen

%c m sc6 set nt i cuno t exceea Allowaol

alue, e me odolo f thi o

on u izes "as me reo" devi on fro he s cified libra n

I int f ract d sensor emoonent' n conj tien th a s 1stic comoin tion the o er unce inties of e ins ent on to m sure trie' prece var' ble a the unc ainties i calibr ng t instru::yrntatione'7 nI E ation

.2-1, Z

+ 5 < TA the in,tdacti effect yof the e7rors i he act a the sen r, and th_

f as mea rea" ' lue Z, a peciffe in Table 292-1, in arcerte# span,s of a errors are considered.

i tne statTstical fummation of

'rors as mea in th nalysi excitding thosyassociare'd with racx d ft and th ccura of their measupement.

a sensor

.s tne d* ference, ?ti perce scan /between Jffe trip pe,IA or To Allowa-tooint, d the y a

usea i the analyp4 s for rea,ctorstrio.

R pt Rack Ffror is 'the "as mea read cevi ion, in opecent span, for'the affejrted channel from $he specif d trio oint. S pr Sensar Erretris either he "as frasurea"Aeviation f the se nsor from4ts cal #5rattori point or he val soeciffat in Tabl 2.2-1, n

percent s in, fr the jmilysis a tion Use of,rilouation

.2-1 al ws for a sens drift actonr an increared rackjfrift far,rbr, and p TABigEVENTS.

/

vides threshold valu o r R, l

The methodology to derive the trip setcoints is based upon comoining all of the uncertainties in the cnannels.

Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties.

Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these unce-tainty magnitudes.

Rack drift in excess of the Allowaole Value exhibits the behavior that the racX has not met its allowance.

Seing that there is a small statistical chance that this will happen, an infrecuent excessive drift is expected.

Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious proolems and should warrant further investigation.

SUMMER - UNIT 1 3 2-3 Amenament No. 35

a 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated.

Allowable Values for the reactor trip setpoints have been specified in Table 2.2-1.

Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accomodate this error.

The methodology to derive the trip setpoints is based upon combining all of the uncertainties in the channels.

Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

i

)

i SUMMER - UNIT 1 8 2-3 Amendment No. 36, j

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATf0N SYSTEM INSTRUMENTATION

-i.IMITING CONDITION FOR OPERATION 3.3.2 The Engineereo Safety Feature Actuation System (ESFAS) instrumentation 1

cnannels ano interlocks snown in Table 3.3-3 shall be OPERABLE with their trio setooints set consistent with the values snown in the Trio Setooint column of Table 3.3-4 ana with RESPONSE TIMES as snown in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

i ACTION:

With an ESFAS Instrumentation or Interlock Trip Setooint trip less a.

conservative than the value snown in the Trip Setooint Column but i

more conservative than the value shown in the Allowanle Value Column of Table 3.3-4, adjust the Setooint consistent with the Trip'Setpoint value.

With an ESFAS Instrumentation or Interlock Trip Setpoint. less conse o.

f & w..-T -tive than the value snown in the Allowaole Value Column of Table 3.3-4,

~

qh.

' ust the Setooint consistent with the Trio Setooint value~of I

A I

T

.3-4, ano determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-l ied for the affected channel or, was sa

)

2.

Declare the cn 1 inoperable and apply the appl e ACTION' statement require. s of Table 3.3.3 until t annel is restored l

to OPERABLE status wi

'ts setpoint adjus consistent with the

~

Trip Setpoint value.

EQUATION 2.2-l' Z

+ S 1 TA here:

k f

Z

= the value fra lumn Z of Table 3.3-4 he affected channel,.

R = the "a asureo" value (in cercent span) of ra error for the,

{

l aff ed channel, S

either the "as measured" value (in percent span) of the f

i or.

error, or the value in column 5 of Table 3.3-4 for the affe

{

channel, and TA = the value from column TA of Table 3.3-4 for the affected channel t

With an ESFAS instrumentation channel or interlock inoperaole take the c.

ACTION shown in Table 3.3-3.

N M ER - UNIT ljG 3/4 6 ( Amenament No. 73, 18.

[hchamittinopeyah awl ypL3 Yls.J15opplicahL ACTr Tid L 3 34 adl 4._ cMand6s usbred to h ONME cTatts wr#d

'C SdgohT adjvfed consisTeaf with h%p $sg-trans.

%mmwww~

j 3/4 3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a.

With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint Column but more conservative than the value shown in the Allowable Value Column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.

b.

With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value Column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to its OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

c.

With an ESFAS instrumentation channel or interlock inoperable take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the engineered safety feature actuation system instrumentation surveillance requirements specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

1 SUMMER - UNIT 1 3/4 3-15 Amendment No. 433-93; I

78 -101 3

1 3/4.3 INSTRUMENTATICN I

SURVEILLANCE REOUIREMENTS f

4. 3. 2.1 Eacn ESFAS instrumentation enannel ano interlock ano the automatic actuation logic ana relays snail be demonstrateo OPERABLE by performance of tne engineereo safety feature actuation system instrumentation surveillance recuirements spec 1fied in Table 4.3-2.

a.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of eacn ESFAS function shall be cemonstrated to be within the limit at least once Der 18 months.

l Eacn test snall include at least ene train such that botn trains are tested at least once per 36 months anc one enannel per function such that all enannels are tested at least once per N times 18 months wnere N is the total nuncer of reaunaant enannels in a specific E5FAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

1 gd 4 ele YbiS f"f SU MER - UNIT 1 3/4 3-15a Amenoment No.13. 101

~,

w

'l

.b I

THIS PAGE LEFT BLANK INTENTIONALLY.

4 SUMMER - UNIT 1 3/4 3-15a Amendment No. 44r-101,

TABLE 3.3-4 m

7 EllGINEERED SAFETY FEATURE ACTUATI0ll SYSTEH INSTRUNEllTATI0ll TRIP'SETPolitTS E

4 Functional Unit

~

t Al owance (TA)

Z

~ !P lS Trip Setpoint Allowable Value i

l.

SAFETY INJECTION, REACTOR TRIP, FEEDWATER ISOLATION, CONTROL I

R00H ISOLATION, START DIESEL r

GENERATORS, CONTAINHENT C00LIllG i

FANS AND ESSENTIAL SERVICE WATER.

a.

Manual Initiation ilA ilA flA ilA

-. lA s.

I b.

Automatic Actuation Logic

) llA

\\

NA flA flA ilA

(

\\

w c.

Reactor Building Pressure-3.0 i

0.71 1.5 13.6 psig 13.06 psig a

1 High 1

\\

\\,

10.71 1.5

>1850 psig

>1839 psig d.

Pressurizer Pressure--Low 13.1 e.

Di f ferential ' Pressure 3.0

\\

0.07 1.5 197 p,sIg kl06 psi' Between Steamlines--Il10h 1.5 l

gil' k

>675 psig

>635 ps l (1') '

^

\\

f.

Steamline Pressure--Low 20.0.

10.71 1.5 0

2.

REACTOR DUILDIllG SPRAY a.

Manual Initiation IIA Nk tiA'

' llA A

flA\\

b.

Automatic Actuation Logic flA ilA NA A

and Act'uation Relays' f

i l

c.

Reactor Dullding Pressure-IO 0.71 1.5

<12.05 psig -

12.31 psig H10h'3.(Phise*'A isolation

~

aligns spray system dis-charge valves and NaOH tank-suction' valves)

(1) Time > cnact:= ate

..t i l i-i' ta '--" '--

^

~

~

~

~ -.....

. ~ -.

TABLE 3.3-4 y

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

$x Functional Unit Trip Setpoint Allowable Value E

Q 1.

SAFETY INJECTION, REACTOR TRIP, FEEDWATER ISOLATION, CONTROL ROOM ISOLATION, START DIESEL GENERATORS, CONTAINMINT COOLING FANS AND ESSENTIAL SERVICE WATER.

a. Manual Initiation NA NA

b. Automatic Actuation Logic NA NA

c. Reactor Building Pressure-53.6 psig 53.86 psig gg High 1 y

d. Pressurizer Pressure--Low 21850 psig 31839 psig

e. Differential Pressure 597 psig 5106 psi Between Steamlines--High

f. Steamline Pressure--Low 1675 psig 1635 psig(l) 2.

REACTOR BUILDING SPRAY

a. Manual Initiation NA NA

b. Automatic Actuation Logic NA NA and Actuation Relays R.

c. Reactor Building Pressure-

$12.05 psig

$12.31 psig 3

High 3 (Phase 'A' isolation s

aligns spray system discharge E

valves and Na0H tank suction valves)

(1) Time constants utilized in. lead lag controller for steamline pressure-low are as follows:

Ti 2 50 secs.

12s 5 secs.

~

mc5 TABLE 3.3-4 (Continuedl m:o EliGIllEERED SAFETY FEATURE AClllAT!0ft SYSTEH IliSTRuttEllTATI0il TRIP SETPOIliTS i

g

_Iotal

~

/

q Functional unit A M ance (TA)

• Z 5

Trip Setpoint Allowable Value

\\

e 3.

CONTAllitiEliT ISOLATI0!i

)

s A

a.

, Phase "A" Isolation

's l

x, 1.

Manual flA 'N ilA liA flA 2.

Safety Injection See 1 above for pll safety injection setpoints and allowable values 3.

Automatic Actuation Logic [i flA fjA'T flA liA and Actuation Relays

.b.

Phase "B" Isolation

[

/

1.

Automatic Actuation flA flA flA NA liA

}

Logic and Actuation

}

N l

Relays f

N h

\\ q'.5 2.

Reactor Building L. 3).

0.71 112.05 psig

$12.31 psig Pressure-liigh 3

/

Purge and Ekhaust Isolation c.

1.

Safety Injection See 1 above for all safety injection setpoints and allowable values 2.

Containment Radioactivity NA fiA liigh

. N 3.

Automatic Actuation

. IIA NA NA flA lik

~F Logic and Actuation Relays g

re

.O

• Trip setpoints shall be set to ensure that the limits of Specification 3.11.2.1 are not exceeded.

~

I B r'.i, afa<. /../. 2. I i-4

---

a

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Functional Unit Trip Setpoint Allowable Value E

f 3.

CONTAINMENT ISOLATION

a. Phase "A" Isolation

1. Manual NA NA

2. Safety Injection See 1 above for all safety See 1 above for all safety injection setpoints and allowable injection setpoints and allowable values values

3. Automatic Actuation Logic NA NA

{

and Actuation Relays

{

b. Phase "B" Isolation

1. Automatic Actuation Logic NA NA and Actuation Relays

2. Reactor Building 512.05 psig

$12.31 psig Pressure-High 3

c. Purge and Exhaust Isolation

1. Safety Injection See 1 above for all safety See 1 above for all safety y

injection setpoints and allowable injection setpoints and allowable g

values values

2. Containment Radioactivity High z

3. Automatic Actuation Logic NA NA 7

and Actuation Relays

• Trip setpoints shall be set to ensure that the limits of ODCM Specification 1.2.2.1 are not exceeded.

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TABLE 3.3-4 (Continued)

~

E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

Functional Unit Trip Setpoint Allowable Value E

4 4.

STEAM LINE ISOLATION

a. Manual NA NA

b. Automatic Actuation Logic NA NA and Actuation Relays

c. Reactor Building Pressure-56.35 56.61 High 2

d. Steam Flow in Two Steamlines-5 a function defined as follows:

5 a function defined as follows:

t>,

High, Concident with A apcorresponding to 40% of full A ap corresponding to 44% of full steam flow between 0% and 20%

steam flow between 0% and 20%

4 load and then a ap increasing load and then a ap increasing linearly to a ap corresponding to linearly to a Ap corresponding to 100% of full steam flow at full 114.0% of full steam flow at full load load Tavg - Low-Low 2552.0'F 2548.4*F

e. Steamline Pressure-Low 2675 psig 1635 psig(1)

E e

S s

E

-i (1) Time constants utilized in lead lag controller for steamline pressure low are as follows:

~

11 2 50 secs.

12 s 5' secs.

n.

E 9

1ABLE 3.3-4 9:

Functional Unit

{s _fptal Allowarice (IA) 2 5,

Trip Setpoint.

Allowable Value c:*i S.

TURBINE TRIP AND FEEDWATER

('

~i ISOLATION w

j Steam Generator Water a.

5. 0 2.18 1.5 sB2.4% of (34.2% of nariow Level - High-liigh jg yarrowrange ranDe instrument

/

q instrument span span 6.

EHERGENCY FEEDWATER

\\s

\\

a.

Manual flA NA fiA NA NA

'\\

b.

Automatic Actuation Logic HA

\\

tiA fiA HA tiA

/

/

//, 3,/

• t c.

Steam Generator Water 12.0

/

9.18 1.5

>l2% of span

> TD>2% of spari 1 on Level - Low-Low Y

['

from 0% to 30%

0% to 305 RTP in-as

/

RIP increasing creasing linearly I linearly to

> Q. of span from

>30.0% of span 30% Ao 2001 RTP -

from 30% to 100%

RIP 2

d. & f.

Undervoltage-ESF Bus

>S760 volts with

>$652 volts with a a <0.25 second 20.?75 second time time delay delav y

>6576 volts with

>6511 valts with a g

a <3.0 second 33.3 second time I

E time delay delay 3e j

TABLE 3.3-4 (Continued)

E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

E Functional Unit Trio Setpoint Allowable Value E

4 5.

TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water

$82.4% of narrow range instrument

$84.2% of narrow ra.~.ge instrument Level - High-High span span 6.

EMERGENCY FEEDWATER a.-Manual NA lNA lNA

b. Automatic Actuation Logic NA y

'c.-Steam Generator Water 112% of span from 0% to 30%

211.2% of span from 0% to 30T%

level - Low-Low RTP increasing linearly to 230.0%

RTP increasing linearly to 229.2%

a

/>

of span from 30% to 100% RTP of span from 30% to 100% RTP I

m

'd.

& f.

Undervoltage-ESF Bus l25760Voltswitha50.25second 25652 Volts with a 50.275 second time delay time delay 26576 Volts with a 53.0 secona 25511 Volts with a 53.3 second l time delay time delay N

2it an

.m m

v

.;_ 2.

2 :y:n My ~- c; ( g-g.

.c.

j~ ;..

i f 3.,

v-l<

E g

TABLE 3.3-4 (Continued}

ENGINEERED SAFEIY FEATURE ACTUATION SYSTEM INSTRilMENTAT

_E

-4 functional Unit m -- -

m Q

Allowance (TA)

Z S

Trip 5etpoint Allowable Value e

e.

Safety Injection

/

g Se 1 above (all 51 Setpoints >

1 g.

Trips of Hain feedwater U

HA NA NA NA NA Pumps

\\

f h.

Suction transfer on Low I

NA

\\

NA NA 3442 ft. 4in.

3441 ft. 3 in.

Pressure 7.

LOSS OF POWER k

\\

I i

s y

a.

7.2 kv Emergency Bus NA I

Undervoltage (Loss of.

HA NA 35760 volts with 35652 volts with a y

Voltage) g a $0.25 second 50.275 second time

\\

H t

time delay delay

\\

=

b.

7.2 kv Emergency Bus

'NA

\\ NA NA 16576 volts 16511 volts with a-Undervoltage i

\\

with a <3.0

<3.3 second time j

\\

second time' delay I

\\

{ delay 8.

AUTOMATIC SWITCHOVER TO s

//

\\'

CONTAINMENT SUMP q

e

-\\

RWST Level Low-Low f NA NA k a.

NA 318%

115%-

b.

Automatic Actuation Logic HA NA NA NA NA f

and Actuation Relays 4

(2)

Pump suction head at which transfer is: initiated is stated in ef fective water elevati condensate storage tank.

on in the-am ---

.m u.

m 2-

. +,,,,

,v..,

y 4

TABLE 3.3-4 (Continued) g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

5 Functional Unit Trio Setpoint Allowable Value

e. Safety Injection See 1 above (all SI Setpoints)

See 1 above (all SI Setpoints)

~

g. Trips of Main Feedwater Pumps NA NA

h. Suction transfer on Low 2442 ft. 41n. (2) 2441 ft. 3 in.

Pressure 7.

LOSS OF POWER

a. 7.2 kv Emergency Bus 15760 volts with a 10.25 second 25652 volts with a 20.275 second w2 Undervoltage (Loss of Voltage) time delay time delay

[

b. 7.2 kv Emergency Bus 36576 volts with a 53.0 second 16511 volts with a 53.3 second F

Undervoltage time delay time delay 8.

AUTOMATIC SWITCHOVER TO CONTAINMENT SUMP

a. RWST Level Low-Low 218%

315%

b. Automatic Actuation Logic NA NA and Actuation Relays i

S.

a r-E.

(2) Pump suction head at which transfer is initiated is stated in effective water elevation in the condensate storage tank.

l

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1

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

$x Functional Unit Trip Setpoint Allowable Value EE 4

9.

ENGINEERED SAFETY FEATURE-ACTUATION SYSTEM INTERLOCKS INTERLOCKS

a. Pressurizer Pressure, P-11 1985 psig 21974 psig &

$1996 ps'g

b. Tavg LOW-Low, P-12 552*F 2548.4*F &

$555.6*F

c. Reactor Trip, P-4 NA NA Y'

Me a

c E

I l

3/4.3 INSTRLIMENTATf0N BASES 3/4.3.1 ano 3/4.3.2 5Y512M INSTRUMENTATIONREACTOR TRIP AND ENGINEERED S Feature Actuation System Instrumentation ano inte y

associateo action ana/or reactor trip will be initiated when the parameter e

monitored by eacn cnannel or comnination thereof reaches its setootnt a channel to be out of service for testing or mainten

, 2) the maintaining an accrocriate level of reiliability of the Reactor Protection Engineerea Safety Features instrumentation and, 3) sufficient syste capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall

-i reliability, reounaancy, and diversity assumeo available in the facility design for the protection and mitigation of accident and transient conditi The integrated operation of each of these s ons.

assumotions usea in the accident analyses. ystems is consistent with the cacao 111ty is maintainea comoaraole to the origi periccic surveillance tests performed at the minimum frequencies are The sufficient to demonstrate this cananility.

Specified surveillance and surveillance and maintenance outage times have been determined in accorca with WCAP-10271, " Evaluation of Surveillance Freauencies and Out of S Times for Reactor Protection Instrumentation System," and supplements ervice Surveillance intervals and out of service times were dete report.

{

on maintaining an appropriate level of reliability of the Reactor Protection j

System ana Engineerea Safety Features instrumentation.

i Setpoints specified in Table 3.3-4 are the nominal value bistables are set for each functional unit.

adjusted consistent with the nominal value wnen thA setcoint is considered to be within the band allowea for calibration accuracy. e "as measured" setpoint is To accomoaate the instrument drift assumed to occur between operational tests ano the accuracy to which setpoints can De measured and calibrated Allowaole Values for the setpoints have been specified in Table 3.3-4 Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accomodate this error.

" :stier-' ;rni.

ine) afor/etermini the OPE ILITY a chan when it rip se nt is ound t yexceed Allowa Value he me dology of is opt lizes 1e "as suread iation om the - ecified ibratic oint to rack a sensor omeonent in can' ction w' a stat ical ce nation o the.

er un. taintie f the trumen ion to mpesure t a ' the un taintie in cal-ating instrumentation.

Equatio

.3 1 recess v able

+R+c TA, th intera

,ve ef s of thapd'rrors i he rack the sensor and the s mea ed" va s of rors a considere.

Z, as s

SUMMER - UNIT 1 B 3/4 3-1 Amendment No. 101

<, ' o 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Protection System and Engineered Safety Feature Actuation System Instrumentation and interlocks ensure that 1) the associated action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoints, 2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to ba out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation and, 3) sufficient system functions capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Specified surveillance and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out nf Service Times for Reactor Protection Instrumentation System," and supplements to that report.

Surveillance intervals and out of service times were determined based on maintaining an aporopriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

The Engineered Safety Feature Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A setpoint is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the setpoints have been specified in Table 3.3-4.

Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since all allowance has been made in the safety analysis to accommodate this error.

l The methodology to derive the trip setpoints is based upon combining all of the uncertainties in the channels.

Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected.to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this SUMMER - UNIT 1 B 3/4 3-1 Amendment No. 101

V' 1

NSTRUMENTATf0N BASES REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (continueo) specified in T e 3.3-4 incercentIan,is ~

statis al summa n of errors assue in the alysis excd ding tho associa with th ensor a racx drif and the uracy of p(betwee eir measu ent.

T or Total 3 owance the dif# rence, i ercent sc e trip tooint anp'the valu used in t analysis or the ac tion.

R Rack Epror is thef as measur i

de' ' tion, in# percent so, for the fected J:tiannel freerthe specjffed trip coint. for Sensor tror is e' e

er the/as measureo# deviatipri of the sensor f ation po' or th percen,tr' pdm its cal

'alue spectfTed in Ta We 3.3-4, span, fr the analy s assu ons.

Useso? Eaue.tiosf 3.3-1 al s for as or drift ctor, an ' crease ack drif v

e for RE TABLE EV S

ctor,anf$rovides[ threshold a

/

The dhedology to derive the trip setpoints is based upon c'ombining all of the uncertainties in the channels.

Inherent to the determination of the trio setooints are the magnituces of these channel uncertainties.

Sensor and racx instrumentation utilized in these enannels are expected to be cacaole of coerating r. thin the allowances of these uncertainty magnitudes.

Rack rfrift in excess of the Allowable Value exhibits the behavior that the rack has not met its al' mance.

Being that there is a small statistical chance that this will happen, an infrecuent excessive orift is expected.

Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

The measurement of response time at the specified freousncies provides assurance that the reactor trip and the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demon-strated by any series of secuential, overlapping or total channel test measurements provided that sucn tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite, or offsite test measurements or 2) utilizing replacement sensors with certified response times.

The Engineered Safety Features Actuation System senses selected plant parameters and deterwines whether or not predetermined limits are being If they are, the signals are combined into logic matrices sensitive exceeded.

to combinations indicative of various accidents, events, and transients. Once the reouired logic combination is completed, the system sends actuation signals to those engineered safety features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consecuences of a steam line Dreak or loss of coolant accident 1) safety injection pumps start and automatic valves position, 2) reactor trip, 3) feed-i water isolation, 4) startup of the emergency diesel generators, 5) containment spray pumps start and automatic valves position, 6) containment isolation,

7) steam line isolation, 8) turoine trip, 9) auxiliary feedwater pumps start and automatic valves position, 10) containment cooling fans start and auto-matic valves position,11) essential service water pumps start and automatic valves position, ano 12) control room isolation and ventilation systems start.

SU)HER - UNIT 1 B 3/4 3-la Amenament No. 35. 101 l

j

r e s INSTRUMENTATION BASES REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (continued) l-will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

The measurement of response time at the specified frequencies provides assurance that the reactor trip an<i the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demon-strated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite, or offsite test measurements or 2) utilizing replacement sensors with certified response times.

The Engineered Safety Features response times specified in Table 3.3-5 which include sequential operation of the RWST and VCT valves (Notes 2 and 3) are based on values assumed in the non-LOCA safety analyses. These analyses are for injection of borated water from the RWST.

Injection of borated water is assumed not to occur until the VCT charging pump suction isolation valves are closed following opening of the RWST charging pumps suction valves. When the sequential operation of the RWST and VCT valves is not included in the response times (Note 1) the values specified are based on the LOCA analyses.

The LOCA analyses take credit for injection flow regardless of the source.

Verification of the response times specified in Table 3.3-5 will assure that the assumptions used for the LOCA and non-Loca analyses with respect to the operation of the VCT and RWST valves are valid.

The Engineered Safety Features Actuation System senses selected plant l

parameters and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those engineered safety features components whose aggregate function best serves the requirements of the condition. As an example, the following actions 1

may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident 1) safety injection pumps start and automatic valves position, 2) reactor trip, 3) feed-water isolation 4) startup of the emergency diesel generators, 5) containment spray pumps start and automatic valves position, 6) containment isolation,

7) steam line isolation, 8) turbine trip, 9) auxiliary feedwater pumps start and automatic valves position, 10) containment cooling fans start and auto-matic valves position, 11) essential service water pumps start and automatic valves position, and 12) control room isolation and ventilation systems start.

SUMMER - UNIT 1 B 3/4 3-la Amendment No. -3h--101-i