ML20064G554

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Proposed Tech Specs Re Limiting Safety Sys Settings for Reactor Trip Sys Instrumentation Trip Setpoints ESF Actuation Sys Instrumentation Setpoint
ML20064G554
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 03/11/1994
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SOUTH CAROLINA ELECTRIC & GAS CO.
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ML20064G548 List:
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NUDOCS 9403160230
Download: ML20064G554 (42)
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS r

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 chnwn i, the A11awahle1 Values column of Table 2.2-1.r7p a~ce the cnannel in t1e trippeo condition witn;n i hour,]

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

1. t Equation 2.2-1 was satisfied for the affect annel and adjust th point consistent with the Trip Setpo alue of I Table 2.2-1, or
2. Declare the channel ino able and apply pplicable ACTION statement requirement of ification .1 until the channel is j

restored to OPERABLE status h it tpoint adjusted consistent with the Trip Setpoint value, i \ i

! EQUATION 2.2-1 + R + S < TA i l

l where:

I Z= the value for col Z of Table 2.2-1 for he affected channel, I R= the "as meas d" value (in percent span) of ck error for the affected nel, S= eith he "as measured" value (in percent span) o 'he sensor er , or the value is column S of Table 2.2-1 for ti affected i I

nnel, and ,

T the value from column TA of Table 2.2-1 for the affected ch el. g Y _- _. - -

clecluc %c%ul . Leo ushle a$d Qlu -t( w &w

        1. Y"*NkO ts natora k Waus srahr wduts egpt 4 k3,I uxfGL-fhl$g[4- ,, g wA & 9 sqa m.

SlM4ER - UNIT 1 2-4 .

9403160230 940311 PDR ADOCK 05000395 P PDR

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.

SUMMER - UNIT 1 2-4 Amendment No.

1AM [ 2.2-1 m REALIOR IHIP $YSitH INSTRi!MtNIAll0N IHIP St IP0lHIS C

Y functional Unit Z S Irlp Setpoint Allowable Value A losait.e (TA)

'l . Manual Reactor Trip ( Not\pplicable NA N[

l NA NA

- 2. Power Range, Neutron flux liigh Sepoint 7.5 4.56 . 0 $109% of RTP $111.2% of RTP Low Setpoint .3 4.56j 0 g $25% of RTP $21.2% of RTP

3. Power Range, Heutron flux 1. O. O l 15% of RTP with a time $6.3% of RTP with a time liigh Positive Rate ly { constant 22 seconds constant 22 seconds
4. Power Range, Neutron flux / 1.6 \ j6.5 0 ) 55% of RIP with a time $6.3% of RTP with a time liigh Negative Rate 8;a m ( g ,/ ( constant 2 2 seconds constant 22 seconds

.S. Intermediate Range, j 17.0 8.4 0

$25% of RTP $31% of RTP Neutron flux y )

6.

7.

Source Range, Neutron flux (

Overtemperature AT /

17.0[' \l0.0 0

[ $105 cps $1.4 x 105 cps 10.3 / .8 1.6 See note 1 See note 2 l

/ & l.2*

g 8. Overpower AT I 1.6 See note 3 See note 4 5/2 1.}6 g 9. Pressurizer Pressure-Low (g 1.5 21810 psig 21859 psig

)I.1 0.7k g 10. Pressurizer Pressure-High / / 3.1 0.71 ' q 1.5 j 52380 psig $2391 psig '

11. Pressurizer Water Level-flighi / 5.0 2.18 1.5 $92% of instrument $93.8% of instrument 7/ span span h 12. Loss of flow- 'I\ [ 2.5 1.48 190% of loop design- 288.9% of loop design flow
  • flow * -
  • Loop design flow = 94,870 gpm l'

?

O RTP - RATED THEllMAL POWER

- - :." '" :;;; in Am (=::4 ..a :.a s. 7.m...;m, r.u x.;. - - --

_________________1________._ ____-_. _ ________ __

TABLE 2.2 1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Functional Unit Trip Setpoint Allowable Value

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 12 seconds constant 32 seconds

4. Power Range, Neutron Flux 55% of RTP with a time $6.3% of RTP with a time High Negative Rate constant 22 seconds constant 12 seconds
5. Intermediate Range, Neutron Flux 525% of RTP 531% of RTP
6. Source Range, Neutron Flux 5105 cps 51.4 x 105cps
7. Overtemperature AT See note 1 See note 2
8. Overpower AT See note 3 See note 4
9. Pressurizer Pressure-Low 21870 psig 11859 psig
10. Pressurizer Pressure-High 52380 psig 52391 psig
11. Pressurizer Water Level-High $92% of instrument span $93.8% of instrument span
12. Loss of Flow 190% of loop design flow
  • 188.9% of loop design flow *
  • Loop design flow = 94,870 gpm RTP - RATED THERMAL POWER I

n 9'

TABLE 2.2-1 (continued) ,

c REACTOR 1 RIP SYSTEM INSTRUMENTATION TRIP SETPOINIS y functional Unit ota Allowance (TA) Z_ , Trip 5etpoint

13. Allowable Value Steam Generator Water level low-Low 1.0 M l l >12% of span from
11. 2 .

>19<2% of span from y 7'7 / D to 30% RTP 5 to 3a% RIP 1 increasing lin- Increasing linearly early to >30.0% of to >gSee: of span span fros 30% to 100% RTP froin 30%"1e 100%

14.

Steam /feedwater flow Mis- 16.0 RTP ggg,'g, 13 24 1.5/ <40% of full Match Colacident With <42.5% of full

1. 5 iteam flow at RIP iteam flow at RIP Steam Generator Water Level 12.0 11.2.

4" tow-to" >12% of span from >18d% of span from M5 /7 3 ** 385 "T" increasing lin- 3 ** 38% "3P early to 130.GE of facreasing ifnearly span from 30% to to >JSeft of span from 3 o 100%

1005 ATP RTP

15. Undervoltage - Reactor 2.1 g,g Coolant Pump .28 0.23 14830 volts 14760
16. Underfrequency - Reactor 7.5 Coolant Pumps 0 0.1 >57.5 Hz >57.1 Hz

{a 17. Turbine Trip A. Low irlp System Pressure MA g B. Turbine Stop Valve 4A HA NA 1800 psig 1750 psig

, a Closure NA NA 11% open 11% open ee sanoa s.ga m euwes

l.

1

TABLE 2.2-1 (continued) f REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l

Functional Unit Trip Setpoint Allowable Value

13. Steam Generator Water 312% 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 $42.5% of full steam flow at RIP 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 14830 volts 24760 Coolant Pump
16. Underfrequency - Reactor 257.5 Hz 157.1 Hz Coolant Pumps
17. Turbine Trip A. Low Trip System Pressure 2800 psig 1750 psig B. Turbine Stop Valve Closure 21% open 21% open RTP - RATED THERMAL POWER

I j'

TADtE 2.2-1 (continued]

REACTOR 1 RIP SYSTEH INSTRllHElliAT1011 TRIP SEIPOINTS

' Total Functional unit ilowance (IA) I Trip Setpoint Allowable Value h 18. Safety Injection Input . li- NA - (A NA flA

- f ror. ESF i

19. Reactor Trip System Interlocks A. Intermediate Range NA NA .. HA >7.5 x 10 6% >4.5 x 10 6%

Neutron Flux, P-6

(

Indication Tndication B. Iow Power l'eactor Trips Block, P-7

a. P-10 input 7. 5 0 $10% of RTP f.56 $12.2% of RTP m b. P-13 input 7.5 4.56 0 4 <10% turbine <12.2% of turbine impulse pressure Impulse pressure j

i equivalent equivalent C. Power Range Neutron 7. 5 4.56 0 <38% of RTP Flux P-8 't

<40.2% of RIP l

D. Low Setpoint Power 7.5 4.56 0 >10% of RTP Range Neutron Flux, P-10 >7.8% of RIP i

E. Turbine Im 7.5 4 56 0 Pressure pulse P-13 Chamber <10% turbine

/<12.2% turbine Impulse pressure jiressure equivalent equivalent a

s F. Power Range Neutron -

7.5 4.56 0 Flux, P-9 [ ) 150% of RTP $52.2% of RIP H

20. Reactor Trip Breakers HA NA NA NA NA y 21. Automatic Actuation Logic NA NA NA flA flA RIP = RATED 111ERMAL POWER g; <
t. ,

.?

TABLE 2.2-1 (continuad)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Functional Unit Trip Setpoint Allowable Value

18. Safety Injection Input from ESF NA NA
19. Reactor Trip System Interlocks A. Intermediate Range Neutron Flux, P-6 27.5 X 10 6% indication 24.5 X 10 6% indication B. Low Power Reactor Trips Block, P-7
a. P-10 input 510%ofRTP $12.2% of RTP
b. P-13 input $10% turbine 512.2% of turbine impulse pressure equivalent impulse pressure equivalent C. Power Range Neutron Flux P-8 $38% 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 F. Power Range Neutron Flux, P-9 550% of RTP 552.2% of RTP
20. Reactor Trip. Breakers NA NA
21. Automatic Actuation Logic NA NA RTP - RATED THEPJiAL POWER

iABlL 2.2-6 ( Loi.' i nued )

REACIOR 1 RIP SYSTEM [HU RIMENTATION THIP SilPOINIS

{:o Nola!10!i (;!nt inued) h NOTE 1: (Continued)

.5 a

and f, (al) is a function of the indicated dit f eren:e between top and bot tom detecto. s of the power-range nuclear ton chambers) with gains to be selected bas 2d on measured instrument response during plant startup tests such that:

(1) for qt - qb between - 21 percent and i 4 per unt t, (al) 0 where qt d"d 4h d'" P ' ""I H A II II IIII NII^I POWER in the top and bottom halves of the coro respettively, and qt *4h is t tal lilEHilAt. POWER in perte.. of RATED TilERMAl POWER.

(11) for each percent that the magnitude of yt - 4 3 sceeds -24 percent, the al trip setpoint shall be m, automatically reduced by 2.27 percent of lis talue at RATED TilERMAt POWER.

b (lit) for each percent that the magnitude of qt - 4) exceeds +4 percent, the al trip setpoint shall be automatically reduced by 2.34 percent of its salue at HAILD lilERMAl. POWER. I NOTE 2:

Ihe channel's maximum trip setpoint shall not exceei its computed trip point hy more than 2.2 percent al Span.

HOIE 3: OVERPOWER AI (T,5)

- ! :, S) -

{

AT SAT " K-K d 6 (1 + ta SI T-K' T-T k

3 Where: ai =

as defined in Note 1 o f, =

as defined in Note 1

~

m K, 5 1.0875 K 1 D

5 0.02/*F for increasing ave age temperature and 0 for decreasing average temperature s, S

, g

=

The function generated bv :he rate-lag controller for I avg dynamic g a compensatlon so

, .._m__._ _ _ _ ___..___________ ____. _ . _ _ . _ _ , _ _ _ . . . _

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

NOTE 1: (Continued) and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range-i nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) for qt - 9b between -24 percent and +4 percent f (AI) = 0 where qt and qb are percent RATED THERMAL 3

POWER in the top and bottom halves of the core respectively, and gt + 4 b is total. THERMAL POWER in percent of RATED THERMAL POWER.

(ii) for each percent that the magnitude of qt - 4b exceeds -24 percent, the AT trip setpoint shall be automatically reduced by 2.27 percent of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of qt - 9b 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 (ta b -1 ATsAT K-K 4

T-K T-T o

5(1 + t3S) 6 Where: AT = as defined in Note 1 AT, = as defined in Note 1 K4 5' 1.0875 K3 > 0.02/*F for increasing average temperature and 0 for decreasing average temperature ,

5 S 3

1+5 S

=

The_ function generated by the rate-lag controller.for T avg dynamic 3 compensation

_ 1 1ABLE 2.2-1 (Continued) 3 E REACTOR 1 RIP SYSTEM INSTRuffEtiTATI0ff IRIP SETPOINIS e

NOTAT10N (Continued)

E

" NOTE 3 (continued)

Y g

, =

Time constant utilized in the rate-iag contrailer for T,yg, 13 > 10 secs. i Ks >

0.00156/ F for I > 1" and Ks = 0 for i < 1" T =

as defined in Notc I i

\

I" <

587.4 F Reference I, g at RATED TilERMAL POWER S =

as defined in Note 1 NOTE 4:

Thepercent 2.4 channel's maximum trip setpoint shall not exceed its computed trip point by more tnan AT Span.

I 2

2 r*

'S . .

. e

- = e- e - w T-=a m 's M-a e+rv+- - e

-l?

TABLE 2.?-1 (Continued)

, REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

NOTE 3: (continued) *

=

t 3 Time constant utilized in rate-lag controller for Tavg* Ia 3 10 secs. I K 2 s 0.00156/*F for T > T" and Kg = 0 for T 5 T~

r T = as defined in Note 1 T~ 5 587.4*F Reference Tavg at RATED THERHAL POWER S =

as defined in Note 1 NOTE 4: The channel's maximum trip setpoint shall not exceed its computed trip point by more than a 2.4 percent AT Span.

i i

7

.-.4 - <- = _. r -v- , - - - w __ _ _- . ..v w w .-_ _ _ _ ____ m _ __ .-_.-.___ __ _ __________.-__m. m_.._ mm_.___

2. 2 LIMITING SAFETY SYSTEM SETTINGS BA1Q 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.

, Setpoints have been selected to ensure that the reactor core and reactorThe Trip coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist theaccidents.

of Engineered Safety Features Actuation System in mitigating the consequences  !

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 Allowable Table 2.2-1.

Values for the reactor trip setpoints have been specifie

Operation with setpoints less conservative than the Trip Setpoint but within the Allowaole Value is acceptable since an allowance has Deen made in the safety analysis to accommodate this error, i

{'j uccu g= mm m m my . vec,v e t, . . q,f.pf r.

set ,

nt i ound t exceed Allowabl alue, ,

o on u izes "as me e me odolo f thi red" devi on fro he s cified libra int f rack ~ d sensor omponent n conj tion n tion the o er uncer inties o th a s 1stic combin e ins ont on to sure proce var ble a 'the unc ainties i alibt instr Fjtiation .2-1, I

+S 1 TA ng t tatiorgV In

/Fack a the sen r, and t hein)siacti effect of thejftrors i he ac mea Z, a red" lue e errors are c idered.

peciffe in Table 1, i erce/ span,sof of trors as ed in th. nalysi exc1 / i he sta4Tstical ation associ "d with e sensor rack d ft and th accura of tM'9ir ing thosment.

measu or To

,sthed'ference,japerce e tri Allowa d the y used i the analypfs for reactorArip. span [between)

R pt- Rack j/p petpoint, d

a devi ion, in ppecent span, for# the aff ed cha ror is 'tht! as mea red" se oint. S,pe Sens . Erro i fromffie specif d trip nsor fremd ts ca rati s either point or he vaihe "as easured"/tfeviation f the percent in, fr the ~alysis a-specif ' in Tabl 2.2-1, n tion Use o a sens drift actor an incre ed rack rift fa quation .2-1 al ws for valu or R, TAB EVENTS.

r, and p vides threshold l

of the uncertainties in the channels.The methodology to derive the trip se trip setpoints are the magnitudes of these channel uncertainties. Sensors and Inherent other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift i in excess met its allowance. of the Allowable Value exhibits the behavior that the raj will happen, an infrequent excessive drift is expected.Being that there (

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

SUMER - UNIT 1 B 2-3 Amenoment No. 35

1 2.2 LIMITING SAFETY SYSTEM SETTINGS -

BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS  !

l The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal i 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 - i but within the Allowable Value is acceptt!,le since an allowance has been made in the safety enalysis 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.

SUMMER - UNIT 1 8 2-3 Amendment No. 35

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

i I

t

~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 j 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.

3 ACTION:

-l

a. With an ESFAS Instrumentation or Interlock Trip Setpoint trip less I conservative than the value shown in the Trip Setpoint Coluun but l more conservative than the value shown in the Allowable Value Column '!

of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint 1 value. .

b. '

With an ESFAS tive than the value Instrumentation or Interlock shown in the Allowable Trip Value ColumnSetpoint less conserva of Table 3.3-4, b y u , __7 _ ___

~

l

. ust the Setpoint consistent with the Trip Setpoint value~of T ,

.3-4, and 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- '

was sa ied for the affected channel or, .

2. Declare the ch 1 inoperable and apply the appl le ACTION j statement' require s of Table 3.3.3 until t annel'is restored a'  !

to OPERABLE status wi its setpoint adjus consistent with the Trip Setpoint value. .

=J EQUATION 2.2-1 Z + S $ TA here:

k 'o Z = the value fro lumn Z of Table 3.3-4 he affected channel, .j R = the "a asured" value (in percent span) of ra error for the , I ed channel, aff ,

)  !

S either the "as measured" value (in percent span) of the or .l i

, error, or the value in column S of Table 3.3-4 for the affe l d channel, and  !

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

__ j, j c.' With an ESFAS instrumentation channel or interlock inoperable take the ACTION shown in Table 3.3-3.

l e

3/4 ( P ( Amendment No. 73, 78 ,

ma,%

UM4ER - UNIT 1

%y-,- - 5 q

LL 3 3-4 Kd &_ chat g nestosd to in ONAe(s stab & cts ScrgenTadjcted consis%(we% -h%cp fsg- p&e, W .

3/4.3 INSTRUMENTATION SURVEILLANCE REOUIREMENTS-f-

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 Tabla 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. -l ;

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 i

~

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.

i e

and A cfe Ybh PW 1

I 1

)

I l

SUMER - UNIT 1 3/4 3-15a Amendment No.13. 101 l

3/4 3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION i

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 ENGIllEERED 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.

4 SUMMER - UNIT 1 3/4 3-15 Amendment No. 13, 73, 78, 101

i TABLE 3.3-4 -

m 7 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP'SETPOINTS ...

E '

t

Q Functional Unit Al owance (TA) Z S Trip Setpoint Allowable Value

!H \ / .

FEEDWATER ISOLATION, CONTROL \

i ROOM ISOLATION, START OIESEL GENERATORS, CONTAINMENT COOLING l FANS AND ESSENTIAL SERVICE WATER.

a. Manual Initiation NA iA NA NA -

(A

b. Automatic Actuation Logic flA \ NA NA NA NA
c. Reactor Building Pressure-

\ .

w 3.0 \ 0.71 1. - ~<3.6 psig <3.86 psig

) High 1

~

w $

$ d. Pressurizer Pressure--Low .

13.1 10.71 1.5 11850 psig 11839 psig

e. Differential Pressure 3.0 \ 0.87 1.5 197 psig $1,06 psi Between Steamlines--liigh \ 1.5 l
j
f. Steamline' Pressure--Low L 20.0 10 1. 5 1675 psig 635 psig(1)

\ .71

2. REACTOR BUILDING SPRAY \
a. Manual Initiation 1 NA N NA 'NA 'A Automatic Actuation Logic
b. NA / NA NA NA A ,

l and Act'ation u Relays' / k l

c. Reactor Building Pressure - 0 0.71 1.5 $12.05 psig - 2.31 psig High'3-(PhaseA" isolation ' <

aligns spray system dis-charge valves and Na0li tank - -

suction valves) .

(1) Time constants utilized in lead lag controller for steamline pressure-l'ow are as follow '

f TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Functional Unit Trip Setpoint Allowable Value

1. SAFETY INJECTION, REACTOR TRIP, FEEDWATER ISOLATION, CONTROL ROOM ISOLATION, START DIESEL GENERATORS, CONTAINMENT COOLING FANS AND ESSENTIAL SERVICE WATER.
a. Manual Initiation NA NA
b. Automatic Actuation Logic NA NA
c. Reactor Building Pressure- 53.6 psig $3.86 psig High I
d. Pressurizer Pressure--Low 21850 psig 21839 psig
e. Differential Pressure 597 psig $106 psi Between Steamlines--High
f. Steamline Pressure--Low 2675 psig 1635 psig(1)
2. REACTOR BUILDING SPRAY
a. Manual Initiation NA NA
b. Automatic Actuation Logic NA NA and Actuation Relays
c. Reactor Building Pressure- $12.05 psig $12.31 psig High 3 (Phase 'A' isolation aligns spray system discharge  :

valves and NaOH tank suction valves)

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

It 2 50 secs. I2s 5 secs.

vs c -

TABLE 3.3-4 (Continued) ,

I ~

-ENGINEERED SAFETY' FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS .

g

-q- Functional Unit f

[{otal AT h ance (TA) 'Z S_ Trip Setpoint Allowable Value.

$ 3. 'CONTAINHENT ISOLATION

a. , Phase "A" Isolation I

l .- Manual A NA NA NA 'NA 2.

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

3. AutomaticActuationLogic[ NA NA -

NA ..

i and Actuation' Relays L

.b. Phase "B". Isolation '

, 1.' Automatic Actuation NA NA NA NA NA g Logic and Actuation

., Relays .

~

h 2. Reactor Building . 3. 0.71 .5 312.05 psig <12.31 psig Pressure-High 3 y .:

c. Purge and'E'haust x Isolation -
1. Safety Injection See J. above for all safety injection setpoints and allowable values
2. Containment Radioactivity NA * '*

High

3. Automatic Actuation . NA NA NA NA . NA

. '6" -

Logic sad Actuation .- -

Relays

[

g -

a- .

-*

  • Trip setpoints shall be set to ensure that the limits of Specification ~3.11'.2.1 are not exceeded.

f

. 03 % siz e 1. 2. 2. i - I ,.

~

- - _ . . . . . . . , . - , ,_ .+ ': .L

?

TABLE 3.3-4 (Continued)

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

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 _

injection setpoints and allowable injection setpoints and allowable values values

2. Containment Radioactivity * * '

High

3. Automatic Actuation Logic NA NA and Actuation Relays
  • Trip setpoints shall be set to ensure that the limits of ODCM Specification 1.2.2.1 are not exceeded.  ;

b-E TABLE 3.3-4 (Continued) 9 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP Functional Unit e (TA) Z Trip Setpoint

4. Allowable Value Sli M LINE ISOLATION
a. Manual i NA -

NA NA t NA NA

b. Automatic Actuation Logic and Actuation Relays NA k NA NA i NA NA

\

c. Reactor Building Pressure- \

3.0 g O. I 1.5 High 2 $6.35 $6.61

d. Steam Flow in Two Steamlines- 20.0

)t / \

High, Coincident with 113.16 1.5/ < a function < a function defined w ( l. 5 defined as as follows: A ap 2

l4 follows: A AP corresponding to 44%

m 't corresponding of full' steam flow 4

3 to 40% of full between 0% and 20%

t, steam flow load and then a op

?,

'- between 0% and increasing linearly 20% load and to a op corre-3 then a op sponding to 114.0%

i i increasing of full steam linearly to a

{ ap correspond-flow at full load. 1 i ing to 110% of a

full steam flow

, at full load avg - Low-Low

,F, T 4.0 R e. . 71 1 .8 3552.0 F 1548.4"F Steamline Pressure - Low -

20 0 10.71 .5 l

-8

s 3675 psig 1635 psig

.A f (1) m Time t 3 50constants secs. utilized in lead lag controller for steamline pressure low are as follows:

O r2 1 5 secs.

e

.---------u --------.-----___-------,-,----._-,--a- - - - - - - - - - - - - - --,-----a-------,, w - - - w- - -- - a _- ---L__--- - - - , -

f..

TABLE 3.3-4 (Continued)

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

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:

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%

load and then a op increasing load and then a op increasing linearly to a op 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 2635 psig(l) i

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

1-2 50 secs.

1 .I2 5 5 secs.

i

.c _s

.i TABLE 3.3-4 ,

i Functional Unit s tal Allowance (TA) Z C S Trip Setpoint Allowable Value  !

25 5. TURBINE TRIP AND FEEDWATER #

~d / /

w ISOLATION j r

/  !

a. Steam Generator Water 5.0 Level - High-High 2,18 1.5 182.4% of
  • $84. 2% o f narrow :  :

narrow range range instrument instrument span span *

6. EMERGENCY FEEDWATER
a. Manual NA  ! . ,

NA NA NA

, NA b.

t' c.

Automatic Actuation Logic Steam Generator Water Level - Low-Low NA 12.0

/ NA 9.18 NA

1. 5 NA

>12% of span NA

//,2$ ,*

~>ID=23 of span from o

from 0% to 30% 0% to 30% RTP in .

S$ RTP increasing creasing linearly to linearly to of span from

>30.0% of span _30% o 100% RTP-

  • from 30% to 100%

RTP- 2.

d. & f. Undervoltage-ESF Bus

>5760 Volts with >5652 Volts with a a <0.25 second ' <0.275 second time time delay delay

g. _

>6576 volts with >6511 volts with a

. i $3.0 second 33.3secondtime-n time delay delay

-9 u

.[

b

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

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

-l TABLE 3.3-4 (Continued) l' ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS l

Functional Unit Trip Setpoint Allowable Value

5. TURBINE TRIP AND FEEDWATER ISOLATION l a. Steam Generator Water $82.4% of narrow range instrument 184.2% of narrow range instrument Level - High-High span span
6. EMERGENCY FEEDWATER
a. Manual NA NA
b. Automatic Actuation Logic NA NA
c. Steam Generator Water 312% of span from 0% to 30% 311.2% of span from 0% to 30T%

Level - Low-Low RTP increasing linearly to 230.0% RTP increasing linearly to 229.2%

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

d. & f. Undervoltage-ESF Bus 15760 Volts with a 10.25 second 25652 Volts with a 50.275 second l time delay time delay '

16576 Volts with a 53.0 second 26511 Volts with a 53.3 second time delay time delay I

)

1

v, j

m TABLE 3.3-4 (Continued) 7 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPO

$ m

-4 w

Functional Unit Allowance (TA) Z QS , Trip Setpoint Allowable Value

e. Safety Injection y .e Se 1 above (all SI Setpoints >

I

g. Trips of Main Feedwater i NA NA NA NA Pumps NA

, f

h. Suction transfer on Low Pressure NA

\ NA NA } 1442 ft. 4in.( } 1441 ft. 3 in.

7. LOSS OF POWER f \

w S

a. 7.2 kv Emergency Bus Undervoltage (Loss of NA \ NA NA >5760 volts with >5652 volts with a w Voltage) a 50.25 second 30.275secondtime w ..,

time delay delay E b. \

7.2 kv Emergency Bus Undervoltage NA r

\ NA NA 16576 volts 16511 volts with a

\ with a $3.0 $3.3 second time second time delay f , delay r {

8. AUTOMATIC SWITCHOVER TO CONTAINMENT SUMP q / k

/ '

a.

l '

RWST Level Low-Low f NA NA NA ' 318% 11S%

b. Automatic Actuation Logic NA NA NA NA and Actuation Relays NA (2)

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

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Functional Unit Trip 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) 1441 ft. 3 in.

Pressure

7. LOSS OF POWER
a. 7.2 kv Emergency Bus 15760 volts with a 20.25 second 25652 volts with a 20.275 second Undervoltage (Loss of Voltage) time delay time delay
b. 7.2 kv Emergency Bus 16576 volts with a 53.0 second 26511 volts with a 53.3 second Undervoltage time delay time delay
8. AUTOMATIC SWITCH 0VER TO CONTAINMENT SUMP
a. RWST Level Low-Low 218% 215%
b. Automatic Actuation Logic NA NA and Actuation Relays l

l r

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

f 1

u, c-TABLE 3.3-4 (Continued)

' ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION T

E Functional Unit Z hotal ilowance (TA) Z S Trip Setpoint- Allowable Value

~

w 9. ENGINEERED SAFETY FEATURE

'\ ,e

[

ACTUATION SYSTEM INTERLOCKS '. -

.. I ,

, 1 INTERLOCKS i

a. Pressurizer Pressure, P-11 3.1 A N . 71 1. 5 1985 psig 11974 psig &

,e

/ 's ,

1

<1996 psig b.

avg Low-Low, P-12 I

4.0 / .8

. 71 \, - 552 F 1548.4 F & <555.6 F w c. Reactor Trip, P-4 2 i NA

. Y $.NA NA

= J a

a i

2 et

.O

TABLE 3.3-4 (Continued)

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

9. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS INTERLOCKS
a. Pressurizer Pressure, P-11 1985 psig 31974 psig &

5 1996 psig

b. Tavg Low-Low, P-12 552*F 1548.4*F &

$555.6*F

c. Reactor Trip, P-4 NA NA l

l l

l l

.I i -

3/4.3 TNSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 SYSTEM INSTRUMENTATIONREACTOR TRIP AND ENGINEERED SAFETY FE Feature Actuation System Instrumentation and interlocks en associated action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reiliability 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 of Service Times report. for Reactor Protection Instrumentation System," and supplements to that Surveillance intervals and out of service times were determined based on maintaining an appropriate 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 i within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational i 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 an allowance has been made in the safety incl d analysis to accommodate,this error. . M cptic M pr::i:i W e bee for etermini the OPE ILITY a chan when it rip se nt is ound t exceed Allowa Value he me dology of is opt lizes e "as sured" iation om the ecified rack sensor omponent in can' ction ibrati point fo the. er un taintie f the a stat ical ce nation o trumen ion to mpusure t rocess va able a the un taintie in cal ating instrumptation.

+R+ Equatio s.3-1, t TA, t inter ve eff s of th rrors i he rack the sensor and the s mea ed" va of th rors a considere . Z, as

/

e SUPEER - UNIT 1 B 3/4 3-1 Amendment No. 101

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 be 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 I 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 of 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 appropriate level of reliability of the Reactor Protection i 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 i 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

INSTRUMENTATION BASES REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (continueo) specified in i e 3.3-4 n percen an, is statis al summa n of errors assu in the alysis ex ding tho associa with th ensor a rack drif nd the uracy of eir measu nt. T or Total owance the dif ence, i ercent s. , betwee e trip tpoint a the valu used in t analysis or the at tion. R Rack or is the %s measur '

de tion, in/ percent sp , for the facted annel fr 'hespecj/iedtrip point. for Sensor tror is '

er the as measure deviati of the sensor fpom its cal ation po or th alue spec ed in T e 3.3-4, perceny span, fr the analy s assu ons. Usee f Ecuati 3.3-1 al s for as or drift ctor, an ' crease ack drif v e for RE TABLE EV S.

ctor,an/5rovides[ threshold The d hodology to derive the trip setpoints is based upon c'ombining 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 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 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 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 Actuation System senses selected plant 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 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. 25, 101

{

w .

INSTRUMENTATION BASES REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (continued)

I 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 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 sequential, over16pping 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 parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic catrices 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 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. 35, 101

Enclosure 2'to Document Control Desk letter TSP 930002 Page 1 PROPOSED TECHNICAL SPECIFICATION CHANGE REQUEST - TSP 930002 VIRGIL C. SUMMER NUCLEAR STATION DESCRIPTION AND SAFETY EVALUATION DESCRIPTION OF AMENDMENT REQUEST Existing Technical Specification (TS) Table 2.2-1, " Reactor Trip System ,

Instrumentation Trip Setpoints," and TS Table 3.3-4 " Engineered Safety Feature Actuation System Instrumentation Trip Setpoints," present setpoint information for each function in a five column format. The five columns of information are:

Total Total Allowance (TA) is the difference, in percent Allowance: instrument span, between the nominal trip setpoint and value used in the safety analysis limit'for the trip setpoint.

Z: Z, in percent span, is the statistical summation of errors assumed in the analysis, excluding those associated with the sensor and rack drift and the accuracy of their measurement.

S: S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in the table, in percent span, from the analysis assumptions.

Trip Setpoint: Nominal value at which the trip is set.

Allowable Allowable Value is a value chosen to accommodate the Value: instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated. 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 accommodate this error.

The five column format included provisions which sometimes eliminated the need for formal reporting through a Licensee Event Report (LER). The issuance of 10 CFR 50.73 changed the filing requirements associated with a LER when an Allowable Value was exceedcd. According to 10 CFR 50.73, filing a LER would not be required in response to the loss of a single channel; only-a loss of a function would require a LER. Therefore, the benefit of the five column methodology was no longer needed to prevent filing a.LER. l The Trip Setpoints in TS Table 2.2-1 prevent the reactor core and reactor coolant system from exceeding their safety limits during normal operation and design basis operational occurrences and assist the Engineered Safety I

Enclosure 2 to Document Control Desk Letter TSP 930002 Page 2 Features (ESF) Actuation System in mitigating the consequences of accidents.

The Trip Set;;oints for the ESF Actuation System are presented in TS Table-3.3-4. The setpoints, in accordance with the Allowable Value, provided in TS Table 3.3-4 cnsure that the consequences of Design Basis Accidents (DBAs) will be acceptable, provided the unit is operated from within the Limiting Condition for Operation (LC0) at the onset of the DBA and the equipment functions as designed. These Technical Specifications provide the setpoint information needed to determine the setpoint operability of the trip function.

The current VCSNS Model 03 steam generator narrow range water level Process MeasurementAccuracy(PMA)uncertaintyandprotectionsystemsetpointshave been recalculated to account for additional PMA uncertainties. PMA uncertainties are based on the type of measurement done but are not directly related to the accuracy of the measurement device; however, overall instrument channel accuracy is affected.

For Function 13 and Function 14 of TS Table 2.2-1 and Function 6.c of TS Table 3.3-4 (steam generator water level low-low), the recalculated PMA uncertainties result in changes to the Allowable Value column. Specifically, the Allowable Value is changed from 210.2% to 211.2% of span from 0 to 30%

RTP increasing linearly and from 228.2% to > 29.2% of span from 30% to 100%

RTP.

SAFETY EVALUATION The current VCSNS Model D3 steam generator narrow range water level Process Measurement Accuracy (PMA) uncertainty and protection system setpoints have been recalculated to account for additional uncertainties that were not in the original PMA value. PMA uncertainties are based on the type of measurement taken, but are not directly related to the accuracy of the measurement device; however, overall instrument channel accuracy is affected.

Previously, a random value of e.0% of span was used for the PMA uncertainty in the setpoint uncertainty calculations for all steam generator design models. This value was based on the density variation as a function of power and level and the assumption that calibration was done for 50% power conditions. For several steam generator models, a fluid velocity effect was known to introduce a significant bias in the low direction that was incorporated into the protection system setpoints.

Improved understanding of AP level measurement system uncertainties (G. E.

Lang and J. P. Cunningham, " Delta-P Level Measurement Systems,"

Instrumentation, Controls, and Automation in the Power Industry, Vol. 34, Proceedings of the Thirty-Fourth Power Instrumentation Symposium, June 1991),

has led to a reinvestigation of the Steam Generator Level PMA uncertainties.

The conclusions are that two additional uncertainty components must be accounted for explicitly (i.e., reference leg temperature changes from calibration temperature and downcomer subcooling) and that fluid velocity effects must be considered for all steam generator models. These uncertainty

Enclosure 2 to Document Control Desk Letter TSP 930002 Page 3 components are not considered to be random in nature and must be treated as biases.

A protection system setpoint study was originally done by Westinghouse to determine the instrument uncertainties for all Reactor Trip and ESFAS protection functions (WCAP-11770). Neither the safety analysis limit (SAL) nor the TS low-low steam generator water level setpoint require a change.

Although no changes are required to the Trip Setpoint column of the TS, changes are required to Allowable Value columns of TS Table 2.2-1, " Reactor Trip System Instrumentation Trip Setpoints," and TS Table 3.3-4, " Engineered Safety Feature Actuation System Instrumentation Trip Setpoints," to accommodate the additional PMA uncertainties associated with the narrow range steam generator water level setpoints. The proposed change to TS Tables 2.2-1 and 3.3-4 present setpoint information for each function in a two column format. The proposed two-column format of the TS tables provide setpoint information used to determine the setpoint operability of the trip function. The two columns of information are:

Trip Setpoint: Nominal value at which the trip is set.

Allowable Allowable Value is a value chosen to accommodate the Value: instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated. 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 accommodate this error.

An integrated safety evaluation was prepared to review the following areas to determine the effect of additional PMA uncertainties for the narrow range steam generator water level protection setpoints:

LOCA and LOCA-Related Accidents Non-LOCA Accidents Steam Generator Tube Rupture Containment Integrity Setpoint Evaluation Technical Specifications Fluid and Auxiliary Systems Performance Mechanical Systems Performance I&C Systems Performance Equipment Qualification Radiological Consequences Emergency Operating Procedures Component Performance Probabilistic Risk Analysis l

The additional PMA uncertainties impact the Protection System Setpcint j calculations, the excessive feedwater fics malfunction event (non-LOCA

Enclosure 2 to Document Control Desk Letter TSP 930002 Page 4 accidenti and the VCSNS Technical Specifications. .No other areas were affecteo 4 the additional PMA uncertainties for the narrow range steam generator water level protection functions.

An evaluation of the protection system setpoints provided the following narrow range steam generator water level PMA uncertainties for the Model D3 steam generators for the low-low level protection functions:

Process Pressure Variations + 0.3 % spen Reference Leg Temperature Variations + 0.4 % span Fluid Velocity Effects 0.0 % span Downcomer Subcooling Effects + 0.6 % span Total PMA Effects + 1.3 % span Using these values and the Westinghouse statistical setpoint methodology, the narrow range steam generator water level protection system setpoints were recalculated. The study shows that the narrow range steam generator water level low-low Technical Specification and SAL trip setpoints are acceptable; however, changes to the values presented in the Technical Specification Allowable Value column are required.

On the VCSNS Model D3 steam generators, the following are the elevations of the key dimensional parameters:

Lower Narrow Range Tap: 333 inches above tubesheet ( 0% of span)

Upper Narrow Range Tap: 566 inches above tubesheet (100%ofspan)

Mid Deck Plate Elevation: 545 inches above tubesheet ( 91% of span)

The steam generator water level low-low nominal trip setpoint is maintained at the present value of 12% of span between 0% and 30% power, then linearly increasing to 30% of span at full power. This setpoint is acceptable from an operational standpoint based Lpon plant operating experience.

The effect of additional PMA uncertainties for the VCSNS Model D3 steam generator narrow range water level protection setpoints has been evaluated against the standards of 10CFR50.59 and does not represent an unreviewed safety question based on the following justification.

1. Will the probability of an accident previously evaluated in the FSAR be ,

increased?

No. The probability of accidents previously evaluated in the FSAR will not be increased. There is no change in the affected protection function response times. Although there is an increase in instrument channel uncertainties, these are not initiators of any transient, and the analysis shows that the FSAR acceptance criteria for the postulated design basis events have been satisfied. Therefore, the probability of occurrence is not affected.

n Enclosure 2 to Document Control Desk Letter .

TSP 930002 Page 5 l

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2. Will the consequences of an accident previously evaluated in the FSAR be increased? '

No. The consequences of sccidents previously evaluated in the FSAR will not be increased. The affected protection functions will respond l

within the assumed times. The evaluation indicates that although-the '

instrument channel uncertainties increases, the FSAR acceptance criteria for the postulated design basis events h;ve been satisfied.

Since it has been concluded that the transient results are unaffected by this parameter modification, it is concluded that the consequences  ;

of an accident previously evaluated are not increased.

3. May the possibility of a different accident than already evaluated in the FSAR be created?

No. There is no significant possibility of creating an accident which is different than any already being evaluated in the FSAR. The affected protection functions will respond in a manner consistent with the current FSAR analyses. The change to the Allowable Value Column does not create the possibility of a new or different kind of accident from any accident previously evaluated, and does not affect the assumed accident initiation sequences. No new operating configuration is being  ;

imposed by the change to the Allowable Value Columns that would create  ;

a new failure scenario. In addition, no new failure modes are being '

created for any plant equipment. Therefore, the types of accidents defined in the FSAR continue to represent the credible spectrum of events to be analyzed which determine safe plant operation.

4. Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

No. There is no significant increase in the probability of a previously evaluated malfunction of equipment important to safety in the FSAR. There is no additional hardware introduced into the protection system. The instrument channel uncertainties increased for the new PMA uncertainties increased. The evaluation demonstrates that the FSAR acceptance criteria for the postulated design basis events have been satisfied. The changes to the Allowable Value Column will not adversely affect system performance or safety system functions assumed in the accident analyses. .The original design specifications such as for seismic requirements, electrical separation, and environmental qualification are unaffected. The revised setpoints will ,

not adversely affect the operation of the Reactor Protection System or '

any other device required for accident mitigation. In addition, no new failure modes are being created for any plant equipment. Therefore, probability of a malfunction of equipment-important to safety previously evaluated in the FSAR will not be increased.

f Enclosure 2 to Document Control Desk Letter TSP 930002 Page 6

5. Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

No. The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. The current failure modes as analyzed are unchanged. With no change to the analyzed failure modes, the FSAR acceptance criteria for the postulated design basis events remain satisfied. The changes to the Allowable Value Column will not adversely affect the ability of existing components and systems or the integrity of the fission product barriers to mitigate the radiological dose consequences of any accident. Both the margin to DNB and fuel temperature limits remain protected. In addition, the offsite dose predictions previously assumed are unaffected and remain within the acceptance criteria. Therefore, the radiological consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not increase.

6. May the possibility of a malfunction of equipment important to safety be different than any already evaluated in the FSAR be created?

No. There is no significant possibility of creating a malfunction of equipment important to safety different than any already evaluated in the FSAR. There is no change to hardware or plant procedures as a result of this evaluation. Operation remains consistent with the FSAR assumptions. All original design and performance criteria continue to be met, and no new failure modes have been created for any system, component, or piece of equipment. No new single failure mechanisms have been introduced nor will the core operate in excess of pertinent design basis operating limits. Therefore, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR has not been created.

7. Will the margin of safety as described in the bases to any technical specification be reduced?

No. The margin of safety as defined in the Bases of the Technical Specifications will not be reduced. All hitial conditions of the FSAR with respect to steam generator level are maintained, and the results of the FSAR remain valid. Therefore, the change in the Allowable Value Column does not involve a reduction in the margin of safety.

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Enclosure 3 to Document Control Desk letter TSP 930002 1 Page 1 PROPOSEb TECHNICAL SPECIFICATION CHANGE REQUEST - TSP 930002 VIRGIL C. SUMMER NUCLEAR STATION DETERMINATION OF N0 SIGNIFICANT HAZARDS CONSIDERATION DESCRIPTION OF AMEkDMENT REQUEST Existing Technical Specification (TS) Table 2.2-1, " Reactor Trip System Instrumentation Trip Setpoints," and TS Table 3.3-4, " Engineered Safety Feature Actuation System Instrumentation Trip Setpoints," present setpoint information for each function in a five column format. The five columns of information are:

Total Total Allowance (TA) is the difference, in percent instrument Allowance: span, between the nominal trip setpoint and value used in the safety analysis limit for the trip setpoint.

Z: Z, in percent span, is the statistical summation of errors assumed in the analysis, excluding those associated with the sensor and rack drift and the accuracy of their measurement.

S: S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in the table, in percent span, from the analysis assumptions.

Trip Setpoint: Nominal value at which the trip .is set.

Allowable Allowable Value is a value chosen to accommodate the Value: instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be_ measured and calibrated. 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 accommodate this error.

The five column format included provisions which in some cases eliminated the  !

need for formal reporting through a Licensee Event Report (LER). The .

l issuance of 10 CFR 50.73 changed the filing requirements associated with a l LER when an Allowable Value was exceeded. According to 10 CFR 50.73, filing a LER would not be required in response to the loss of a single channel;.only as a result of a function would a LEP, be required. Therefore, the benefit of the five column methodology was'no longer needed to prevent filing a LER. j The Trip Setpoints in TS Table 2.2-1 prevent the reactor core and reactor coolant system from exceeding their safety limits during normal operation and l design basis operational occurrences and assist the Engineered Safety Features (ESF) Actuation System in mitigating the consequences of accidents.

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Enclosure 3 to Document Control Desk Letter TSP 930002 Page 2 The Trip Setpoints for the ESF Actuation System are presented in TS Table 3.3-4. The setpoints, in accordance with the Allowable Value, provided in TS Table 3.3-4 ensure that the consequences of Design Basis Accidents (DBAs) will be acccptable, provided the unit is operated from within the Limiting Condition for Operation (LCO) at the onset of the DBA and the equipment functions as designed. These Technical Specifications provide the setpoint information needed to determine the setpoint operability of the trip function.

The current VCSNS Model D3 steam generator narrow range water level Process Measurement Accuracy (PMA) uncertainty and protection system setpoints have been recalculated to account for additional PMA uncertainties. PMA uncertainties are based on the type of measurement performed but are not directly related to the accuracy of the measurement device; however, overall instrument channel accuracy is affected.

For Function 13 and Function 14 of TS Table 2.2-1 and Function 6.c of TS Table 3.3-4 (steam generator water level low-low), the recalculated PMA uncertainties result in changes to the Allowable Value column. Specifically, the Allowable Value is changed from 210.2% to 211.2% of span from 0 to 30%

RTP increasing linearly and from 228.2% to 229.2% of span from 30% to 100%

RTP.

BASIS FOR DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Although no changes are required to the Trip Setpoint column of the VCSNS Technical Specifications, changes are required to other columns of TS Table 2.?-1, " Reactor Trip System Instrumentation Trip Setpoints," and TS Table 3.3-4, " Engineered Safety Feature Actuation System Instrumentation Trip  ;

Setpoints," to accommodate the additional PMA uncertainties associated with the VCSNS Model D3 steam generator narrow range water level setpoints. For the steam generator water level low-low trip function, a TS change is required to the Allowable Value columns.

Pursuant to 10 CFR 50.92 each application for amendment to an operating license must be reviewed to determine if the proposed change involves a significant hazards consideration. The amendment describing the technical specification changes associated with the additional PMA uncertainties for the narrow range steam generator water level protection functions has been reviewed.and deemed rot to involve significant hazards considerations. As discussed below, all applicable acceptance criteria are satisfied and the conclusions presented in the VCSNS FSAR remain valid. Thus, the proposed Technical Specification changes do not constitute an unreviewed safety question and the accident analyses support the changes. The basis for this determination follows:

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Enclosure 3 to Document Control Desk Letter l TSP 930002 Page 3

1. Operation of VCSNS in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The probability or consequences of accidents previously evaluated in the FSAR will not be increased. There is no change in the affected protection function response times or the Technical Specification Trip Setpoint. The changes to the Allowable Value column for the steam .

generator water level low low reactor trip function do not invalidate the design basis acceptance criteria for the transients evaluated.

The narrow range steam generator water level setpoints are part of the accident mitigation response and are not themselves an initiator for any transient. Since it has been concluded that the transient results are unaffected by the parameter modifications, it is concluded that the probability or consequences of an accident previously evaluated are not increased.

2. The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The changes to the Allowable Value columns for the steam generator water level low-low reactor trip function will not introduce any new accident initiator mechanisms. The affected protection functions will respond in a manner consistent with the current FSAR analyses. No new failure modes or limiting single failures have been identified.

Furthermore, the setpoint adjustment does not affect the assumed accident initiation sequences. In addition, no new operating configuration is being imposed by the setpoint adjustment that would create a new failure scenario. Since the safety and design requirements continue to be met and the integrity of the reactor coolant system pressure boundary is not challenged, no new accident scenarios have been created. Therefore, in light of the above, an accident which is different than any already evaluated in the FSAR will not be created as a result of this change.

3. The proposed license amendment does not involve a significant reduction in a margin of safety.

Although the additional PMA uncertainties for the narrow range steam generator water level trip functions will require a change to the plant Technical Specifications,it will not invalidate the design basis j acceptanca criteria presented in the FSAR accident analyses. All  !

initial conditions of the FSAR with respect to the steam generator level are maintained, and the results of the FSAR remain valid.

Therefore, there is no reduction in the margin to safety as defined in the Bases of the VCSNS Technical Specifications.

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Enclosure 3 to Document Control Desk Letter TSP 930002 Page 4

6.0 CONCLUSION

I The various facets of the VCSNS licensing basis that are potentially affected i by the additional PMA uncertainties for the narrow range steam generator -l water level trip functions have been evaluated. All applicable acceptance l criteria are satisfied. In conclusion, while Technical Specification changes '

are necessary to reflect the additional PMA uncertainties for the narrow range steam generator water level trip functions, all the conclusions presented in the FSAR remain valid, l

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