ML20070J788

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Proposed TS Pages for TS Tables 2.2-1, Reactor Trip Sys Instrumentation Trip Setpoints & 3.3-4, ESFAS Instrumentation Trip Setpoints
ML20070J788
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Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 07/20/1994
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SOUTH CAROLINA ELECTRIC & GAS CO.
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NUDOCS 9407250317
Download: ML20070J788 (48)
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3 1:

. SAFETY LIMITS AND LTMITING 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 oi Table 2.2-1 adjust the setpoint consistent with the Trip Setpoint value. i

b. With the reactor trip system instrumentation or interlock setpoint less conservative than the value shown in the Allowable Values column of .-

.rrmo ....J!ti;. ith'- 1 L;_., !i Table,2J-1 pig t h ; ......... ; :.. w.

w...

e r. . . . . . . . .. . . . . v n iv u. nuui . w i me s .

wea'auwse wu .s w. sw u uwws w a sus ta is K. us ws s m ilit blin k hhgun w i via .

I 3.d:dgtth: ::tp f-t :: :':t:-t

'th th; T.;r i t,..ic.t ..l.s vi -

. clare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

EQUAfl0N 2.2-1 Z . R . '; i TA where:

Z= the value for co Z of Table 2.2-1 f a- fected channel, R= the "as measured" value ( rc tian)ofrackerrorforthe affected channel, S= either the "as red" value (in percent ) of the sensor error, or value is column S of Table 2.2-1 he affected chan s , and i L the value frc: column- TA of Tabla 2.2-1 fer-the-ef-feeted-chann . s li l

l' 1:

t.

SUt94ER - UNIT 1 2-4 l

9407250317 DR 940720 ADOCK 05000395  !

PDR  !

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1

2.2 LIMITING SAFETY SYSTEM SETTINGS l 1

REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS l 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 channelin Table 3.31.

ACTION:

a. With a reactor trip system instrumentation or interlock setpoint less conservative than the value shown in the Tria Setpoint column of Table 2.21 adjust the setpoint consistent wit 2 the Trip Setpoint value.

1

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.

I l

l l

l l

l SUMMER - UNIT 1 2-4 Amendment No.

m G

o TABLE 2.2-1 i

REACTOR 1 RIP SYSTEM INSTRUMENTATION TRIP SETPOINTS U

H W

[ Functional Unit n ; ;-, ,_ '"1 Z  ; irlp Setpoint Allowable Value

1. Manual Reactor Trip II Applicable NA I NA NA
2. Power Range, Neutron Flux High Setpoint .5 4.56 0 '

Low Setpoint 5109% of RTP 5111.2% of RTP t l.3 4.56 0 $25% of RTP $27.2% of RTP

3. Power Range, Neutron Flux .6 High Positive Rate 0.5 C 55% of RTP with a time 56.3% of RTP with a time constant 12 seconds constant 22 seconds
4. Deleted

't S. Intermediate Range, 17.0 .4 Neutron Flux C 525% of RTP $31% of RTP

6. Source Range, Heutron Flux 7.0 10.0 0 5 105 cps 51.4 x 105 cps
7. Overtemperature AT  : 4.7 2.2 1, 5 See note 1 See note 2

& l. 3**

8. Overpower AT .1
2. 1. 5 See note 3 See note 4 9.

l Pressurizer Pressure-Low .1 0.7 1. i 21870 psig 21859 psig

10. Pressurizer Pressure-liigh .9 4

5.0 0.i: $2380 psig 52391 psig a

R 11. Pressurizer Water Level-liigh ' .0 2.18 g .

592% of instrument $93.8% of instrument span span E

2 12. Loss of Flow iJ 1. '/'

L g .o ,

' *. 5 190% of loop design 288.9% of loop design

,e flow * - -

flow

  • C"
  • Loop desi n flow = 94.500 gpm N" 1' RTP 6 RAT D THERM 4LtPOWER wa

{ : . y' r...

. o.' a.6-T isi.,/, _..J !.5 f:: "r:::1.':;. ".. __. -

TABLE 2.2-1 -

m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS C

! Functional Unit Trio Setpoint Allowable Value 7

! 1. Manual Reactor Trip NA NA 1

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 15% of RTP with a time <6.3% of RTP with a time High Positive Rate constant >2 seconds constant >2 seconds
4. DELETED
5. Intermediate Range, Neutron Flux <25% of RTP <31% of RTP ,

y 6. Source Range, Neutron Flux < 105 cps < 1.4 x 105 cps

7. Overtemperature AT See note 1 See note 2
8. Overpower AT See note 3 See note 4
9. Pressurizer Pressure-Low >1870 psig >1859 psig
10. Pressurizer Pressure-High $2380 psig $2391 psig
11. Pressurizer Water Level-High 192% ofinstrument span 193.8% ofinstrument span
12. Loss of Flow >90% ofloop design flow * >88.9% ofloop design flow
  • 1[

t ll

  • F, F

h

  • Loop design flow = 94,500 gpm l y RTP -RATED THERMAL POWER '

y I

TABLE 2.2-1 (continued}

o

, REACTOR TRIP SYSTEM INSTRUMEllTATI0tj TRIP SETPOINTS 0

1 d 'otal Functional Unit Alowance(TA) Z S Trip Setpoint Allowable Value
13. Steam Generator Water Level Low-Low Barton Transmitter a .0 5.1 1. '

127.0% of span 226.1% of span Rosemount Transmitter i .0 5.1 1, 7 127.0% of span 125.7% of span

14. Steam /Feedwater Flow His-  : 6.0 13 4 1 5/ <40% of full <42.5% of full Match Coincident With 15 steam flow at RTP steam flow at RTP o Steam Generator Water Level a Low Barton Transmitter 7 .0 5.1 17 127.0% of span 326.1% of span Rosemount Transmitter 7 .0 5.1 17 227.0% of span 225.7% of span
15. Undervoltage - Reactor 2 .1 1.28 0.23 Coolant Pump 24830 volts 24/60 vot*s
16. Underfrequency - Reactor .5 0 01 Coolant Pumps a

->57.5 Hz >57.1 liz

17. Turbine Trip A. Low Trip System Pressure 1 / NA i, 1800 psig 1750 psig B. Turbine Stop Valve jk NA 31% open 31% open Closure i

h RTP - RATED THERMAL POWER a

n

. y. ;d; } :t

TABLE 2.2-1 (continued) -

m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

i Functional Unit Trio Setpoint Allowable Value E

y 13. Steam Generator Water

- LevelLow-Low Barton Transmitter > 27.0% of span >26.1% of span Rosemount Transmitter >27.0% ofspan >25.7% of span

14. Steam /Feedwater Flow Mis- 140% offull steam flow at RTP 142.5% of full steam flow at RTP Match CoincidentWith Steam Generator Water Level Low-Low Barton Transmitter >27.0% of span >26.1% of span Rosemount Transmitter >27.0% of span >25.7% of span m 15. Undervoltage - R.eactor >4830 volts >4760 volts .

& Coolant Pump

16. Underfrequency- Reactor > 57.5 Hz > 57.1 Hz Coolant Pumps
17. Turbine Trip A. Low Trip System Pressure >800 psig >750 psig B. Turbine Stop Valve Closure >1% open >1% open E

8 a

n

',. RTP -RATED THERMAL POWER

I j TABLE 2.2-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

= ~

' Functional Unit

^

.! ;----..-- G .e i  ; Trip Setpoint Allowable Value h

18. Safety Injection Input NA > L NA NA

- from ESF

19. Reactor Trip System Interlocks A. Intermediate Range l A N h 4 >7.5 x 10 *% >4.5 x 10 8%

Neutron Flux, P-6 Indication Indication

8. Low Power Reactor Trips Block, P-7
a. P-10 input .5 4.56 ( $10% of RTP $12.2% of RTP m b. P-13 input .5 4.56 i <10% turbine <12.2% of turbine 4 Tapulse pressure Tapulse pressure i equivalent equivalent C. Power Range Neutron '5 .56 (

-<38% of RTP -<40.2% of RTP Flux P-8 D. Low Setpoint Power '5

. 4. E >10% of RTP >7.8% of RTP Range Neutron Flux, P-10 ~ ~

E. Turbine Impulse Chamber '. 5 4.56 e <10% turbine *<12.2% turbine Pressure P-13 Impulse pressure iiressure equivalent equivalent y F. Power Range Neutron .5 4.56 $50% of RTP Flux, P-9 $52.2% of RTP a .

g 20. Reactor. Trip Breakers l NA N NA NA

& 21. Automatic Actuation Logic j L ""

"" L NA NA g

RTP = RATED INERMAL POWER g

. - m. -_ ' _--___m_ _ . *_ _ _ _ m-

__ __ _- ________ _ _ _ - _ _ _ _ _ _ _ 2 ___ _ ._

TABLE 2.2-1 (continued) m RE ACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS S

I i Functional Unit Trio Setpoint Allowable Value

18. Safety InjectionInputfrom ESF NA NA

~

19. Reactor Trip System Interlocks A. Intermediate Range Neutron Flux, P-6 27.5 X 10S% indication 24.5 X 10A % indication B. LowPowerReactorTrips  !

Block, P-7 -

a. P-10 input <10% of RTP < 12.2% of RTP ,
b. P-13 input 110% turbine impulse pressure 112.2% of turbine impulse I equivalent pressure equiv& lent C. Power Range Neutron Flux P-8 <38% ofRTP <40.2% of RTP D. Low Setpoint Power Range Neutron Flux, P-10 210% ofRTP 27.8% of RTP E. Turbine Impulse Chamber $10% turbine impulse pressure 112.2% turbine pressure Pressure, P-13 equivalent equivalent F F. Power Range Neutron Flux, P-9 < 50% of RTP < 52.2% of RTP m ,

h 20. Reactor Trip Breakers NA NA E 21. Automatic Actuation Logic NA NA E.

m i

y RTP - RATED THERMAL POWER

I

%Chy O* ~

m l %C. \ -

o e

~ TABLE 2.2-1 (Continued)

E e REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

- NOTATION NOTE 1: OVERIEMPERATURE AT (1 + t,S) ,

ATsAT, K, - K,, T-T 4 K 3( P - P') - f(AI) 3 Where: AT =

Measured AT by RID Instrumentation AT, 5 Indicated AT at RATED THERMAL POWER Ki 5 1.23 l-y K 2 2 0.0292/*F

= l 1 + t,S

=

The funtion generated by the lead-lag controller s

I + '28 for T avg dynamic compensation i

si ,t2 =

Time constants utilized in lead-lag controller for T avg, t, t 28 secs.,

t 2-5 4 secs.

T = Average temperature. *F l T' 5 Indicated Tavg at RATED THERMAL POWER, 572.0*F s T's 587.4*F l

K3 1 -0.00161/ psi P =

Pressurizer pressure, psig y P' 3 2235 psig, Nominal RCS' operating pressure u li g S =

Laplace transform operator, sec-1 EW a

n

.O

m TABLE 2.2-1 (continued) i S g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 7 NOTATION C

l 3 NOTE 1: OVERTEMPERATURE AT

~

~

(1 + 1 1S) ,.

ATsAT K-K T-T +K3 ( P - P') - f,( AD o 1 2 (1 + t 3 3 Where: AT = Measured AT by RTD Instrumentation AT, s Indicated AT at RATED THERMAL POWER K 1 s 1.23 l K,~

2 0.0292/ F l l

m 1+xS1 a = The funtion generated by the lead-lag controller for T,,,

I + '28 dynamic compensation 11,12 = Time constants utilized in lead-lag controller for T,,,,1 1 2 28 secs.,

12 s4 secs.

T = Average temperature, *F T' s Indicated T,,, at RATED THERMAL POWER,572.0 F s T's 587.4*F l gg K3 2 0.00161/ psi l 4

,L R P = Pressurizer pressure, psig

'+* 9 ig P' 2 2235 psig, Nominal RCS operating pressure

.", S = Laplace transform operator, sec-1 1

m e

.lM '

W TABLE 2.2-1 (Continuedl

% ch^ :g [tm i

REACTOR TRIP SYSTEM INSTRUMENTATION 1 RIP SETPOINTS 9.l.m. j*

{ ~~ ~

s l

HOTATION (Continued) 1 NOTE 1: (Continued) and f (AI) is a function of the Indicated difference between top and bottom detectors of the power-range nuclea,r tests such ton that: chambers; with gains to be selected based on measured instrument response (1) for qt - 9b ,

between -35 percent and 46 percent f, (al) = 0 where qt and qb are percent RATED TilERMAL l POWER in the top and bottom halves of the core respectively, and qt + bAis total THERMAL POWER in percent of RATED THERMAL POWER.

(11) for each percent that the magnitude of qt - Ab exceeds -35 percent, the AT trip setpoint shall be l

[ automatically reduced by 2.46 percent of its value at RATED TilERMAL POWER.

(iii) 1 for each percent that the magnitude of qt ~ 9b exceeds +6 percent, the AT trip setpoint shall be l automatically reduced by 3.29 percent of its value at RATED THERMAL POWER.

NOTE 2: I The channel's Span. maximum trip setpoint shall not exceed its computed trip point.by more than 2.2 percent AT NOTE 3: OVERPOWER AT (t s) l  !

3 ATSAT a K-K 4 T-K T-T 3(1 + t.,S) 8 2

,i o

Where: AT =

as defined in Note 1 g AT, =

as defined in Note 1 3(n K, < 1.078

" l K

5 E

0.02/*f for increasing average temperature and 0 for decreasing average temperature 1, s

% =

8 The function generated by the rate-lag controller for T

f+5 3 compensation . .

9- >

'.. .. avg dynamic m .

si -i i .:

9 TABLE 2.2-1 (continued)

$ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (continued)

E NOTE 1: (Continued) 4

~

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

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

(i) for q,- q, between -35 percent and + 6 percent 1f (AI) = 0 where q, and q, are percent RATED THERMAL l POWER in the top and bottom halves of the core respectively, and q, + q, is total THERMAI., POWER in percent of RATED THERMAL POWER.

(ii) for each percent that the magnitude ofq,- q, exceeds -35 percent, the AT trip setpoint shall be l automatically reduced by 2.46 percent ofits value at RATED THERMAL POWER. 1 (iii) for each percent that the magnitude of q,- q, exceeds + 6 percent, the AT trip setpoint shall be l 1 automatically reduced by 3.29 percent ofits value at RATED THERMAL POWER. l 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

- I 6 3s)

AT SAT, K-K, 4 T - K, T- T q}

$E 4 m

- ~

{ Where: AT = as defined in Note 1 7(( AT, = as defined in Note 1 g K4 s 1.078 l K3 2 0.02/"F for increasing average temperature and 0 for decreasing average temperature 1

1* S

= The function generated by the rate-lag controller for T,,, dynamic I+1 3 S compensation

. __l

a 4

0 0

n TABLE 2.2-1 (Continuedl g REACTOR TRIP SYSTEM INSTRUMENTATION 1 RIP SETPOINTS I

NGTATION (Continued)

NOTE 3:

(continued)q 8

f f3

=

Time constant utilized in rate-lag controller for T avg > 53 1 10 secs.

K, t 0.00198/*F for T > T~ and gK = 0 for I 5 T~ l T =

as defined in Note 1

,o T~ $ Indicated Tavg at RATED THERMAL POWER, 572.0*F s T'#s 587.4*F l Es S =

as defined in Note 1 NOTE 4: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.3 percent AT Span.

l 8

8&

8 n

r

.+

m TABLE 2.2-1 (continued) .,

'! REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS f

c:

NOTATION (continued)

} NOTE 3: (continued)

.- 13 = Time constant utilized in rate-lag controller for T,,,,1 3 210 secs. l 1

K, 2 0.00198/*F for T > T~ and K, = 0 for T s T~ l T = 'as defined in Note 1 T~ s Indicated T,,, atRATED THERMAL POWER,572.0*F s T"5 587.4 F 1 Y S = as definedin Note 1 5

NOTE 4: The channel's maximum trip setpoint shall not exceed its computed trip point by more than  ;

. 2.3 percent AT Span . I  ;

5 E.

R i

E

[ E

=

1

'l t

l No My h%

2.1 SAFETY LIMITS

    • T%m \.

BASES

_2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and products cladding possible perforation to the reactor which would result in the release of fission coolant.

Overneating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could i result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation arai therefore related to ONB. THERMAL POWER and Reactor Coolant Temoerature and This relation has been developed to predict the DNB flux and Press the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux ratio (DNBR) defineo as the ratio of the heat flux thi would cause DNB at a particular indicative of the margin to DNB. core location t'o the local heat flux, is l l

The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum ON8R of the limiting rod during Condition I and II l

events is greater than or equal to the ONBR limit of the ONB correlation being used. l The correlation DNBR limit is established based on the entire applicable  !

experimental data set such that there is a 95 percent probability with 95 l '

percentlimit.

DNBR confidence that DNB will not occur when the minimum DNBR is at the 1

In meeting this design basis, uncertainties in plant operating parameters, l nuclear and thermal parameters, and fuel fabrication parameters are considered  !

statistically such that there is at least a 95 cercent probability with 95 percent i confidence equal to the level ONBR that the minimum ONBR for the limiting rod is greater than or -

limit. I The uncertainties in the above plant parameters are-used to determine the plant DNBR uncertainty. This ONBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.  !

In addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses. l The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressere and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

SUMMER - UNIT 1 B 2-1 Amencment No. 75

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented ay restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive claddin v  ;

from nucleate boiling (DN $)and temperatures the resultantbecause of the onset sharp reduction in heatof departure transfer I coefficient. DNB is not a directly measurable parameter during operation and l therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

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

The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Conc ition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used. The correlation DNBR limit is established based on the entire applicable l experimental data set such that there is a 95 percent probability with 95 oercent confidence that DNB will not occur when the minimum DNBR is at the JNBR limit. l In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability with 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DN BR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values ofinput parameters without uncertainties.

In addition, margm has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.

The curves of Figure 2.1-1 show the loci of points of THERM AL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

SUMMER - UNIT 1 B 2-1 Amendment No. %,

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 l Setpoints have been selected to ensure that the reactor core and reactor I 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 Syr, tem in mitigating the consequences of accidents. The setpoint for a reactor trip system or interlock function  !

l 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 l in the safety analysis to accommodate this error. 3A...g..m.a e. v v m v.. l m n included for cetermining the OPtKASIL11Y of a channel when its trip . .~ 3 iet t is found to exceed the Allowable Value. The methodology of sption izes the "as measured" deviation from the specified cal ation

>oint for r and sensor components in conjunction with a sta ical combin t-i

' ion of the othe neertainties of the instrumentation to m ure the process l

'ariable and the un ainties in calibrating the instr tation. In j

quation 2.2-1, Z + R + < TA, the interactive eff of the errors in the ack and the sensor, and the ' s measured" valu l f the errors are considered .

., as specified in Table 2.2-1, ercent s , is the statistical summation if errors assumed in the analysis ex d those associated with the sensor ind rack drift and the accuracy of t r l surement. TA or Total Allowance  ;

s the difference, in percent sga , between trip setpoint ar.d the'value ased in the analysis for re tir trip. R or Rack r is 'the "as measured" l

deviation, in percent , for the affected channel he specified trip ietpoint. 5 or S Error is either the "as measured" de on of the Lensor from alibration point or the value specified in Table - -1, in iercent , from the analysis assumptions. Use of Equation 2.2-1 a s for

< s r drift factor, an increased rack drift factor, and provides a thr Ic

._.a' "I"^"?"' E 9?E"! .

t -

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. Sensors and 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 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 3 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 SUMMER - UNIT 1 B 2-3 Amenoment No. 35 l

I

' j l

l 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 Set points 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 cont 'quences of accicients. 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 teste 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 accommodate 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, ma be indicative of more serious problems and should warrant further investi ation.

l l

1 i

SUMMER -UNIT 1 B 2-3 Amendment No. 66, l l

l l

l

= CA f%  !

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)-

The various reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the Reactor Protection System reacnes a preset or calculated level.

In addition to redundant channels and trains, the design approach provides a Reactor Protection System which monitors numerous system variables, therefore, providing protection system functional diversity.

The Reactor whenever reactor trip is Protection initiated. System initiates a turbine trip signal This prevents the reactivity insertion that would otherwise result from excessive reactor system cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trio The Reactor Protection System includes manual reactor trip capability.

Power Rance. Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a high and low range trip setting. The low setpoint trip provides protection during subcritical and low -

power operations to mitigate the consequences of a power excursion beginning from low power, and the high setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

The low setpoint trip may be manually blocked above P-10 (a power level-of approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated below the P-10 setpoint.

Power Rance. Neutron Flux. Hich Rates The Power Stange Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low  ;

l trips to ensure that the criteria are met for rod ejection from mid-power. I I

Intermediate and Source Rance. Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup to mitigate the consequences of an SUMMER - UNIT 1 B 2-4 Amendment No. 75 ~

l LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

The various reactor trip circu,its automatically open the reactor trip breakers whenever a condition monitored by the Reactor Protection System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Protection System which monitors numerous system variables, therefore, providing protection system functional diversity. 'fhe Reactor Protection System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive reactor system cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trip The Reactor Protection System includes manual reactor trip capability.

Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a high and low range trip setting. The low setpoint trip provides protection during suberitical and low power operations to mitigate t ae consequences of a power excursion beginning from low power, and the high setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

The low setpoint trip may be manually blocked above P-10 (a power level of approximately 10 percent of RATED TH8RM AL POWER) and is automatically reinstated below the P-10 setpoint.

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid-power.

Intermediate and Source Range, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup to mitigate the consequences of an SUMMER - UNIT 1 B 2-4 Amendment No. 76,

b CM  %

SL \,

LIMITING SAFETY SYSTEM SETTINGS BASES

{ %~

PressurizerPressure(Continued)

On decreasing power the low setpoint trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10 percent of full power equivalent)1 and on increasing power, automatically reinstated by P-7.

The high setcoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The pressurizer high water level trip is provided to prevent water relief through the pressurizer safety valves.

On decreasing power the pressurizer high water level trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10 percent of full equivalent); and on increasing power, automatically reinstated by P-7.

_ Loss of Flow The Loss of Flow trips provide core protection to prevent DNB by mitigating the consequences coolant pumps. of a loss of flow resulting from the loss of one or more reactor On increasing power above P-7 (a power level of approximately 10 percent of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10 percent of full power equivalent), an automatic reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow.

Above P-8 (a power level of approximately 38 percent of RATED THERMAL POWER) an automatic reactor trip will occur if the flow in any single loop drops below 90 percent of nominal full loop flow. Conversely on decreasing power between P-8 and the P-7 an automatic reactor trip will occur on loss of flow in more than one loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Level The steam generator water level low-low trip protects the reactor from loss of heat sink in the event of a sustaired steam /feedwater flow mismatch retulting from loss of normal feedwater. The specified setpoint provides 2

allowances for starting delays of the auxiliary feedwater system.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The steam /feedwater flow mismatch in coincidence with a steam generator low water level trip is not used in the transient and accident analyses but is -

included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than or equal to 40% of full steam flow at RTP. The Steam Ger.erator Low Water level portion of the trip is activated when the water level drops below the low l

SUMMER - UNIT 1 8 2-6 Amendment No. ,

i

LIMITING SAFETY SYSTEM SETTINGS l BASES l 1

Pressurizer Pressure (Continued) 1 On decreasing power the low setpoint trip is automatically blocked by P-7 l (a power level of approximately 10 percent of F.ATED THERMAL POWER with turbine i impulse chamber pressure at approximately 10 percent of full power equivalent); l and on increasing power, automatically reinstated by P-7. l The high setpoint trip functions in conjunction with the pressurizer  !

relief and safety valves to protect the Reactor Coolant System against system l overpressure. l Pressurizer Water Level l The pressurizer high water level trip is provided to prevent water relief i through the pressurizer safety valves. On decreasing power the pressurizer high ,

water level trip is automatically blocked by P-7 (a power level of approximately l 10 percent of RATED THERM AL POWER with a turbine impulse chamber pressure at {

approximately 10 percent of full equivalent); and on increasing power, j automatically reinstated by P-7. ,

i Loss of Flow The Loss of Flow trips provide core protection to prevent DNB by mitigating l the consequences of a loss of flow resulting from the loss of one or more reactor )

coolant pumps. i On increasing power above P-7 (a power level of approximately 10 percent l of RATED THERMAL POWER or a tur aine impulse chamber pressure at approximately i 10 percent of full power equivalent), an automatic reactor trip will occur if l the flow in more than one loop drops below 90% of nominal full loop flow.

Above P-8 (a power level of approximately 38 percent of RATED THERMAL POWER)  ;

an automatic reactor trip will occur if the flow in any single loop drops below 90 percent of nommal fullloop flow. Conversely on decreasing power between P-8 and the P-7 an automatic reactor trip will occur on loss of flow in more than one loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Level  ;

The steam generator water level low-low tri rotects the reactor from l loss of heat sink in the event of a sustained steamffeedwater flow mismatch  ;

resulting from loss of normal feedwater. The specified setpoint provides allowances for starting delays of the auxiliary feedwater system.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The steam /feedwater flow mismatch in coincidence with a steam generator low water level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor  ;

Protection System. This trip is redundant to the Steam Generator Water Level i Low-Low trip. The Steam /F eedwater Flow Mismatch portion of this trip is I activated when the steam flow exceeds the feedwater flow by greater than or equal to 40% of full steam flow at RTP. The Steam Generator Low Water level  ;

portion of the trip is activated when the water level drops below the low l SUMMER - UNIT 1 B 2-6 Amendment No.

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-ve than the value shown in the Allowable Value Column of Table 3.
1. bjett": fetra4at r n:i:t:nt _it, '

. T ,y $. mpo ;,,, . . L J--

T; m :.? d . ~ e:t........ ...... ::-;.. c : th:t a r ti r 2.2 0 n::: "+Mi-d 'r ::. o ficuou m. ... . . .d r ,

/.1/eclarethechannelinoperableandapplytheapplicableACTION statement requirements of Table 3.3.3 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

I EQUATION 2.2 1 e.

  • n + 3 i ta where:

Z = the value alumn Z of Table 3.3-4 f e affected channel, R = the "as measured" value nt span) of rack error for the ,

affected channel, 1 S = either t measured" value (in percen of the sensor I err ,

or the vg0ue in column S of Table 3.3-4 for ffected )

annel, and  !

T/r - Qc valu. T i v. sui um. , TA vi ToLiu 0.0 4 ivr w e afi= u md d.:nn d .

l

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

1 3/4 3-15 Amendment No. 73, 78 , 101 SUt+1ER - UNIT 1

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

1 3.3.2 The Engineered Safety Feature Actuation System (ESFAS) instrumentation l channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip I 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. ,

l 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 i more conservative than the value shown in the Allowable Value Column i of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint I value. l
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 channelinoperable 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 inoperabb kake the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS l

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 FE ATURES RESPONSE TIME of each ESFAS l 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 s,ecific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

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

3/4.3 TNSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demoDstrated 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. 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 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. j

/ m.%e.p~p sh nas4detek MS py .

i i

l l

i l

1 3/4 3-15a Aneendment No.13,101 SUPNER - UNIT 1

THIS PAGE TO BE DELETED DUE TO REPAGINATION l

1 l

l l

l i

l l

i SUMMER - UNIT 1 3/4 3-15a Amendment No.10,101,

- < ~

l . . . .

l u, .

-lm TABLE 3.3-4 I

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

g .

w  :

Q . Functional Unit I~ p -

, J h. _ . . 'T ". ;  ;' 1 Trip Setpoint Allowable Value ,

1. SAFETY INJECTION, REACTOR TRIP,  ! .

i FEEDWATER ISOLATION, CONTROL l ROOM ISOLATION, START DIESEL- l .

GENERATORS,' CONTAINMENT COOLING

' FANS AND ESSENTIAL SE,RVICE WATER. .

a. Manual Initiation b4 h4 NA - ,4A a
b. Automatic Actuation Logic N\ NA >% NA NA w c. . Reactor Building Pressure- 3.0 0.71 1, 5 $3.6 psig 13.86 psig t

1 High 1 gg ,

w ij

j d. Pressurizer Pressure--Low
3.1 10.71 1, 5 >1850 psig M839p'sig y - .
e. Differential Pr' essure ' .0 .87 1 5/ 197.psig $l06 psi Between Steam' fines--High 1 5 4 ti!
f. Steamline' Pressure--Low !0. 0 . 10. 1 5 >675 psig

_ k635psig(1)

2. REACTOR BUILDING SPRAY- ,
a. Manual Initiation NA NA ' NJ 'NA 'A ' ,

i

b. Automatic Actuation Logic NA NA A  ;

and Act'ation Relays-

~

u l;

'c . Reactor Building Pressure- i -- -a n 71 ,

'k 112.05 psig - 12.31 psig, High.3 (Phase 'A' i. solation .

aligns spray system dis- . - .

charge valves and NaOH tank- -

suction valves) .

(1) . Time constants utilized in lead lag controller for st6amline pressure-Tow are as follow)

~

T > 50 secs T $ 5 secs. - -

..t .

- . ._ ~.

TABLE 3.3-4 .

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

@ Functional Unit Trip Setpoint Allowable Value

!E

1. SAFETY INJECTION, REACTOR

@ TRIP, FEEDWATERISOLATION, 4 g CONTROL ROOM ISOLATION, START DIESEL GENERATORS, CONTAINMENT COOLING FANS AND ESSENTIAL SERVICE WATER.

a. ManualInitiation NA NA
b. Automatic Actuation Logic NA NA
c. Reactor Building Pressure-High 1 s3.6 psig s3.86 psig

$ d. PressurizerPressure--Low 21850 psig 21839 psig

[ e. Differential Pressure s97 psig s106 psi u Between Steamlines-High

f. Lemline Pressure-Low 2675 psig 2635 psig(1)
2. REACTOR BUILDING SPRAY
a. Manuallnitiation NA NA
b. Automatic Actuation Logic NA NA g and Actuation Relays f

E

c. Reactor Building Pressure-High 3(Phase 'A' isolation s12.05 psig s12.31 psig z 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:

Ti 2 50 secs. T2 s 5 secs.

E - -

TABLE 3.3-4 (Continued) 3 5 ENGINEERED SAFETY' FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS .

E Tm.a i '

y Functional Unit y L _ . a 'TT. ',  ;  ; Trip Setpoint Allowable Value

3. CONTAINMENT ISOLATION
a. , Phase "A" Isolation
1. Manual iA NA N. L NA 'NA
2. Safety Injection f ee 1 a ove for pil afety injection setpoints and allowable values
3. Automatic Actuation Logic FA i N \ NA -

NA i

and Actuation Relays  %

4

.b. Phase "B" Isolation

, 1. Automatic Actuation F A NA N\ NA NA

} Logic and Actuation

, Relays e

g 2. Reactor Building .0 .71 1. 5 $12.05 psig $12.31 psig Pressure-High 3 Purge and' Exhaust Isolation

c. . .
1. S5fety Injection .ee above for all s fety injection setpoints and allowable values Containment Radioactivity *
2. l 'A NA NL
  • High
3. ' Automatic Actuation . J A NA L

NA . Nk

. '3" -

Logic and Actuation - .

Re1ays .

g- . .'

g n -

2

- O D"U n.1,.1 1

  • Tripsetpointsshallbesettoensurethatthelimitsof)

~

m Specification 3,4+ rect are not exceeded.

9

.=.

I TABLE 3.3-4 (Continued) .

ENGINEERED SAFETY FE ATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS m

5 Functional Unit Tdp Setpoint Allowable Value g

se 1

3. CONTAINMENTISOLATION E.-. a. Phase "A" Isolation e

~

1. Manual NA NA
2. Safety Injection See I above for all safety injection setpoints See 1 above forall allowable values
3. Automatic Actuation Logic NA NA and Actuation Relays
b. Phase "B" Isolation w 1. Automatic Actuation Logic NA NA D and Actuation Relays
2. Reactor Building $12.05 psig 512.31 psig Pressure-Iligh 3
c. Purge and ExhaustIsolation
1. Safety Injection See 1 above for all safety injection setpoints See 1 above forall allowable values
2. Containment Radioactivity 3

Iligh a

{g

3. Automatic Actuation Logic NA NA and Actuation Relays

. S F

.I

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

TABLE 3.3-4 (Continued)

E I ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

' TM Functional Unit i12 _: r: (ra;  ;

Trip Setpoint Allowable Value

4. STEAM LINE ISOLATION

[

a. Manual f4 NA N\ NA NA
b. Automatic Actuation Logic N4 NA Ni NA NA and Actuation Relays ,
c. Reactor Building Pressure- 3 ,0 0.7 1 5 -<6.35 <6.61 High 2 -
d. Steam Flow in Two Steamlines- 2. ). 0 -

.16 1 5/ < a function < a function defined High, Coincident with , 1 5 defined as as follows: A Ap w follows: A AP corresponding to 44%

1 corresponding of full steam flow ,

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

4 steam flow load and then a Ap between 0% and increasing linearly 20% load and to a Ap corre-then a Ap sponding to 114.0%

increasing of full steam linearly to a flow at full load.

Ap correspond-ing to 110% of full steam flow at full load g T avg

- L w-Low .71 2552.0 F 1548.4 F l 1 e. Steamline Pressure - Low I ^^

12.7  :.:L 2675 psig 1635 psig(1)

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

e o

t i 1 50 secs. I2 < 5 secs.

TABLE 3.3-4 (Continued) -

,y 7 ENGINEERED SAFETY FE ATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS m

h Functional Unit Trio Setpoint Allowable Value R

x 8 4. STEAM LINE ISOLATION E

- a. Manual NA NA e

~

b. Automatic Actuation Logic NA NA and Actuation Relays
c. Reactor Building Pressure- < 6.35 < 6.61 IIigh 2
d. Steam Flow in Two Steamlines- Sa function defined as follows: A Ap Sa function defined as follows: A Ap

, IIigh, Concident with - corresponding to 40% of full steam flow corresponding to 44% of full steam flow

- < - - between 0% and 20% load and then a Ap between 0% and 20% load and then a Ap g

2 increasing linearly to a Ap corresponding to increasing linearly to a Ap corresponding to u 110% offullsteam flowat fullload 114.0% offull steam flow at fullload Tave - Low-Low > 552.0*F > 548.4*F

e. Steamline Pressure-Low 1675 psig 2635 psig(D i

8 re 5

t

,y (1) Time constan*g utilized in lead lag controller for steamline pressure low are as follows:

Il 2 50 secs. .* I2 s 5 secs.

t l

L TABLE 3.3-4 (Continuedl ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATIUN TRIP SETPOINTS 3 T.+e+

] Functional Unit '"- -^^ ' 7 c- Irip Setpoint

. Allowable Value S. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water Level - High-High Barton Transmitter  ; 0.8 1 4 1 7 579.2% of span $81.0% of span Rosemount Transmitter 1 0.8 .4 7 579.2% of span 1 181.0% of span
6. EMERGENCY FEEDWATER y a. 1tanual t A NA N; NA NA
b. Automatic Actuation Logic t A N NJ NA flA
c. Steam Generator Water Level - Low-Low Barton Transmitter 5.1 227.0% of span Rosemount Transmitter a s

' 326.1% of span 227.0% of span 325.7% of span

d. & f. Undervoltage-ESF Bus 25760 Volts with 25652 Volts with a a 50.25 second 50.275 second time time delay delay

>6576 volts with >6511 Volts with a g a 53.0 second 33.3secondtime time delay delay R

a n

TABLE 3.3-4 (Continued)  ;

ENGINEERED SAFETY FE ATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

3 Functional Unit Trio Setooint Allowable Value 8

5. TURBINE TRIP AND FEEDWATER

@ ISOLATION 3

- a. Steam Generator Water Level-IIigh-High Barton Transmitter 179.2% ofspan 181.0% of span Rosemount Transmitter $79.2% of span 181.0% ofspan

6. EMERGENCY FEEDWATER
a. Manual NA NA g b. Automatic Actuation Logic NA NA
c. Steam GeneratorWater 4 Level- Low-Low Barton Transmitter > 27.0% of span > 26.1% of span RosemountTransmitter > 27.0% of span > 25.7% of span ,

d.& f. Undervoltage-ESF Bus >5760 Volts with a <0.25 second > 5652 Volts with a <0.275 second time time delay delay

>6576 Volts with a <3.0 second >6511 Volts with a <3.3 second time delay time delay a

Et a

n

O w

ll m

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

= e Functional Unit , . l ' . _ _ c. a. 'Ta'

Trip Setpoint Allowable Value l

e. Safety Injection :e 1 above (all SI Se o' nts)
g. Trips of Main Feedwater 1A NA i lA NA NA Pumps
h. Suction transfer on Low j a l A 1442 ft. 4in.(2) 441 ft. 3 in.

Pressure

7. LOSS OF POWER w a. 7.2 kv Emergency Bus (A NA IIA 15760 volts with 15652 volts with a a Undervoltage (Loss of a 50.25 second 10.275 second time w Voltage) time delay delay
b. 7.2 kv Emergency Bus l lA 'IA 16576 volts 16511 volts with a Undervoltage with a 53.0 $3.3 second time second time delay delay B. AUTOMATIC SWITCHOVER TO CONTAINMENT SUMP
a. RWST Level Low-Low 4 NA N/ >18% 115%
b. Automatic Actuation Logic 2  ::7.

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

TABLE 3.3-4 (Continued) .

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

3 Functional Unit Trio Setpoint Allowable Value IE i e. Safety Injection See 1 above (all SI Setpoints) See 1 above (all SI Setpoints) h g. Trips of Main Feedwater Pumps NA NA H

~

h. Suction transfer on Low >442 ft. 4 in. G >441 fL 3 in.

Pressure

7. LOSS OF POWER
a. 7.2 kv Emergency Bus Undervoltage >5760 volts with a <0.25 second >5652 volts with a <0.275 second time (Loss of Voltage) time delay delay u b. 7.2 kv Emergency Bus Undervoltage >6576 volts with a <3.0 second > 6511 volts with a < 3.3 second time delay
  • time delay
8. AUTOMATIC SWITCIIOVERTO

$ CONTAINMENT SUMP

a. RWST Level Low-Low >18% >15%
b. Automatic Actuation Logic NA NA and Actuation Relays i

S

' it

, a

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

a p TABLE 3.3-4 (Continued)

I m

x ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS i

ttreat i E Functional Unit a'__,,,, ;;/'

$ Trip Setpoint Allowable Value

r. 9. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS INTERLOCKS
a. Pressurizer Pressure, P-ll 1.1 .71 1 .5 1985 psig >1974 psig &

$1996 psig

b. T, Low-Low, P-12 4.0 . .8 552 F >548.4 F & $555.6 F
c. Reactor Trip, P-4 km ' - ' ~~ N5 - ' 44 4 NA NA T

Di -

a a

i

%n O

1 L

TABLE 3.3-4 (Continued) .

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS m . -_

S g Functional Unit Trio Setpoint Allowable Value x

8

9. ENGINEERED SAFETY FEATURE E ACTUATION SYSTEM INTERLOCKS g _

- INTERLOCKS

a. Pressurizer Pressure, P-11 1985 psig > 1974 psig &

11996 psig

b. TavsLow-Low, P-12 552*F >548.4*F &

1555.6*F

c. Reactor Trip, P-4 NA NA U.

Y a

a-e i

a n

O

3/4.3 INSTRUMENTATION 8ASES 3/4.3.1 and 3/4.3.2 SYSTEM INSTRUMENTATIONREACTOR TRIP AND ENGINEERED SAFETY 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 valvas 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 an allowance has been made in the safety analysis to acccmmodate.this error. 7 ;,.th:.f. p ........ .~

% d;d f:r d;t mi.ir.; O.: ^ S /C U ii vr s channel when it etpoin ;

3 s founa w the Allowabie Value. The method is option i

tilizes the "as meas tion from 1 ied calibration point for l

ack and sensor components in c i he other uncertaint ith a statistical combination of a nd the unc e instrumentati 'ure the process varial le i + -

es in calibrating the instrumentat on. Lation 3.3< ,

u .., ;5,2TA, the winteractive W ";; . _ ' .2 effects

.. :f Nof the errors in the racc

& - ca = h ad 7 M%

SUMER - UNIT 1 8 3/4 3-1 Amendment No. 101

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3/4.3 INSTRUMENTATION l

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 1 capability is avaifable 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 frec uencies 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 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 measuredd 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 an 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 i m 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 l 1

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

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

" n-if!J in T C c 3. 3, , i n pe n.en u spou, ia ihm a wa i. ; a u n.4 sumacion ui b;rr$?!Pa ed in the analysis excluding those associated with the sneef Tn'd qack drift an e- racy of their measurement. TA or owance is the difference, in percen .

etween the tr' .oint and the value used i n the analysis for the actuation. . .

rror is the "as measured" deviation, in percent span, for fectYdN 1 from the specified trip setpoint. S or Sensor E s either the "as measu viation of the s ensor from its p ion point or the value specified in -

.4, in percent s Fom the analysis assumptions. Use of Equation 3.3-1 a y a se rift factor, an 7:= = d , J i :T; Tm. v., anu prov mes a 1.nre>Juii.c u,m  % oronor m g-E r m ,

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 allcwance. 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.

'IM? k iheEngineeredSafetyFeaturesActuationSystemsensesselectedplant parameters and determines whether or not predetermined limits are being A 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.

Amendment No. 35, 101 SUMMER - UNIT 1 B 3/4 3-la

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I INSERT %

The Engineered Safety Features response times specified in Tcble 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.

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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, 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 beated water is assumed not to occur until the VCT charging pump suction isolatie 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

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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 si gnals to those engineered safety features components whose aggregate function aest 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 aumps start and automatic valves position,2) reactor trip,3) feed-water isolation,4) startup of the emergency diesel generators,5) containment spray pump 3 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 anc ventilation systems start.

SUMMER - UNIT 1 B 3/4 3-la Amendment No. 05,101, i

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. Attachment II to Document Control Desk Letter TSP 930015, Supplement 1 Page 1 of 2 l

l PROPOSED TECHNICAL SPECIFICATION CII ANGE REQUEST TSP 930015, SUPPLEMENT 1, ADMINISTRATIVE CHANGES 1 VIRGIL C. SUMMER NUCLEAR STATION DESCRIPTION AND SAFETY EVALUATION DESCRIPTION OF SUPPLEMENT TO AMENDMENT REQUEST South Carolina Electric & Gas Company (SCE&G) proposes to modify the Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications (TS) to change the format ofTS Tables 2.2 1 and 3.3-4 from five columns to two columns. Existing 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 functional unit in a five column format. The five columns ofinformation are:

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

I The Trip Setpoints in TS Table 2.21 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 Features (ESF) Actuation System in l mitigating the consequences of accidents. )

The Trip Setpoints for the ESF Actuation System are presented in TS Table 3.3-4. The l setpoints, in accordance with the Allowable Value provided in TS Table 3.3 4, ensure that the consequences of Design Basis Accidents (DB As) will be acceptable. The basic assumption is  !

that the unit is being operated from within the Limiting Condition for Operation (LCO) at l the onset of the DBA and the equipment functions as designed. These TS tables provide the l setpoint information needed to determine the setpoint operability of the trip function.

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- Attachment II to Document Control Desk Letter TSP 930015, Supplement 1 Page 2 of 2

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l The five column format included provisions which in some cases eliminated the need for

! formal reporting through a Licensee Event Report (LER). The issuance of10 CFR 50.73 l l changed the filing requirements associated with an LER when an Allowable Value was

. exceeded. According to 10 CFR 50.73, an LER would not be required in response to the loss of a single channel. Only upon loss of a function would an LER be required. As a result, SCE&G desires to delete the TA, Z, and S columns. The remaining columns, the Trip Setpoint and Allowable Value, provide the necessary setpoint information to insure 1 continued safe operation of the plant. l In addition, several minor typographical errors and omissions from previously approved TS amendments ( AM 104 and 108) are requested. These items are described in detail in i Attachment l. )

l SAFETY EVALUATION The proposed TS amendment to Tables 2.2-1 and 3.3-4 is intended to delete three columns of information that is no longer used in either the daily operation of the plant or as an aid in determining event reportability. This information contains specific equipment and system tolerances and uncertainties and is used to establish setpoints and allowable values. The safety analysis limits are unchanged, all safety system setpoints are unchanged, and all Allowable Values remain unchanged. The proposed license amendment is administrative in I

nature since there is no impact on hardware, software, or setpoints. The two column format -

will continue to provide the setpoint information used to determine the setpoint operability of -

the trip function. I l

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  • AttachmentIII to Document Control Desk Letter TSP 930015, Supplement 1 Page 1 of 3 i

PROPOSED TECHNICAL SPECIFICATION CHANGE REQUEST TSP 930015, SUPPLEMENT 1, ADMINISTRATIVE CHANGES VIRGIL C. SUMMER NUCLEAR STATION NO SIGNIFICANT HAZARDS EVALUATION DESCRIPTION OF SUPPLEMENT TO AMENDMENT REQUEST South Carolina Electric & Gas Company (SCE&G) proposes to modify the Virgil C. Su;nmer Nuclear Station (VCSNS) Technical Specifications (TS) to change the format of TS Tables 2.2-1 and 3.3-4 from five columns to two columns. Existing 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 functional unit in a five column format. The five columns ofinformation are:

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

Z: Z, in percent span, is the statistical summation oferrors 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 instrument Value: drift assumed to occur between operational tests and the accurney 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 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 Features (ESP) Actuation System in mitigating the consequences of accidents.

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 acceptable. The basic assumption is that the unit is being operated from within the Limiting Condition for Operation (LCO) at the onset of the DBA and the equipment functions as designed. These TS tables provide the setpoint information needed to determine the setpoint operability of the trip function.

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o o Attachmant III to Document Control Desk Letter TSP 930015, Supplement 1 Page 2 of 3 The five column format included provisions which in some cases 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 an LER when an Allowable Value was j l

exceeded. According to 10 CFR 50.73, an LER would not be required in response to the loss of '

a single channel. Only upon loss of a function would an LER be required. As a result, SCE&G desires to delete the TA, Z, and S columns. The remaining columns, the Trip l Setpoint and Allowable Value, provide the necessary setpoint information to insure continued safe operation of the plant.

In addition, several minor typographical errors and/or omissions from previously approved TS amendments (AM 104 and 108) are requested. These items are described in detail in Attachment I.

! B ASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION 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.

This amendment request which describes administrative changes to TS Tables 2.2-1 and 3.3-4 has been reviewed and deemed not to involve significant hazards considerations. As discussed below, all applicable acceptance criteria are satisfied and the conclusions presented in the VCSNS FSAR remain valid. The basis for this determination follows:

l 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.

This change does not alter or delete any setpoints or Allowable Values and as such has no affect on any assumptions used for accident analysis. No hardware or software changes are involved, so no common mode or common cause failures can occur as a result of this change. This change has no impact on the daily operation of VCSNS.

The performance of periodic calibrations and channel checks will assure the setpoints remain within tolerance. Since this amendment request affects only information that is no longer used in the daily operation of the plant and has no impact on accident analysis, 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.

This change revises two TS tables which contain both eetpoints and Allowable Values as well as other information for safety trip functions. However, the revision only deletes three columns of data that were used in determining the operability of one channel of the safety function. These values are also used in determining the setpoints and are based on measured or published tolerances and uncertainties.

Although these columns are being deleted, no changes to any hardware, software, or setpoints will occur. Since these changes do not have any plant impact, no new failure mechanisms are introduced. Only the information not used on a daily basis is being removed from these tables; this will not create the possibility of a new or different kind of accident from any accident previously evaluated. l 1

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, Attachment III to Document Control Desk Letter l TSP 930015, Supplement 1 Page 3 of 3

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

This change revises the format of TS Tables 2.21 and 3.3-4 which list the setpoint and Allowable Values for safety trip functions. The data that is being removed from these tables was used to establish clear reportability requirements for any portion of one channel of any of the listed safety trip functions. Since the reporting requirements have changed and an LER is not required if one coincident channel is inoperable, this data is no longer used in daily operations. The margin of safety was established when setpoints and Allowable Values were determined, and no changes to these values are involved. There is no reduction in a margin of safety that could affect the plant, SCE&G employees, or the public.

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