ML20059G915

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Proposed Tech Specs Re Main Steam Line Radiation Monitor Sys High Radiation Trip Function for Initiating Automatic Reactor Scram & Automatic Closure of Main Steam Line Isolation Valves
ML20059G915
Person / Time
Site: Limerick  Constellation icon.png
Issue date: 10/29/1993
From:
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20059G907 List:
References
NUDOCS 9311090185
Download: ML20059G915 (39)
v • d • e


Text

c- -

d

-l ATTACHMENT 2 l

LIMERICK GENERATING STATION ,

UNITS 1 AND 2' .

i Docket Nos. 50-352 l' 50-353 License Nos. NPF-39 i NPF-85  :

TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 91-03-0  ;

LIST OF AFFECTED PAGES ,

Unit 1 Unit 2 .

2-4 2-4 4 8 2-8 8 2-8 ,

3/4 3-3 3/4 3-3 3/4 3-6 3/4 3-6 ,

3/4 3-7 3/4 3-7 s 3/4 3-11 3/4 3-11 3/4 3-18 3/4 3-18 ,

3/4 3-23 3/4 3-23 .j 3/4 3-26 3/4 3-26  ;

3/4 3-27  ;

3/4 3 3/4 3-31 3/A 3-31 i 3/4 6-19 3/4 6-19 .1 3/4 6-22 3/4 6-22 )

3/4 6-24 3/4 6-24 1 3/4 6-31 3/4 6-31 1 8 3/4 3-1 B 3/4 3-1 l B 3/4 3-2 B 3/4 3 -i '

B 3/4 3-3* B 3/4 3-3*

  • - Page revised to accomodate relocation of information from TS Bases Section 3/4.3.3.

)

9311090185 DR ADOCK931029

% l 05000352 ILi eDn y

1 1

l s

f UNIT 1 TS Pages l 1

1 l

i i

TABLE 2.2.1-1 .'

[: REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS M

Eo ALLOWABLE Q FUNCTIONAL UNIT TRIP SETPOINT VALUES

, 1. Intermediate Range Monitor, Neutron Flux-High s 120/125 divisions s 122/125 divisions of full scale of full scale jj 2. Average Power Range Monitor:

(( a. Neutron Flux-Upscale, Setdown s 15% of RATED THERMAL POWER s 20% of RATED THERMAL POWER

b. Neutron Flux-Upscale
1) During two recirculation loop operation:

a) Flow Biased s 0.58 W+ 59% with s 0.58 W+ 62% with a maximum of amaximumof b) High Flow Clamped s 116.5% of RATED $ 118.5% of RATED THERMAL POWER THERMAL POWER

2) During single recirculation loop operation:

a) Flow Biased s 0.58 W+ 54%, s 0.58 W+ 57%,

b) High Flow Clamped Not Required Not Required OPERABLE OPERABLE rp c. Inoperative N.A. N.A.

d. Downscalt 2 4% of RATED 23% of RATED THERMAL POWER THERMAL POWER
3. Reactor Vessel Steam Dome Pressure - High s 1037 psig s 1057 psig
4. Reactor Vessel Water Level - Low, level 3 2 12.5 inches above instrument 2 11.0 inches above zero* instrument zero
5. Main Steam Line Isolation Valve - Closure s 8% closed s 12% closed
6. DELETED DELETED DELETED b
7. Drywell Pressure - High s 1.68 psig s 1.88 psig
8. Scram Discharge Volume Water Level - High
a. Level Transmitter s 260' 9 5 8" elevation ** s 261' 5 5 8" elevation
b. Float Switch s 260' 9 5 8" elevation ** s 261' 5 5 8" elevation
9. Turbine Stop Valve - Closure s 5% close s 7% close
10. Turbine Control Valve - Closure, Trip Oil Pressure - Low 2 500 psig 2 465 psig
11. Reactor Mode Switch Shutdown Position N.A. N.A.
12. Manual Scram N,A. N.A.
  • See Bases Figure B 3/4.3-1.
    • Equivalent to 25.45 gallons / scram. discharge volume.

11MITING SAFETY SYSTEM SETTINGS

. . BASES  ;

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

4. Reactor Vessel Water Level-Low '

The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease. The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high er.ough above the fuel to assure that there is adequate protection for the fuel  !

and pressure limits.  !

5. Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit ,

the amount of fission product release for certain postulated events. The MSIVs are closed automatically from measured parameters such as high steam flow, low reactor .

l water level, high steam tunnel temperature, and low steam line pressure. The MSIVs l l closure scram anticipates the pressure and flux transients which could follow. MSIV  :

closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety 1 Limits. ,

6. DELETED

! l 7 Drywell Pressure-Hiah  ;

High pressure in the drywell could indicate a break in the primary pressure  ;

boundary systems or a loss of drywell cooling. The reactor is tripped in order  !

to minimize the possibility of fuel damage and reduce the amount of energy being l added to the coolant and to the primary containment. The trip setting was .

selected as low as possible without causing spurious trips. J l

I I

I i

i I

i LIHERICK - UNIT I B 2-8 l 1

. .=. , . , , .1

TABLE 3.3.1-1 (Continued) .'

C M REACTOR PROTECTION SYSTEM INSTRUMENTATION ,

E 9

APPLICABLE MINIMUM c- OPERATIONAL OPERABLE CHANNELS z  ; FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM fa) ACTION

. 6. DELETED' DELETED DELETED DELETED l

7. Drywell Pressure - High 1,2(h) 2 1 4
8. Scram Discharge Volume Water Level - High
a. Level Transmitter 1, 2 2 1

, 5(i) 2 3 s

b. Float Switch 1, 2 2 1 6

[ 5(i) 2 3

9. Turbine Stop Valve - Closure 1(j) 4(k) 6
10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 1(j) 2(k) 6
11. -Reactor Mode Switch Shutdown Position 1, 2 2 l' 3, 4 2 7 4

5 2 3

12. Manual Scram 1, 2. 2 1 3, 4 2 8 5 2 9 i____________________._...._._,.__.__..-..._.____..;,-..,; . _ . _ , _ , ...,.....__,.,..,__..._;_,.,-.._.~......._..-..._-._-._

6-TABLE 3.3.1-2

((*

1. Intermediate Range Monitors:

gj a. Neutron Flux - High N.A.

-d b. Inoperative N.A.

2. Average Power Range Monitor *:
a. Neutron Flux - Upscale, Setdown N.A.
b. Neutron Flux - Upscale
1) Flow Biased 50.09
2) High Flow Clamped 50.09
c. Inoperative N.A.

$$ d. Downscale N.A.

3. Reactor Vessel Steam Dome Pressure - High 50.55 gl
4. Reactor Vessel Water Level - Low, level 3 s1.05
5. Main Steam Line Isolation Valve - Closure 50.06
6. DELETED DELETED 'l
7. Drywell Pressure - High N . A .'
8. Scram Discharge Volume Water Level - High
a. Level Transmitter N.A.
b. Float Switch N.A.
9. . Turbine Stop Valve - Closure 50.06
10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 50.08**

.11. Reactor Mode Switch Shutdown Position N.A.

12. Manual Scram N.A.
  • Neutron detectors.are. exempt from response time testing. Response time shall.be measured -

from the detector output.or from the input of the fi"st electronic component in the channel.

** Measured from start of turbine' control valve fast closure.

. _ - _ . .-_ ___ -______-- ____-_ _ _-_ -_____ ____-_ - ---_ - ____. = .. ..- . =.-.

TABLE 4.3.1.1-1 .'

C

, M REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS B

Q CHANNEL OPERATIONAL

, CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH c FUNCTIONAL UNIT CHECK TEST CALIBRATION (a) SURVEILLANCE RE0VIRED 5

+ 1. Intermediate Range Monitors:

~ a. Neutron Flux - High S/U,S(b) S/U(c),W R 2 S W(j) R 3,4,5

b. Inoperative N.A. W(j) N.A. 2,3,4,5
2. Average Power Range Monitor (f):
a. Neutron Flux - S/U,S(b) S/U(c),W SA 2 Upscale, - Setdown S W(j) SA 3,5(k)
b. Neutron Flux - Upscale g 1) Flow Biased S,0(g) S/U(c),Q W(d)(e),SA 1
2) High Flow Clamped S S/U(c),Q W(d)(e), SA 1
c. Inoperative N.A. Q(j) N.A. 1, 2, 3, 5(k)
d. Downscale S Q SA 1
3. Reactor Vessel Steam Dome Pressure - High. S Q R 1, 2(h)
4. Reactor Vessel Water Level-Low, level 3 S Q R 1, 2
5. Main Steam Line Isolation Valve - Closure N.A. Q R 1
6. DELETED- DELETED DELETED DELETED DELETED I
7. Drywell Pressure - High. S Q R 1, 2
8. Scram Discharge Volume-Water Level - High-
a. Level Transmitter- S 0 R 1,2,5(i)
b. Flo'at Switch N.A. Q- .R~ 1, 2, 5(i)

T ^

TABLE 3.3.2-1 .

E:

Eg ISOLATION ACTUATION INSTRUMENTATION 5

.p MINIMUM APPLICABLE

, ISOLATION OPERABLE CHANNELS OPERATIONAL ,

c: TRIP FUNCTION SIGNAL (al PER TRIP SYSTEM (b) CONDITION ACTION 5

-d 1. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level
1) Low, Low-Level 2 B 2 1,2,3 21
2) Low, Low, Low-level 1 C 2 1,2,3 21
b. DELETED DELETED DELETED' DELETED DELETED' l
c. Main Steam Line Pressure - Low P 2 1 22  ;

$$ d. Main Steam Line

,, Flow - High E 2/line 1, 2, 3 20 i

II e. Condenser Vacuum - Low Q 2 1, 2**, 3** 21

f. Outboard MSIV Room Temperature - High F(f) 2 1,2,3 21
g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High F(f) 14 1, 2, 3 21
h. Manual Initiation NA. 2 1,2,3 24
2. _RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
a. Reactor Vessel Water Level low --Level 3 A 2 1, 2, 3 23
b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High ~V 2 1, 2, 3' 23 j l c. Manual' Initiation _NA 1 1,2,3 24 7 a

_.._..__._.._______._________.._.__________.____.__.___________________..__m._____~;__,,...w._.__J-., .m.._..,_,_,,,u...,__,m,_, ._,,,,.,_...,.,,,,._.,_,..,,.,.,,,m.,,,, . . _ , , . . . , , , , , . , , , ,

TABLE 3.3.2-2 -

C M ISOLATION ACTUATION INSTRUMENTATION SETPOINTS S

Q ALLOWABLE

, TRIP FUNCTION TRIP SETPOINT VALUE C

g 1. MAIN STEAM LINE ISOLATION

~

a. Reactor Vessel Water Level
1) Low, Low - Level 2 2 - 38 inches
  • 2 - 45 inches
2) Low, Low, Low - Level 1 2 - 129 inches
  • 2 - 136 inches
b. DELETED DELETED DELETED . l'
c. Main Steam Line Pressure - Low 2 756 psig 2 736 psig w d. Main Steam Line 2 Flow - High s 108.7 psid s 111.7 psid
e. Condenser Vacuum - Low 10.5 psia 210.1 psia /s 10.9 psia.
f. Outboard MSIV Room Temperature - High s 192*F s 200*F
g. Turbine Enclosure - Main Steam .

Line Tunnel Temperature - High s 165 F $ 175'F

h. Manual Initiation N.A. N.A.
2. RHR SYSTEM SHUTOOWN COOLING MODE ISOLATION
a. Reactor Vessel Water level Low - Level 3 2 12.5 inches
  • 2 11.0 inches
b. Reactor Vessel (RHR Cut-in 1

Permissive.) Pressure - High s 75 psig 5 95 psig

c. Manual Initiation N.A. N.A.

~

h

-- TABLE 3.3.2-3 '

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME  !

TRIP FUNCTION RESPONSE TIME (Seconds)# j

1. MAIN STEAM LINE-1 SOLATION
a. Reactor Vessel Water Level l
1) Low, low - Level 2 5 13(a)** '
2) Low, Low, Lov - Level 1 s 1.0*/s 13(a)**
b. DELETED DELETED
c. Main Steam Line Pressure - Low s_l.0*/s 13(a)**.

r

d. Main Steam Line '

Flow - High 5 0.5*/s 13(a)**

I

e. Condenser Vacuum - Low H.A.
f. Outboard MSIV Room Temperature - High N.A. j
g. Turbine Enclosure - Main Steam 1 Line Tunnel Temperature - High N.A.  ;
h. Manual Initiation N.A. .i
2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
a. Reactor Vessel Water Level l Low - Level 3 . ,3(a)
b. Reactor Vessel.(RHR Cut-In Permissive) Pressure - High N.A.  ;
c. Manual Initiation N.A.
3. REACTOR WATER CLEANUP SYSTEM ISOLATION
a. RWCS A Flow - High s 13##
b. RWCS Area Temperature - High -N.A.
c. RWCS Area Ventilation a Temperature - High N.A.
d. SLCS Initiation N.A.
e. Reactor Vessel Water Level -

Low, Low - Level 2 5 13(a)

f. Manual Initiation N.A.

LIMERICK - UNIT I 3/4 3-23 l _

)

)

TABLE 3.3.2-3 (Continued) l

. ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME j TRIP FUNCTION RESPONSE TIME-(Secondsl#-

f. Outside Atmosphere To Refueling Area a Pressure - Low N.A.
g. Reactor Enclosure Manual  :

Initiation N.A. ,

h. Refueling Area Manual Initiation N.A.  ;

I TABLE NOTATIONS (a) Isolation system instrumentation response. time specified includes 10 seconds .;

diesel generator starting and 3 seconds for sequence loading delays.  ;

(b) DELETED l' i

  • Isolation system instrumentation response time for MSIV only. No diesel -l generator delays assumed for MSIVs. *
    • Isolation system instrumentation response time for associated valves i except MSIVs.
  1. Isolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time shown in Tables 3.6.3-1, 3.6.5.2.1-1 and 3.6.5.2.2-1 for valves in each c valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.  !
    1. With 45 second time delay.

l i

\

i i

I LIMERICK - UNIT 1 3/4 3-26 I

\

y TABLE 4.3.2.1-1 .'

C

?! ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE' REQUIREMENTS 5

CHANNEL OPERATICNAL ,

, CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH c: TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLM$CE REQUIRID 5

-d

1. MAIN STEAM LINE ISOLATION
a. Reactor Vessel Water Level ##
1) Low, Low, level 2 S Q R 1, 2, 3
2) Low, Low, low - Level 1 S Q R 1, 2, 3
b. DELETED DELETED DELETED DELETED DELETED I
c. Main Steam Line Pressure - Low S M R 1

$$ d. Main Steam Line w Flow - High S M R 1,2,3

$ e. Condenser Vacuum - Low S M R 1, 2**, 3**

f. Outboard MSIV Room Temperature - High S M R 1, 2, 3
g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High S M R 1, 2, 3 -
h. Manual Initiation N.A. R N.A. 1, 2, 3
2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
a. Reactor Vessel Water Leve1###-

Low - Level 3 5 Q R ' 1, 2, 3

b. Reactor Vessel (RHR Cut-In 'S Q R 1, 2, 3-Permissive) Pressure ## - High
c. Manual-Initiation N.A. R N.A. 1,.2, 3

_ _ _ - - __-____-________:______-__.-____-._--._.--...=._...-....-- . . . - -

- [: TABLE 4.3.2.1-1-(Continued)

M ISOLATION ACTUATION lNSlRUMENTATION SURVEILLANCE REQUIREMENTS E3 CHANNEL 0PERATIONAL CHANNEL FUNCTIONAL . CHANNEL . CONDITIONS FOR WHICH Q TRIP FUNCTION . CHECK TEST CALIBRATION SURVEILLANCE REQUIREL E 7. SECONDARY CONTAINMENT ISOLATION

q a. Reactor Vessel Water Level ##

Low, Low - Level 2 S Q R 1 , 2 ,l 3

b. Drywell Pressure ### - High S Q R 1, 2, 3
c. 1. Refueling-Area Unit 1 Ventilation Exhaust Duct Radiation - High S M R *#
2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S M R *#
d. Reactor Enclosure Ventilation 23 Exhaust Duct Radiation - High S M R 1, 2, 3 u, e. Outside Atmosphere To. Reactor d, Enclosure a Pressure - Low N.A. M Q 1, 2, 3
f. Outside Atmosphere To Refueling Area a Pressure - Low N.A. M Q -
g. Reactor Enclosure Manual Initiation' N.A. R N.A. 1, 2, 3
h. Refueling Area
  • Manual Initiation N.A. R N.A.

handling irradiated fuel in the refueling area secondary containment, or (2 during CORE

  • Required when (1)3) during operations with a potential for draining -the reactor vessel with)the vessel.

ALTERATIONS, head-removed and or:(fuel in the vessel. ,

    • When not administratively bypassed and/or when any turbine stop valve is open.

L#During operation of the associated Unit- 1 or Unit 2 ventilation exhaust system.

'##These trip functions (la, 2b, 3e, 6a, 6h, and 7a) are common to the ECCS actuation trip function. '

      1. These trip functions (2a, 6b, and 7b) are common to-the RPS and ECCS actuation trip functions.

l

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

jf TABLE 3.6.3-1

!! PART A - PRIMARY CONTAINMENT ISOLATION VALVES 9

Ei INB0ARD OUTBOARD ISOL.

PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S)-, NOTES P&lD Ca

s. NUMBER BARRIER BARRIER TIME.IF APP. IF APP.

(SEC)(26) (20) 0038 CONTAINMENT INSTRUMENT 59-1005B (CK) NA 59 GAS SUPPLY - HEADER 'B' HV59-129B 7 C,H,S 003D-2 CONTAINMENT INSTRUMENT 59-1112(CK) NA GAS SUPPLY TO ADS VALVES HV59-1518 45 M 59 E&K

007A(B,C,D) MAIN STEAM LINE HV41-lF022A 5* C,E,F,P,Q 6 '41 l 4" 'A'(B,C,D) (B,C,D) 4

= HV41-lF028A 5* C,E,F,P,Q 6

.l 4

)$ 1 40' 001B 45 EA 6

  • F,K.P)

XV40-101B NA 6,1

.EdPATB, THIS TABLE) 008 MAIN STEAM LINE DRAIN HV41-IF016 30 C,E,F,P,Q 4. 41 HV41-lF019 30 C,E,F,P,Q NA 41 009A- FEEDWATER

.41-lF010A(CK) HV41-1F074A(CK) NA.

41-1036A(CK) NA

'HV41-130B 45 HV41-133A 45-HV41-109A NA 32 HV41-lF032A NA LHV55-lF105 (CK). 30 7' HV44-lF039(CK) NA (X-98) 41-1016(X-98, NA 31 X-44)

1 TABLE 3.6.3-1 (Continued)

PART A - PRIMARY CONTAINMENT ISOLATION VALVES e

E INBOARD OUTBOARD ISOL.

' PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID 4 c: NUMBER BARRIER BARRIER TIME.IF APP. IF APP.

g (SEC)(26) (20)

~

025 DRYWELL PURGE SUPPLY HV57-121(X-201A) 5** B,H,S,U,W,R,T 3,11,14 57-HV57-123 5** B,H,S,U,W,R,T 3,11,14 HV57-109 6** B,H,S,U,W,R,T 11 (X-201A)

HV57-131 5** B,H,S,U,W,R,T 11 (X-201A)

HV57-135 6** B,H,S,U,W,R,T 11 HYDROGEN RECOMBINER "B" HV57-163 9 B,H,R,5 3,11,14 INLET w FV-C-00-101B 90 B,H,R,5 11 026 DRYWELL PURGE EXHAUST HV57-Il4 5** B,H,5,U,W,R,T 3,11,14,33 57 i' HV57-111 15** B,H,S,U,R,T 11-Z SV57-139 5 10 HV57-Il5 6** B,H,S,U,W,R,T 11,33 HV57-ll7 5** B,H,S,U,R,T 11 SV57-145 5 B,H,R,S 11 HYDROGEN RECOMBINER "A" HV57-161 9 B,H,Ps,S 3,11,14 INLET FV-C-00-101A 90 B,H,R,S 11 027A CONTAINMENT INSTRUMENT 59-1128(CK) NA 59 GAS SUPPLY TO ADS VALVES HV59-151A 45 M H,M,&S 028A-1 RECIRC LOOP SAMPLE HV43-IF019 10 8 43 HV43-lF020 10 B 028A-2 DRYWELL H2/02 SAMPLE SV57-132 5 B,H,R,5 11 57 SV57-142 5 B,H,R,5 11 028A-3 DRYWELL H2/02 SAMPLE SV57-134 5 B,H,R,S 11 57 t SV57-144- 5 B,H,R,S 11 E____----_-----______.-____._____ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . .

TABLE 3.6.3-1 (Continued) .

g PART A - PRIMARY CONTAINMENT ISOLATION VALVES

.g b INBOARD OUTBOARD ISOL.

PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID s; NUMBER BARRIER BARRIER TINE.IF APP. IF APP.

g (SEC)(26) (20)

~

040G-1 ILRT DATA ACQUISITION 60-1057 NA 11- 60 60-1058 NA 11 040G-2 '60-1071

ILRT DATA ACQUISITION NA 11 60 60-1070 NA 11 ,

040H-1 CONTAINMENT INSTRUMENT 59-1005A(CK) NA 59.

GAS SUPPLY - HEADER 'A' HV59-129A 7 C,H,5 042 STANDBY LIQUID CONTROL 48-lF007(CK) NA 48 (X-116) HV48-lF006A 60 29

^

043B MAIN STEAM SAMPLE HV41-lF084 10 8 41

? HV41-1F085 10 B 044 RWCU ALTERNATE 41-1017 NA 5,31 41 RETURN 41-1016(X-9A, NA X-98)

PSV41- ll2 NA 045A(B,C,D) LPCI INJECTION ' A'(B,C,0) HV51-lF041A(B,C,NA 9,22 51

- 42A(B,C, 7 9,22 D) .

HV51-1F017A 38 (B,C,D) 050A-1 DRYWELL PRESSURE HV42-147B 45 10 42' INSTRUMENTATION 053 DRYWELL CHILLED WATER HV87-128 .

60 C,H- 11 87 SUPPLY - LOOP 'A' ~ HV87-120A . 60 C,H 11 HV87-125A 60 C,H 11

-uww- m w w1m ts D'wmvm m.--wai--4-i m #'a-. _ms-- s- y-w mar a mw-+2, sea _mu-m+w2-m-har w. i mn-mwm-.m uu_-am.imaa24a rm_----A a--a _m.a-amu -

s Ih TABLE 3.6.3-1 (Continued) lb p; PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES n

- INBOARD OUTBOARD ISOL.  !

ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID' 25 PENETRATION FUNCTION

-4 NUMBER BARRIER BARRIER TIME.IF APP. IF APP.

-- (SEC)(26) (20)

- 41 003A-1 INSTRUMENTATION 'D' -- XV41-1F0700 1 MAIN STEAM LINE FLOW XV41-lF0730

'A' XV43-1F003A 1 43-003A-2 INSTRUMENTATION --

RECIRC PUMP SEAL PRESSURE .

XV55-1F024A 1- 55 003C-1 INSTR. - HPCI STEAM FLOW -- ,

XV55-IF024C 1 55 003C-2 INSTR. - HPCI STEAM FLOW --

S XV41-lF070A - 41 003D-1 INSTR. 'A' MAIN STEAM -- 1 LINE FLOW XV41-IF073A C,E,F,P,Q L 6' 007A(B,C.D) INSTR. 'A'(B,C,D) MAIN (HV41-lF022A(B, 5* ~41-STEAM LINE PRESSURE SEE PART A HV41-1F028A 5* - C,E,F,P,Q 6 C,D) THIS TABLE). 8,C,D) SEE , ART A . THIS TABLE) HV40-lF0018. 45 EA 6-F,K,P) SEE , ART A THIS TABLE) XV40-10 1B(F, 1,6 K,P) 

                                                                                                                 ~

XV42-1F0458 l' 42 020A-1 INSTR - RPV LEVEL -- XV51-102B .1 51 020A-2 INSTR 'B' LPCI DELTA P -- INSTR - 'D' LPCI DELTA'P- -- XV51-103B 1 51' 020A-3 - XV42-lF045C 1 ~-42 0208-1 INSTR - RPV LEVEL .-- XV51-102C' 1 51 0208-2' INSTR 'CLPCI. DELTA P -- ______.m._..._.__-_.,_.-.m___.__u..:_.m__.-.._.m___m_____._.m_.___._________.____--._m._.____-_,______-.m.-.ms .!sm'-m___.u...-_-_m_mu_:= e- uw w-sam, w- m _ _ _ . _-_-w.a sm _ea, +--e=*m_ news--.w _.:-e._mai,---- emewamcuwo vow, 

   . 3/4.3 INSTRUMENTATION                                                                    l BASES 3J4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram.to:

1

a. Preserve the integrity of the fuel cladding.
b. Preserve the integrity of the reactor coolant system. j
c. Minimize the energy which must be adsorbed following a j loss-of-coolant accident, and
d. Prevent inadvertent criticality.

This specification provides the limiting conditions for operation ' necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service ' because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance. { The reactor protection system is made up of two independent trip systems.  ! There are usually four channels to monitor each parameter with two channels in a each trip system. The outputs of the channels in a trip system are combined i in a logic so that either channel will trip that trip system. The tripping of-both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified > surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Specification , Improvement Analyses for BWR Reactor Protection System, as a proved by the 1 NRC and documented in the NRC Safety Evaluation Report (SER) letter to T. A. Pickens from A. Thadani dated July 15, 1987. The bases for t e trip settings of RPS are discussed in the bases for Specification 2.2.1. l Automatic reactor trip upon receipt of a high-high radiation signal  ! from the Main Steam Line Radiation Monitoring, System was removed as the result  : of an analysis performed by General Electric in NED0-31400A. The NRC approved ~ the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15,1991). a The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses. No credit was taken for those channels with response times indicated as not a ) Response time may be demonstrated by any series of sequential,overlapping pplicable. or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times. I LIMERICK - UNIT 1 B 3/4 3-1 i

l INSTRUMENTATION , r BASES . 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION i This specification ensures the effectiveness of the instrumentation used.to-  ! mitigate the consequences of accidents by prescribing the OPERABILITY trip  ! setpoints and response. times for isolation of the reactor systems. When  ! , necessary, one channel may be inoperable for brief intervals to' conduct required  : surveillance. i Specified surveillance intervals and maintenance outage times.have been determined in accordance with NEDC-30851P, Supplement 2, " Technical Specification Improvement Analysis for BWR Instrumentation Common to RPS and i ECCS Instrumentation" as approved by the NRC and documented in the NRC Safety  ; Evaluation Report (SER) (letter to D. N. Grace from C. E. Rossi dated January 6, ' 1989). Automatic closure of the MSIVs upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the j result of an analysis performed by General Electric in NED0-31400A. The NRC 1 a approved the results of this analysis as documented in the SER (letter to i George J. Beck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15, 1991). l Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on  ! safety. The setpoints of other instrumentation, where only the high or low-end of .the setting have a direct bearing on safety, are established at a level away i from the normal operating range to prevent inadvertent actuation of the systems involved. , Except for the MSIVs, the safety analysis does not address individual sensor I response times or the response times of the logic systems to which the sensors 1 are connected. For D.C. operated valves, a 3 second delay is assumed before the  : valve starts to move. For A.C. operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency diesel ,

generators. In this event, a time of 13 seconds is assumed before the valve '

y' starts to move. in addition to the pips break, the failure of the D.C. operated  ! is concurrent with the i valve 10-second is assumed; diesel startup thus and the signal delay (sensor the 3 second response) load center loa ding delay.- The safety  : analysis considers an allowable inventory loss in each case which in turn  ! determines the valve speed in conjunction with the 13-second delay. It follows that checking the valve speeds and the 13-second time for emergency power  ; establishment will establish the response time for the isolation functions. i Operation with a trip set less conservative than its Trip Setpoint but i within its specified Allowable Value is acceptable on the basis that the ' difference between each Trip Setpoint and the Allowable Value is an allowance  ! for instrument drift specifically allocated for each trip in the safety t , analyses. ' 3/4.3.3 EMERGENCY CORE C00llNG SYSTEM ACTUATION INSTRUMENTATION 9 4 The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness . of the systems to provide the design protection. Although the instruments are " listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time. LIMERICK - UNIT 1 B 3/4 3-2 , i l 

  .       INSTRUMENTATION                                                                                             I
      +

BASES Specified surveillance intervals and maintenance outage times have been  ! determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical ' Specification Improvement Methodology Actuation Instrumentation)," as approve (with Demonstration for BWR ECCSd by the NRC.an (letter to D. N. Grace from A. C. Thadani dated December 9,1988'(Part 1) and - letter to D. N. Grace from C. E. Rossi dated December 9, 1988 (Part 2)) i Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the ' difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety , analyses. 3/4.3.4 RECIRCUL ATION PUMP TRIP ACTUATION INSTRUMENTATION l The anticipated transient without scram l system provides a means of limiting the consequ(ATWS) recirculation pump tripences of the  ; of a failure to scram during an anticipated transient. The response of the i plant to this postulated event falls within the envelope of study events in ( General Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR. i The end-of-cycle recirculation pump trip (EOC-RPT) system is a supplement to l the reactor trip. During turbine trip and generator load rejection events, the ~i E0C-RPT will reduce the likelihood of reactor vessel level decreasing to level  ;

2. Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in i order to reduce the void collapse in the core during two of the most limiting '

pressurization events. The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the i turbine control valves.  ; A fast closure sensor from each of two turbine control valves provides input i to the E0C-RPT system; a fast closure sensor from each of the other two turbine i control valves provides inout to the second EOC-RPT system. Similarly, a - position switch for each of two turbine stop valves provides input to one E0C-RPT system; a position switch from each of the other two stop valves provides i input to the other E0C-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The  : operation of either logic will actuate the EOC-RPT system and trip both 1 J recirculation pumps. I Each E0C-RPT system may be manually bypassed by use of a keyswitch which is  ; administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room. The E0C-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e., 175 ms. Included in this time are: the response time of the sensor, the time allotted for breaker arc suppression, and the response time of the system logic. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. ~ LIMERICK - UNIT 1 B 3/4 3-3 4 i em , - - - - _ _ , , - . - - . - , - - - 

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

9 i 8 I f i 4 4 UNIT 2 a l TS Pages ' i l l i I I I i l 4 i I i l l i 1

TABLE 2.2.1-1 Es REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 

                     #4                                                                                                                   ALLOWABLE FUNCTIONAL UNIT                                               TRIP SETPOINT                       VALUES' i                    1. Intermediate Range Monitor, Neutron Flux-High            s 120/125 divisions               s 122/125 divisions c;                                                                                    of full scale                      of full scale Ej                    2. Average Power Range Monitor:
a. Neutron Flux-Upscale, Seldown s 15% of RATED THERMAL POWER s 20% of RATED THERMAL POWER
b. Neutron Flux-Upscale
1) During two recirculation loop operation:

a) Flow Biased s 0.58 W+ 59%, with s 0.58 W+ 62%, with a maximum of a maximum of b) High Flow Clamped s 116.5% of RATED s 118.5% of RATED THERMAL POWER THERMAL POWER

2) During single recirculation loop operation:

a) Flow Biased s 0.58 W+ 54%, s 0.58 W+ 57%, no b) High Flow Clamped Not Required Not Required jo OPERABLE OPERABLE

c. Inoperative N.A. N.A.
d. Downstale 2 4% of RATED 2 3% of RATED-THERMAL POWER THERMAL POWER
3. Reactor Vessel Steam Dome Pressure - High s 1037 psig s 1057 psig
4. Reactor Vessel Water Level - Low, Level 3 2 12.5 inches above instrument 2_11.0 inches above zero* instrument zero
5. Main Steam Line Isolation Valve - Closure s 8% closed s 12% closed
6. DELETE 0 DELETED' DELETED 1;
7. Drywell- Pressure - High 5 1.68 psig s 1.88 psig
8. Scram Discharge Volume Water Level - High l
a. Level Transmitter s 261' 1 1/4" elevation ** s 261' 9 1/4" elevation i b. Float Switch s 261' 1 1/4" elevation ** _s 261'.9 1/4" elevation l

l

  • See Bases Figure B 3/4.3-1.
                                          ** Equivalent to 25.58 gallons / scram discharge volume.

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

4. Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease. The scram setting-was chosen f ar enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.
5. Main Steam line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events. .The MSIVs are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature, and low steam line pressure.

The MSIVs closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits. 6 DELETED l 7 DrYwell Pressure-High High pressure in the drywell could indicate a break in the primary pressure boundary systems or a loss of drywell cooling. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and to the primary containment. The trip setting was selected as low as possible without causing spurious trips. LIMERICK - UNIT 2 E 2-8 __A

4 

                  .C                                                                                                                                    TABLE 3.3.1-1 (Continued)
  • x 5- REACTOR PROTECTION SYSTEM INSTRUMENTATION R .

c: APPLICABLE MINIMUM 5 OPERATIONAL OPERABLE CHANNELS I CONDITIONS PER TRIP SYSTEM (a) ACTION FUNCTIONAL UNIT [ DELETED DELETED DELETED

6. DELETED ,
7. Drywell 1 Pressure - High 1,2(h) 2 1
8. Scram Discharge Volume Water 'r Level - High
a. Level Transmitter 1, 2 2 1 w

2 5(i) 2 3 Y w b. Float Switch 1, 2 2 1 5(i) 2 3

9. Turbine Stop Valve - Closure 1(j) 4(k) 6
10. Turbine Control Valve Fast Closure, ~

2(k) 6 Trip Oil Pressure - Low 1(jj

11. Reactor Mode Switch Shutdown t Position 1, 2 2 1
                                                                                                                                                                                 ' 3, 4                2                              7                             1 5             2                              3
12. Manual Scram .1, 2 2 1 3, . 4 2 8-5 2 9

_ . _ _ _ . . . _ . _ _ _ . _ _ . . _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ __._._______.____m._ . . . ~~...___.,__.....,__._.._.-__......_._.,__-_....m. _ .-~.-~...~---_....--...._._4 . . , . _ . .

[: TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES 7; RESPONSE TIME 7c FUNCTIONAL UNIT (Seconds) t i i c 1. Intermediate Range Monitors: 

     =

Z a. Neutron Flux - High N.A.

b. Inoperative N.A.
2. Average Power Range Monitor *:
a. Neutron Flux - Upscale, Setdown N.A.
b. Neutron Flux - Vascale 1 Flow Biased 50.09 2)) High Flow Clamped 50.09 ,
c. Inoperative N.A. ,

ls d. Downscale N.A. r

3. Reactor Vessel Steam Dome Pressure - High 50.55'

[

4. Reactor Vessel Water Level - Low, level 3 $1.05
5. Main Steam Line Isolation Valve - Closure. 50.06
6. DELETED DELETED l
7. Drywell Pressure - High N.A..
8. Scram Discharge Volume Water Level - High
a. Level Transmitter N.A.
b. Float Switch N.A. ,
9. Turbine Stop' Valve - Closure -50.06
10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low ,

so.08**

11. Reactor Mode. Switch Shutdown Position N.A.
12. Manual Scram N.A.  ;
  • Neutron detectors are exempt from response time testing. Response time shall be measured from the detector output or from the input of the first electronic componentLin the channel.
                   ** Measured .from start of turbine control valve fast closure.

4

j

_ _ _ _ _ _ _ _ _ _ . _ - . _ . _ _ _ _ . _ . - _ _ . _ . . .. _. ..__u._.__.______,,.__. _

TABLE 4.3.-l.1-1 r. REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS R CHANNEL OPERATIONAL 

       ,                                                                                          CHANNEL      FUNCTIONAL                   CHANNEL           CONDITIONS FOR WHICH c-                             FUNCTIONAL UNIT                                                 CHECK          TEST                 CALIBRATION (a)       SURVEILLANCE REOUIRED 5

H 1 Intermediate Range Monitors: N a. Neutron Flux - High S/U,S(b) S/U(c),W R 2 S W(j) R 3, 4. 5

b. Inoperative N.A. W(j) N.A. 2,3,4,5
2. Average Power Range Monitor (f):
a. Neutron Flux - S/U,S(b) S/U(c),W SA 2 Upscale, Setdown S W(j) SA 3,5(k)
b. Neutron Flux - Upscale w 1) Flow Biased 5,0(g) S/U(c),Q W(d)(e),SA 1 u

w 2) High Flow Clamped S S/U(c),Q W(d)(e),SA 1

c. Inoperative N.A. Q(j) N.A. 1, 2, 3, 5(k)
d. Downscale S Q SA 1
3. Reactor Vessel Steam Dome Pressure - High S Q R 1, 2(h)  !
4. Reactor Vessel Water Level-Low, level 3 S Q R 1, 2
5. Main Steam Line Isolation Valve - Closure N.A. Q R 1
6. ~ DELETED DELETED DELETED DELETED- DELETED l
7. Drywell Pressure - High S Q R- 1, 2
8. Scram Discharge Volume Water level - High .

a.- Level Transmitter S R 1, 2, 5(i

b. . Float Switch N.A.

Q Q R I,2,5(i))- 

                                                                                                                                                                                                                                                                                             ^

l C TABLE 3.3.2-1 ' 

                            .R ISOLATION ACTUATION INSTRUMENTATION c                                                                                                                             MINIMUM                                         APPLICABLE 5                                                                                                      ISOLATION      OPERABLE CHANNELS                                       OPERATIONAL

[ TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION

1. MAIN STEAM LINE ISOLATION
a. Reactor Vessel Water Level
1) Low, Low-level 2 B 2 1,2,3 21
2) Low, Low, Low-Level 1 C 2 1, 2, 3 21
b. DELETED DELETED DELETED DELETED DELETED l.

b c. Main Steam Line 1 Pressure - Low P 2 1 22 w 

                             ~
d. Main Steam 1.ine Flow - liigh E 2/line 1, 2, 3 20
e. Condenser Vacuum - Low Q 2 1, 2**, 3** 21
f. Outboard MSIV Room Temperature - liigh F(f) 2 1,2,3 21
g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High F(f) 14 1, 2, 3 : 21
h. Manual Initiation .NA 2 1,2,3 24
2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
a. Reactor Vessel Water Level Low - Level 3- A- 2 '1, 2, 3 23
b. Reactor Vessel.(RHR Cut-In
                                                                                       . Permissive) Pressure - High                    V                      :2                                        .l',  2, 3                                   23
c. Manual Initiation NA' 1 .1,/2, 3 24 .;

_ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . . -, _ _ .._ _ . _ ______._.__m._____ . _ . _ _ _ _ . . .. ._.___.-___..._:._._...,________...

r TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E ALLOWABLE i TRIP FUNCTION TRIP SETPOINT VALUE C [] 1. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level
1) Low, Low - Level 2 2 - 38 inches
  • 2 - 45 inches
2) Low, Low, Low - Level 1 2 - 129 inches
  • 2 - 136 inches
b. DELETED DELETED DELETED
c. Main Steam Line Pressure - Low 2 756 psig 2 736 psig de d. Main Steam Line ,

a Flow - High s 108.7 psid s 111.7 psid y' , g; e. Condenser Vacuum - Low 10.5 psia 210.1 psia /s_10.9 psia 1

f. Outboard MSIV Room Temperature - High s 192*F s 200*F
g. Turbine Enclosure - Main Steam I Line Tunnel Temperature - High 5 165'F s 175 F
h. Manual Initiation N.A. N.A.
2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
a. Reactor Vessel Water Level Low - Level 3 2 12.5 inches
  • 2 11.0 inches
b. Reactor Vessel (RHR Cut-in Permissive) Pressure - High 5 75 psig 5 95 psig
c. Manual Initiation N.A. N.A.

____ _ _ ______ ___. =__--_ - - _ _ _ __ __ _ -_ _ __ _ -_ _ __ __ ________ _ _ - . _ _ _ _ - _ . _ _ .- _ __.-- --- . - . _ -

1 TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME L TRIP FUNCTION RESPONSE TIME (Seconds)#

1. MAIN STEAM LINE ISOLATION a, ' Reactor Vessel Water Level
1) Low, Low - Level 2 s 13(a)**
2) Low, Low, Low - Level I s 1.0*/s 13(a)**  ;
b. DELETED . DELETED l .
c. Main Steam Line Pressure - Low $ 1.0*/s 13(a)**
d. Main Steam Line Flow - High s 0.5*/s 13(a)** ,
e. Condenser Vacuum - Low N.A.
f. Outboard MSIV Room Temperature - High N.A.

I g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.

h. Manual Initiation N.A.
2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
a. Reactor Vessel Water Level  ;

Low - Level 3 s 13(a) -

b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High H.A.
c. Manual Initiation N.A.
3. REACTOR WATER CLEANUP SYSTEM ISOLATION
a. RWCS a Flow - High s 13## r
b. RWCS Area Temperature - High N.A.
c. RWCS Area Ventilation a Temperature - High N.A.
d. SLCS Initiation N.A.
e. Reactor Vessel Water Level -

Low, Low - Level 2 5 13(a)

f. Manual Initiation N.A.

LIMERICK - UNIT 2 3/4 3-23

i.  !

l \ l 

                                                                                                                'l ;

TABLE 3.3.2-3 (Continued) i ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME  ! TRIP FUNCTION RESPONSETIME(Secondsi#~

f. Outside Atmosphere To Refueling Area a Pressure - Low N.A.  ;

L g. Reactor Enclosure Manual 1 1 Initiation N.A.

h. Refueling Area Manual Initiation N.A.

l TABLE NOTATIONS . (a) Isolation system instrumentation response time specified includes 10 seconds- , diesel generator . starting and 3 seconds for sequence loading delays. 7 (b) DELETED

  • Isolation system instrumentation response time for MSIV only. No-diesel generator delayr assumed for MSIVs. j
                 ** Isolation system instrumentation response time for associated valves except MSIVs.
                                                                                                                 -l
                  # Isolation system instrumentation response time specified for the Trip--

Function actuating each valve group shall be added to isolation time l shown in Tables 3.6.3-1, 3.6.5.2.1-1 and 3.6.5.2.2-1 for valves in.each  ! valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve. . I 

                 ## With 45 second time delay.

LIMERICK - UNIT 2 3/4 3-26

C TABLE 4.3.2.1-l_ E E ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9 OPERATIONAL-i CHANNEL c CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICil TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE'RE0VIRED

1. MAIN STEAM LINE ISOLATION
a. Reactor Vessel Water Level ##
1) Low, Low, Level 2 S Q R 1,2,3  ;

Low, Low, low - Level 1 S Q- R 1, 2, 3 2)

b. DELETED DELETED DELETED DELETED DELETED
c. Main Steam Line w Pressure - Low S M R 1 s.

w d. Main Steam Line Flow - High S M R 1, 2, 3 h

e. Condenser Vacuum - Low S M R 1, 2**, 3**
f. Outboard MSIV Room Temperature - High S M R 1, 2, 3
g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High S M R 1, 2, 3
h. Manual Initiation N.A. R N.A. 1. 2, 3.
2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION
a. Reactor Vessel Water Level ###

1, 2, 3 Low - Level 3 S Q R R 1, 2, 3-

b. Reactor Vessel (RHR Cut-In' S .Q Permissive) Pressure ## - High~
c. Manual Initiation .N.A. R N.A. 1,~2, 3.

TABLE 4.3.2.1-1 (Continued) ,' c- ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 52 CHANNEL OPERATIONAL 53 CHANNEL FUNCTIONAL CHANNEL . CONDITIONS FOR WHICH 

     ?;                                     TRIP FUNCTION                                                                                                                                                                                     CHECK              TEST                                -CALIBRATION                                                                   SURVEILLANCE RE0VIRE0'

, n 

      '                                     7. SECONDARY CONTAINMENT ISOLATION E                                             a.                          Reactor Vessel Water Level ##

Ej low, low - Level 2 S Q R 1, 2,13

b. Drywell Pressure ### - High S Q R 1, 2, 3
c. 1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High S M R *#

4 2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S M R *#

d. Reactor Enclosure Ventilation o, Exhaust Duct Radiation - High S M R 1, 2, 3 s
e. Outside Atmosphere To Reactor Enclosure A Pressure - Low N.A. M Q 1,2,3 i

[

f. Outside Atmosphere To Refueling Area a Pressure - Low N.A. M Q
g. Reactor Enclosure Manual Initiation N.A. R N.A. 1, 2, 3
h. Refueling Area Manual Initiation N.A. R N.A.
  • Required when (1) handling irradiated fuel in the refueling area secondary containment, or (2) during CORE-ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
                                               **When not administratively bypassed and/or when any turbine stop valve is open.
                                                                                                                                                                                                                                               ~
                                                #During operation of the associated' Unit 1 or Unit 2 ventilation exhaust system.
                                               ##These trip functions (la, 2b, 3e, 6a, 6h, and 7a) are common to the ECCS actuation trip function.
                                             ###These trip functions (2a, 6b, and 7b) are common to the RPS and ECCS actuation trip functions.

TABLE 3.6.3-1 . PART A - PRIMARY CONTAINMENT ISOLATION VALVES B INBOARD OUTBOARD ISOL. PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID Ei NUMBER BARRIER BARRIER TIME.IF APP.lF APP.

q (SEC)(26) (20) 003B CONTAINMENT INSTRUMENT 59-2005B (CK) NA 59 GAS SUPPLY - HEADER 'B' HV59-229B 7 C,H S 003D-2 CONTAINMENT INSTRUMENT 59-2112(CK) NA GAS SUPPLY TO ADS VALVES HV59-251B 45 M 59 E&K 007A(B,C,0) MAIN STEAM LINE HV41-2F022A 5* C,E,F,P,Q 6 41. l
                                                                                'A'(B,C,D)                                  (B,C,0)
                ;g                                                                                                                                                        (B,C,D)

HV40-2F001B 45 EA 6 T' (F,K,P) 03 (XV40-201B NA 6,1 (F,K,P) SEE PART B, THIS TABLE) 008 MAIN STEAM LINE DRAIN HV41-2F016 30 C,E,F,P,Q 4 41 HV41-2F019 30 C.E,F,P,Q 009A FEEDWATER 41-2F010A(CK) NA- 41 HV41-2F074A(CK) NA 41-2036A(CK) NA HV41-2308 45 HV41-233A 45 HV41-209A NA 32 HV41-2F032A(CK) NA HV55-2F105 30 7 HV44-2F039(CK) NA (X-98) 41-2016(X-9B, NA 31 X-44). .__ _ _ ._________ _ _ _ _ ___._ _ _._ _ __.._ . _ _ . . . _ _ . . . _ , - - . . . . . . . . . . . . . . . . _ . . , . . , - , _ _ . . ...,...,.,...,........,m. -

r- TABLE 3.6.3-1 (Continued) 32 g PART A - PRIMARY CONTAINMENT ISOLATION VALVES E i s INBOARD OUTBOARD ISOL. O PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), . NOTES P&ID NUMBER BARRIER BARRIER TIME.IF APP. IF APP. (SEC)(26) (20)- 025 DRYWELL PURGE SUPPLY HV57-221(X-201A) 5** B,H,S,U,W,R,T 3,11,14 57, HV57-223 5** B,H,5,U,W,R,T 3,11,14 HV57-209 6** B,H,S,U,W,R,T 11 (X-201A) HV57-231 5** B,H,S,U,W,R,T 11 (X-201A) 3 HV57-235 6** B,H,S,U,W,R,T 11 HYDROGEN RECOMBINER HV57-263 9 B,H,R,S 3,11,14 y 

                                                             "B" INLET                                                  FV57-00-2018                                         90                      B,H,R,S                                     11,34
               ?     026                                     DRYWELL PURGE EXHAUST             HV57-214                                                                      5**                     B,H,S,U,W,R,T                             3,11,14,33 57
               %                                                                               HV57-211                                                                      15**                    B,H,5,U,R,T                                11 SV57-239                                                                      5                                                                   10 HV57-215                                             6**                     B,H,S,U,W,R,T                               11,33 HV57-217                                             5**                     B,H,S,U,R,T                                11 SV57-245                                             5                       B,H,R,S                                     11 HYDROGEN RECOMBINER               HV57-261                                                                      9                       B,H,R,S                                   3,11,14 "A" INLET                                                  FV57-00-201A                                         90                      B,H,R,5                                     11,34 027A                                    CONTAINMENT INSTRUMENT            59-2128(CK)                                                                  NA                                                                                          59 GAS SUPPLY TO ADS VALVES                                   HV59-251A                                            45                      M H,M,&S                                                                                                                                                                                                                                  ,

028A-1 RECIRC LOOP SAMPLE HV43-2F019 10 B 43 l HV43-2F020. 10 B I 028A-2 DRYWELL H2/02 SAMPLE SV57-232 5 B,H,R,S 11 57- l SV57-242 5 B,H,R,S 11. 028A-3. DRYWELL H2/02 SAMPLE SV57-234 5 8,H,R,S- 11- 57-SV57-244 'S- B,H,R,S 11 

                  ~v                                                  -=.,e               - w-   -.iw   ,.n- ..%. - - -    v         m -3    -                    .---    =:,.,-<ws         ..-- ---       --.--sr       . . .,- - . .r- . -. +-                 -i.e..       ,, . . -

                                                                                                                                                                                                                                         . 6 4

h TABLE 3.6.3-1 (Continued) PART A - PRIMARY CONTAINMENT ISOLATION VALVES t C"- 5 INBOARD OUTBOARD- ISOL. PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. SIGNAL (S), NOTES P&ID N NUMBER BARRIER BARRIER TIME.IF APP. IF APP. (SEC)(26) (20) 040G-1 ILRT DATA ACQUISITION 60-2057 NA 11 60 60-2058 NA 11 t 040G-2 ILRT DATA ACQUISITION 60-2071 NA 11 60 60-2070 NA 11 

                                                                             -040H-1       CONTAINMENT INSTRUMENT                            59-2005A(CK)                    NA                                          59 g                                                                        GAS SUPPLY - HEADER 'A'                                         HV59-229A         7            C,H,S
  • p 042 STANDBY LIQUID CONTROL 48-2F007(CK) NA 48 g (X-116) HV48-2F006A 60 29-043B MAIN STEAM SAMPLE HV41-2F084 10 B 41 HV41-2F085 10 B 044 RWCU ALTERNATE 41-2017 NA 5,31 41 RETURN 41-2016(X-9A, NA X-9B)

PSV41-212 NA 045A(B,C,0) LPCI INJECTION 'A'(B,C,D) HV51-2F041A(B,C, NA 9,22. 51 D)(CK) HV51-242A(B,C,. 7 9,22' D) HV51-2F017A 38 (B,C,0) 050A-1 DRYWELL PRESSURE HV42-247B 45 10 42 INSTRUMENTATION 053 DRYWELL CHILLED WATER HV87-228 60 C,H 11 87. SUPPLY - LOOD 'A' HV87-220A 60 CH1 11 

                                                                                                                                                          .HV87-225A'        60           C,H .                -11

_ _ _ _ _ _ _ _ _ _ _ .______ _ _ _ _ __ _ _ _ _ _ . . . _ . - _ _ _ _ _ .._______..._._.___...u.__.__.. ___-_ - _ . ._ _ . _ . . _ . - _, 

                                                                                                                             . I r_

TABLE 3.6.3-1 (Continued) EE [g PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES R c- INBOARD OUTBOARD 55 ISOL. PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL. 

    -d                                                                                             SIGNAL (S), NOTES    P&ID NUMBER                                       BARRIER            6ARRIER
    ^2                                                                               TIME.IF APP. IF APP.

__ (SEC)(26) (20) 003A-I INSTRUMENTATION 'O' -- XV41-2F0700 1 41 MAIN STEAM LINE FLOW XV41-2F0730 003A-2 INSTRUMENTATION 'A' -- XV43-2F003A 1 43 RECIRC PUMP SEAL PRESSURE 003C-1 INSTR. - HPCI STEAM FLOW -- XV55-2F024A 1 55 003C-2 INSTR. - HPCI STEAM FLOW 

  }{                                                      --

XV55-2F024C 1 55 oi 003D-1 INSTR. 'A' MAIN STEAM -- XV41-2F070A 1 41 3: LINE FLOW XV41-2F073A , 007A(B,C,D) INSTR. 'A'(B,C,D) MAIN (HV41-2F022A(B, 5* C,E,F,P,Q 6 - 41 ~ STEAM LINE PRESSURE C,0) SEE PART A (HV41-2F028A 5* C,E,F,P,Q 6 THIS TABLE) (B,C,0) SEE PART A THIS TABLE) (HV40-2F0018 45 EA 6 (F,K,P) SEE PART A THIS TABLE) XV40-201B(F, 1,6 K,P) 020A-1 INSTR - RPV LEVEL -- XV42-2F045B 1 42 020A-2 INSTR 'B' LPCI DELTA P -- XV51-2028 1 51-020A-3. INSTR 'D' LPCI-DELTA P -- XV51-2038 1 51 020B-1 INSTR RPV LEVEL -- XV42-2F045C 1 42 0208-2 INSTR. 'C' LPCI DELTA P -- 

                                                                     'XV51-202C                                    1    51
 'I                                                                                            '

l l' . . !r 3/4.3 INSTRUMENTATION BASES f 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to: l l a. Preserve the integrity of the. fuel cladding. l

b. Preserve the integrity of the reactor coolant system.
c. Minimize the energy which must be adsorbed following a l l loss-of-coolant accident, and i
d. Prevent inadvertent criticality.

This specification provides the limiting conditions for operation  ! necessary to preserve the ability of the system to perform its intended , function even during periods when instrument channels may be out of service > because of maintenance. When necessary', one. channel may be made inoperable -{ for brief intervals to conduct required surveillance. The reactor protection system is made up of two independent trip systems. I There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. -The tripping of r both trip systems will produce a reactor scram. The system meets the intent t of IEEE-279 fcr nuclear power plant protection systems. Specified  : surveillance intervals and surveillance and maintenance outage' times have been '; determined in accordance with NEDC-30851P, " Technical Specification  ! Improvement Analyses for BWR Reactor Protection System," as approved by the .; NRC and documented in the NRC Safety Evaluation Report (SER) (letter to T. A. ' Pickens from A. Thadani dated July 15, 1987. The bases for the trip settings of RPS are discussed in the bases for Specification 2.2.1. q Automatic reactor trip upon receipt of a high-high radiation signal j from the Main Steam Line Radiation Monitoring System was removed as the result ] of an analysis performed by General Electric in NED0-31400A. The NRC approved _j the results of this analysis as documented in the SER (letter to George J. Beck, n BWR Owner's Group from A. C. Thadani, NRC, dated May 15, 1991). 9 The measurement of response time at the specified frequencies provides , assurance that the protective functions associated with each channel are  ! completed within the time limit assumed in the safety analyses. No credit was 1 taken for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping or ' total channel test measurement, provided such tests demonstrate the total-channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or i (2) utilizing replacement sensors with certified response. times. l l i LIMERICK - UNIT 2 B 3/4 3-1 1  ;

l 

  -           INSTRUMENTATION BASES 3/4.3.2      ISOLATION ACTUATION INSTRUMENTATION                                                      :

1 This specification ensures the effectiveness of the instrumentation used to i mitigate the consequences of accidents by prescribing the OPERABILITY trip i setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required 4 surveillance. l t Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 2, " Technical jl Specification Improvement Analysis for BWR Instrumentation Common to RPS and ECCS Instrumentation" as approved by the NRC and documented in the NAC Safety 4 Evaluation Report (SER) (letter to D. N. Grace from C. E. Rossi dated January 6,  ! 1989). j Automatic closure of the MSIVs upon receipt of a high-high radiation  ! j signal from the Main Steam Line Radiation Monitoring System was removed as the  ! i result of an analysis performed by General Electric in NEDO-31400A. The NRC . approved the results of this analysis as documented in the SER (letter to i George J. Beck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15, 1991). { Some of the trip settings may have tolerances ex licitl stated where both the high and low values are critical and may have a substanti 1 effect on  ! J safety. The setpoints of other instrumentation, where only the high or low end j i of the setting have a direct bearing on safety, are established at a level away  ! from the normal operating range to prevent inadvertent actuation of the systems  ! involved.  ! Except for the MSIVs, the safety analysis does not address individual sensor l response times or the response times of the logic systems to which the sensors 3 are connected. For D.C. operated valves, a 3 second delay is assumed before the 1 valve starts to move. For A.C. operated valves, it is assumed that the A.C.

power supply is lost and is restored by startup of the emergency diesel j generators. In this event, a time of 13 seconds is assumed before the valve J starts to move. In addition to the pipe break, the failure of the D.C. operated is concurrent with the
valve is assumed; 10-second thus the diesel startup andsignal the 3de.ay second (sensor response) load center loa ding delay.The safety

, analysis considers an allowable inventory loss in each case which in turn l determines the valve speed in conjunction with the 13-second delay. It follows -

that checking the valve speeds and the 13-second time for emergency power

establishment will establish the response time for the isolation functions. Operation with a trip set less conservative than its Trip Setpoint but

within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance  :

i for instrument drift specifically allocated for each trip in the safety 1 l analyses. ' 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the , ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and resrsnse times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are i listed by system, in some cases the same instrument may oe used to send the 4 actuation signal to more than one system at the same time. i LIMERICK - UNIT 2 B 3/4 3-2

i INSTRUMENTATION l l BASES  : Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D. . N. Grace from A. C. Thadani dated December 9, 1988 (Part 1) and letter to D. N. ~ Grace from C. E. Rossi dated December 9, 1988 (Part 2)). Operation with a trip set less conservative than its Trip Setpoint but  ! within its specified Allowable Value is acceptable on the basis that the ' difference between each Trip Setpoint and the Allowable Value is an allowance  : for instrument drift specifically allocated for each trip in the safety j analyses.  ! 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram

system provides a means of limiting the consequ(ATWS) recirculation pump tripences of th '

of a failure to scram during an anticipated transient. The response of the l plant to this postulated event falls within the envelope of study events in .j General Electric Company Topical Report NED0-10349, dated March 1971, NED0- . 24222, dated December 1979, and Section 15.8 of the FSAR. j k The end-of-cycle the reactor trip. Duringrecirculation turbine trip pump and trip (E0C-RPT)d generator loa rejection events, thesystem is - ECC-RPT will reduce the likelihood of reactor vessel level decreasing to level

2. Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.

A fast closure sensor from each of two turbine control valves provides input to the E0C-RPT system; a fast closure sensor from each.of the other two turbine  : control valves provides input to the second E0C-RPT system. Similarly, a + position switch for each of two turbine stop valves provides input to one E0C-  ! RPT system; a position switch from each of the other two stop valves provides ' input to the other EOC-RPT system. For each E0C-RPT system, the sensor relay 1 contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop ialves. The j operation of either logic will actuate the E0C-RPT system and trip both [ recirculation pumps. Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administrative 1y controlled. The manual bypasses and the automatic Operating . Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control I room. 1

The EOC-RPT system response time is the time assumed in the analysis between
initiation of valve motion and complete suppression of the electric arc, i.e.,

175 ms. Included in this time are: the response time of the sensor, the time

allotted for breaker arc suppression, and the response time of the system logic.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses. LIMERICK - UNIT 2 B 3/4 3-3 

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