ML20116L270

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Proposed Tech Specs,Revising TS Sections 3/4.3.1,3/4.3.2, 3/4.3.3 & Associated TS Bases Sections 3/4.3.1 & 3/4.3.2 to Eliminate Selected Response Time Testing Requirements
ML20116L270
Person / Time
Site: Limerick  Constellation icon.png
Issue date: 08/08/1996
From:
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20116L256 List:
References
NUDOCS 9608160192
Download: ML20116L270 (17)
v • d • e


Text

- _ - _ _ _ _ _ _ - - _ _

ATTACHMENT 2 UMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos. 50-352 50-353 Uconse Nos. NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 96 09-0 UST OF AFFECTED PAGES Unit 1 Unit 2 3/43-6 3/43-6 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4339 3/4339 83/43-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2 9608160192 960808 2 PDR ADOCK 0500 l

TA8LE 3.3,1-2 i ' r REACTOR PROTECTION SYSTEM RESPONSE TIMES i

' RESPONSE TIME g ISeconds)

FUNCTIONAL UNIT

1. Intermediate Range Monitors:

l 7 N.A.

a. Neutron Flux - High

. - .E N.A.

  • b. Inoperative
2. Average Power Range Monitor *:

N.A.

a. Neutron Flux - Upscale, Setdown
b. Neutron Flux - Upscale 50.09
1) Flow Blased

$0.09

2) High Flow Clamped N.A.
c. Inoperative N.A.

2 d. Downscale 0.j5 Reactor Vessel Steam Dome Pressure - High j 3.

Reactor Vessel Water Level - Low, level 3 sh05

- 4.

50'io6 4

, 5. Hafn Steam Line Isolation Valve - Closure DELETED l ,

6

6. DELETED N.A.

Drywell Pressure - High

.h 7.

Scram Discharge Volume Water Level - High

- 8. N.A.

,, en a. Level Transmitter N.A.

. b. Float Switch

,y 50.06

9. Turbine Stop Valve - Closure

-m

Trip 011 Pressure - Low N.A.

11. Reactor Mode Switch Shutdown Position N.A.
12. Manual Scram Response time shall be measured
  • Neutron detectors are exempt from response time testing.from the detector output or from th

< ** Neasured  := ~ from_ v start ---__~n of turbine control valve fast. closure. - _ _ _ - _

, ll1 g g k I

'* N ""** '"#* rerneining channel including tr%Iunit and relay logic are

)

~

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

1. MAIN STEAM LINE ISOLATION
a. Reactor Vessel Water Level N 1 Low, Low - Level 2  : " ' ;" t4. A . _ )

z)l Low, Low, low - Level 1 - c'

b. DELETED s 1.0{

del.ETE l j

c. Main Steam Line ^ ~ ^

Pressure - Low '

d.

s 1.0{

Main Steam Line Flow - High

e. Condenser Vacuum - Low 50.(

N.A. -

f. Outboard MSIV Room I Temperature - High N.A.
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 q,Q ,

1 i ' ^ ' l- 14. A._ 4

b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High N.A.
c. Manual Initiation N.A. ,
3. REACTOR WATER CLEANUP SYSTEM ISOLATION e_
a. RWCS a Flow - High -

mesif d,A

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  % . i

f. Manual Initiation N.A.

FEB161995 LIMERICK - UNIT 1 3/4 3-23 Anendment No. 29, 89

1

, TABLE 3.3.2-3 (Continued) .

i ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION

  • RESPONSE TIME (Seconds)#

4.

HIGHPRESSdREC06LANTINJECTIONSYSTEM 150LAIION .

a. HPCI Steam Line  %

A Pressure - High l b. '

f Ji R A.) '

HPCI Steam Supply Pressure - Low . _ ,,3 g--- ed . A. _

c. m r HPCI Turbine Exhaust Diaphragm Pressure - High 'uv' N.A.

)

l d. HPCI Equipment Room Temperature - High N.A.

e. HPCI Equipment Room A Temperature - High -

N.A.

f. HPCI Pipe Routing Area - '

Temperature - High N.A.

g. Manual Initiation N.A.

5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION  !

a. RCIC Steam Line -

f,3 A Pressure - High y" r1 A. N i

, .s

b. RCIC Steam Supply Pressure - Low - --

~ )J. A >

c. RCIC Turbine Exhaust Diaphrage Pressure - High

'j N.A.

.. d. RCIC Equipment Room Teeperature - High N.A.

e. RCIC Equipment Roos. .

A Temperature - High N.A.

i

f. RCIC Pipe Routing Area '

Temperature - High N.A.

g. Manual Initiation N.A. .

L1HERICK - LatIT 1 3/4 3-24 . Amendment No. 33 BC138IE 3

r - _.

TABLE 3.3.2-3 (Continued)

ISGLATION SYSTEM INST'KUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (SecondsM

6. PRIMARY CONTAINMENT ISOLATION
a. Reactor Vessel Water Level Low, Low - Level 2 O 1)
2) Low, Low, Low - Level 1 M l,X{ y A.
  1. !!W2). g,
b. Drywell Pressure - High TueBW J4 . A. . .
c. North Stack Effluent Radiation - High N.A.
d. Deleted i e. Reactor Enclosure Ventilation Exhaust Duct - Radiation - High N.A.
f. Deleted  !
g. Deleted
h. Drywell Pressure - High/

Reactor Pressure - Low N.A.

1. Primary Containment Instrument Gas to N.A.

Drywell a Pressure - Low

j. Manual Initiation N.A.
7. SECONDARY CONTAINMENT ISOLATION j
a. Reactor Vessel Water Level Low, Low - Level 2 N.A.
b. Drywell Pressure - High M.A.

l c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High N.A.

2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High N.A. 1
d. Reactor Enclosure Ventilation Exhaust Duct Radiation - High N.A.
e. Deleted

_ l l LIMERICK - UNIT 1 3/4 3-25 Amendment 23,112

' - ~

FEB 21 1996 i

1

.- ve. m s -

i

\

.e. . . _ . - -.-.- ._.. - -_. - -. .

I l.

TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INS 1RtJMENIAT10N RESPONSE TIME i TRIP FUNCTION RESPONSE TIME (Seconds)#

f. Deleted
g. Reactor Enclosure Manual Initiation N.A.
h. Refueling Area Manual Initiation N.A.

TABLE NOTATIONS (a) Isolation system instrumentation response time specified includes 10 seconds A

diesel generator starting and 3 seconds for sequence loading delays.

(b) DELETED Isolation system instrumentation response time for MSIV only. No diesel generator delays assumed for MSIVs.

h Isolation system instrumentation response time for associated valves except MSIVs.

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 valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

    1. With 45 second time delay.

e-, . , . - - -

      1. Sensor is eliminated from response time testing for the MSN actuation logic circuits. Response time testing and conformance to administrative limits for the remaining channel including trip unit and relay logic are required.

LIMERICK - UNIT 1 3/4 3-26~ Amendment No. 6,89,112 FEB 21 1996 i n .

n , - .m

a

  • l TABLE 3.3.3-3 '

l EMERGENCY CORE C00 LING SYSTEM RESPONSE TIMES EG1 RESPONSE TIME fSeconds)

1. CORE SPRAY SYSTEM $ 27
2. LOW PRESSURE COOLANT INJECTION MODE

~

0F RHR SYSTEM s 40 4'

3. AUTCMATIC DEPRESSURIZATION SYSTEM N . A ._.

s 60Y

~

4. HIGH PRESSURE COOLANT INJECTION SYSTEM l
5. LOSS OF POWER N.A.

. , / ,._ , ~ -

  1. ECCS actuation instrumentation is eliminated from response time testing.

l l

1

_~ -

LIMERICK - UNIT 1 3/4 3-39 Amendment No. 102 NOV 0 31995

3/4.3 INSTRUMENTATION

{.

3 BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION -

The reactor protection system automatically initiates a reactor scram to:

I a. Preserve the integrity of the fuel cladding.

) b. Preserve the integrity of the reactor coolant system.

c. Minimize the energy which must be adsorbed following a loss-of-coolant accident, and

' d. Prevent inadvertent criticality.

This specification provides the limiting conditions for operation i 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 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 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 Sgecification Improvement Analyses for BWR Reactor Protection System, as approved by the letter to T. A.

NRC Pickens and fromdocumented A. Thadant dated in the JulyNRC Safety 15, 1987. The Evaluation bases for Report (SER)th(e trip of RPS are discussed in the bases for Specification 2.2.1.

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 NEDO-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. Thadant, NRC, dated May 15,1991).

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 applicable. i Response time may be demonstrated by any series of sequential overlapping or i total channel test measurement, provided such tests demonstraf.e the total l channel response time as defined. Sensor response time verification may be 4 demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

Response time testing for the sensors as noted in Table 3.3.1-2 is not required based on the analysis in NEDO-32291-A. Response time testing for the remaining channel components is required as noted.

- - 1 FEB 161995 LIMERICK - UNIT I B 3/4 3-1 Amendment No. 53, 89

~

'. INST'RUMENTATION i BASES

. 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION t

i This specification ensures the effectiveness of the instrumentation used to tsitigate the consequences of accidents by prescribing the OPERABILITY trip i setpoints and response times for isolation of the reactor systems. When l necessary, one channel may be inoperable for brief intervals to conduct required 4

surveillance.

~

l Specified surveillance intervals and maintenance omge times have been i

determined in accordance with NEDC-30851P Supplement 2 *lechnical Specification Improvement Analysis for BWk Instrumentation Common to RPS and 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) and NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation " as approved by the NRC and documented in the NRC SER (letter to S. D. Floyd from C. E. Rossi dated June 18,1990).

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 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 George J. Beck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15,1991).

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 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 response times or the response times of the logic systems to which the sensors 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 starts to move. In addition to the pipe break, the failure of the D.C. operated is concurrent with the valve 10-second is assumed; thus the diesel startup andsignal the 3delay (sensor second load response) 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 isolstion functions.

Operation with a trip set less conservative than its Trip Setpoint but uthin 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.

3/4.3.3 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 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.

]

~

FEB 1 6 IW5 ERICK - UNIT 1

~

R 1/4 3 2 ' Q- t- M. 33, Md

~ -

Response time testisg for sensors are not required based otithe analysis in NEDO 32291-A. Response time l testing of the remaining channel components is required as.noted in Table 3.3.2-3._. ._ _

- -_ -, .-_n_,---

r_ .

I I =

4 .

3 TABLE 3.3.1-2  :

h m

REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME

5 FUNCTIONAL UNIT fSeconds)

.n 7 1. Intermediate Range Monitors:

g a. Neutron Flux - High N.A.

U b. Inoperative N.A. '

m

2. Average Power Range Monitor *:
a. Neutron Flux - Upscale, Setdown N.A.

}

b. Neutron Flux - Upscale

,, h hg5 k lamped sN

c. Inoperative N.A.
w. d. Downscale N.A.

s.

  • Reactor Vessel Steam Dome Pressure - High 50.55 3.

e 4. ,

Reactor Vessel Water Level - Low, level 3 05

5. ' Main Steam Line Isolation Valve - Closure 50.06
6. DELETED DELETED l N.A. I
7. Drywell Pressure - High 2

so

8. Scram a.

lsc

{evegaroeVo'uneWaterLevel-High TPansm'tter N.A.

. . b. Float Switch N.A. j

9. Turbine Stop Valve - Closure 50.06  ;

irie

  • 10. Tufr'ne p0 gngl ssure Valve .ow ast Closure,

$0.08**

'" N.A.

11. ' Reactor Mode Switch Shutdown Position
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 component in the channel.

- _* M from start of turbine control valve fay closure.

iD , s sensor is eliminssed from response tkne testing for the RPS circuits. Response time testing and r conformance to administrative limits for the remaining channel including trip unit and relay logic are .

.m mquired.

s- -

. -_- .___-____ ____=__~_______-___ ___ ___ _

l TABLE 3.3.2-3

, . . , l ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

1. MAIN STEAM LINE ISOLATION
a. Reactor Vessel Water Level 1 Low, Low - Level 2 ede(u$ner M.A _

2 Low, Low, Low - Level I s 1. 0* ' ' " ' ' "

)

I C tt'O 1 h LETE M ]

b. DELETED
c. Main Steam Line -

Pressure - Low s 1.0* ': " ? '"

k,4 dtt

d. Main Steam Line  :

Flow - High s 0.5*/ M i

e. Condenser Vacuum - Low N.A?

Niidt'

f. Outboard MSIV Room Temperature - High N.A.
g. Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.
h. Manual Initiation N.A.
2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION 1
a. Reactor Vessel Water Level Low - Level 3 Ab EA+ -

b.

A 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 1p4Sif M. A -

.M

b. RWCS Area Temperature - High
c. RWCS Area Ventilation A Temperature - High N.A.
d. SLCS Initiation N.A.
e. Reactor Vessel Water Level -

Low, Low - Level 2 "

- - ,_ AA .- 2

f. Manual Initiation .A O FB1625 LIMERICK - UNIT 2 3/4 3-23 W = t No. 52

~

~ a . - , , a n g . ; . . +, ,

- TABLE 3.3.2-3 (Continued) .

ISOLATION SYSTEM INSTRUNENTATION RESPONSE TIE TRIP FWCTION RESPONSE' TIME (Seconds M '.

4. NIGN PRESSURE COOLANT INJECTION SYSTEM 15DLATION nm'

~

a. IFCI Steam Line A Pressure - High M NA
b. $PCI itea's Supply

_r 3 Pressure - Law -

A A. --

c. HPCI Turbine Exhaust Diaphragm Pressure - High N. A.
d. HPCI Equipment Room -

Temperature - High N.A.

e. IFCI Equipment Room A Temperature - High. N.A.

i f. IFCI Pipe Routing Area Tamperature - High . N.A.

~ .

~

g. Manual Initiation N. A.
5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
a. RCIC Steam Line A Pressure - High 2

9 r,3 _d A. -

b. RCIC Steam Supply Pressure - Law

' i # H.A.

c. RCIC Turbine Exhaust Diaphragm Pressure - High N.A.

I d. RCIC Equipment Room Temperature - High K. A.

e. JtCIC Equipment Room

& Temperature - High N. A.

f. RCIC Pipe Routing Area -

Temperature - High M.A.

g. Manual Initiation , M. A.

e J - -

e LDELICK - UNIT 2 3/4 3-24

~

a25m

<. . 2. .

, .. i j

TABLE 3.3.2-3 (Continued) l 1 ISOLATION SYSTEM INSTRt3tENTATION RLSMNSE TIME TRIP FUNCTION RESPONSE TINE (Secondsif

6. PRIMARY CONTAlletENT ISOLATION
a. Reactor Vessel Water Level
1) Low, Low - Level 2 -5458 NA'-- -

{_ 2) Low, Low, Low - Level 1 M.S A._ .

! b. Drywell Pressure - High M dA i

c. North Stack Effluent l Radiation - High M.A.

i d. Deleted i i l I

e. '

..eactor Enclosure Ventilation Exhaust 4 l Duct - Radiation - High N.A. l 1

f. Deleted i
g. Deleted
h. Drywell Pressure - High/ i'

~

Reactor Pressure - Low N.A.

1. Primary containment Instrument Gas to Drywell A Pressure - Low .

N.A.

j. Manual Initiation N.A.
i. SEC0lN1ARY CONTAllstENT ISOLATION
a. Reactor Vessel Water Level Low, Low - Level 2 N.A.
b. Drywell Pressure - High N.A.

c.1. Refueling Area Unit 1 Ventilation .

" Exhaust Duct Radiation - High N.A.

2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High N.A.

"- ~

d. Riactor Enclosure Ventilation Exhaust Duct Radiation - High N.A.
e. Deleted l

LIMERICK - tal!T 2 3/4 3-25 "

- Amendment No. 74 FEB 21 I!!96

.,.x;

'4 a w.x s 4

. , w&~

72Nw

1. '

~

~

i k

l TABLE 3.3.2-3 (Continued)

}

j ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME

! TRIP FUNCTION RESPONSE TIME (SecondsW i

j f. Deleted l l g. Reactor Enclosure Manual i Initiation N.A.

}

! h. Refueling Area Manual Initiation N.A.

i i

~_ -

N TABLE NOTATIONS

~

[ i

(a) Isolation system instrumentation response time specified includes 10 seconds l diesel generator starting and 3 seconds for sequence loading delays.

i l (b) DELETED i j Isolation system instrumentation response time for MSIV only. No diesel rator delays assumed for MSIVs.

} **

Isolation system instrumentation response time for associated valves
except MSIVs. _

> 1 i # Isolation system instrumentation response time specified for the Trip 3 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 j valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

! ## With 45 second time delay.

1 i

s l' l ### Sensor is eliminated from response time testing for the MSIV actuation logic circuits. Response time

, testing and conformance to administrative limits for the remaining channel including trip unit and i relay logic are required.

i - - --

- - - , . . - . , - . , -  ; 3-n; 3 -

. . .; - , . 3.,. y.,.

~

LIMERICK - UNIT 2 -

'3/4 3 -Amendment No. 52.74 FEB 21 1996

... u ,

a- _

j. . .

4 4

TABLE 3.3.3 3 l

EMERGENCY CORE ctvu ING SYSTEM' RESPONSE TIMES i

i E RESPONSE TIME (Seconds)

1. CORE SPRAY SYSTEM s 27 4
2. LOW PRESSURE COOLANT INJECTION MODE -

2 0F RHR SYSTEM s 40

3. AUTOMATIC DEPRESSURIZATION SYSTEN N. . "
4. HIGH PRESSURE COOLANT INJECTION SYSTEN * [s 60 g l

\

5. LOSS OF POWER N.A.

f-

f. _C T
# ECCS actuation instrumentation is eliminated _from response time testing.

m '

N LIMERICK - UNIT 2 3/4 3-39 Amendment No. 66 D03%

, .. 3/4.3 INSTRUMENTATION

... o BASES 4

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION l

The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cladding.

b.

l Preserve the integrity of the reactor coolant system.

c. Minimize'the energy which must be adsorbed following a s loss-of-coolant accident, and '

d.

, Prevent inadvertent criticality.

M l 4'.

This specification provides the limiting conditions for operation . . .

necessary to preserve the ability of the system to perform its intended 3C "

function even during periods when instrument channels may be out of service l 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 i

i 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 .

i of IEEE-279 for nuclear power plant protection systems. Specified .%% -

i I (

surveillance intervals and surveillance and maintenance outaae times have been i

- determinedinaccordancewithNEDC-30851P,"TechnicalSpcifIcation drh

Improvement Analyses for BWR Reactor Protection System, as a proved by the NRC and documented in the NRC Safety Evaluation Report (SER) letter to TL A';

Pickens from A. Thadant dated July 15, 1987. The bases for t e trip settings

{ of RPS are discussed in the bases for Specification 2.2.1.

{

i

wx .

u w kusun g &w .: -

a 7:iN nA&

Automatic reactor trip upon receipt of a high-high radiation signal' e i

i 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 l

the results of this analysis as documented in the SER (letter to George JPBeck, i BWR Owner's U. Group from 32.m dated May 15,1991).

1< A. C. Thadani, NRC,N w 3 fy M M g 2

.w ^ ,.

3.*M 3 frequencies prov s W:

! The mea @surement of response, time at the y i sWcified

?

i assurance that the protective functions assoctafed'with each channel are"*'i&N;" 4 '

I

)

completed within the time limit assumed in the safety analyses. No credit'was 4-taken for those channels with response times indicated as not a 'W' l

Response time may be demonstrated by any series of ~ sequential, pplicable.

overlapping.or

! total channel test measurement, provided such tests demonstrate the total i

channel response time 'as defined. Sensor response time verification may

! demonstrated by either (1) inplace, onsite or offsite or "test me (2) utilizing replacement w sensors with certified response times. "*:F. ..

, . -_t.o17 p .

w g. jr0 -

Gy j h -

"+

. . , _ . .n .-. - - - . - ~ ~ . . . , -

- - - - - ~ ~ -- -

3 i Response time testing for the sensors as noted in Table 3.3.12 is not required based on the analysis in l NEDO 32291-A. Response time testing for the remaining channel components is required as noted.

( _- - ,. -+ , 4 . ,. ,-. ,

i g , g, g .

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i FEB 161995 7 k' LIMERICK - UNIT 2 B'.3/4 3-1 Amendment No. 17, 52'

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/ ,f i:x * ;c" % + v .

_.~ .a . - { { . L:jX s @. . . p~ *f v, Q - ~

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-

  • INSTRUMENTATION BASES
3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION

' 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. -

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 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)

Isolation andActuation NEDC-31677P-A{ru" Ins mentation Technical" asSpecification a Improvement Analysis for BWR the NRC SER (letter to S. D. Floyd from C.pproved by the NRC E. Rossi dated Juneand documemnted in 18,1990).

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 result of an analysis performed by General Electric in NEDO-31400A. The NRC approved the results or 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).

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 instrumenta". ion, where only the high or low end 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

. response times or the respo,nse times of the logic systems to which the sensors 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 starts to move. In addition to the pipe break valve is assumed; thus the signal delay (sensor, response the failureisofconcurrent the D.C. operated with the 10-second diesel startup and the 3 second load center loa) ding delay. The safety analysis considers an allowable inventory loss in each case which in turn e i determines the valve speed in conj unction 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 tiut 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.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the conseguences 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 grovide the design protection. Although the instruments are ~(

listed actuationby system,to signal more than one system at the same time.n some cases the same ins FEB 161995 MERICK--UNIT .72 ._- . .

4 3/4-3 ~

" r Li. No

~

Response tirne testing for sensors are not required based on the analysis in NEDO42291-A. Response tirne ,

c h testing of the rgchannel cornponents is required as noted in Table 3.3.24 __ .

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