Text

Amends 53 & 17 to Licenses NPF-39 & NPF-85,respectively, Extending Surveillance Test Intervals & Allowable out-of-svc Times for Instrumentation Supporting RPS & ECCS
	ML20086K543
Person / Time
Site:	Limerick
Issue date:	12/02/1991
From:	Chris Miller; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20086K546	List: ML20086K543; ML20086K562;
References
	NUDOCS 9112130161
	Download: ML20086K543 (86)
	v • d • e

~

c-.

T e~

1 i g

(bf

.p

~

o,f UNITED STATES g-NUCLEAR REGULATORY COMMISSION 5-M E WASHINGTON, D C,20555

'4'4...../

PHILADELPHIA ELECTRIC COMPANY I

DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT 1 AhTNOMENT TO FACILITY OPFRATING LICENSE Amendment flo. 53 License No. NPF-39 l '. -

The Nuclear Regulatory Comission (the Comission) has found that A.

The application for amendment by Philadelphia Electric Company

.(the licensee) dated April 26, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; i

C; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering'the health and safety of the'public, and -(ii) that such activities will be conducted in compliance with the Comission's regulations; l'

D.-

The issuance of this amendment will not be inimical to the comon defense and. security or to the health.and safety of the public; and E..

The issuance of this emendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated-in the attachment to this license amendment, and paragraph 2.C.(2)_

of F_acility Operating License No. NPF-39 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 53 _,-are hereby incorporated into this

-license. Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

9112130161 911202 DR ADOCK 050 2

n.

i

-2 3.

This license amendn:ent is effective fif teen (15) days af ter date of issuance.

FOR THE NUCLE REGULATORY COMMISSION Ciaries L Miller, D rector Project Directorate 1-2 Division of Reactor Projects - 1/11

Attachment:

Changes to the Technical Specifications Date of Issuance: December 2, 1991 9

e

ATTACHMENT TO LICENSE AMENDMENT NO. 53 FACILITY OPERATitlG LICENSE NO. fiPF-39 DOCKET NO. 50-352 Replace the following pages of the Appendix A Technical Specifications with the-ettached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.*

Remove Insert XiX XiX XX XX*

3/4 3-1 3/4 3-1 3/4 3-2 3/4 3-2*

3/4 3-5 3/4 3-5 3/4 3-6 3/4 3-6*

3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 3/4 ' 9 3/4 3-9 3/4 3-10 3/4 3-10 3/4 3-15 3/4 3-15*

3/4 3-16 3/4 3-16 3/4 3-17 3/4 3-17 3/4 3-18 3/4 3-18*

3/4 3-27 3/4 3-27 l

3/4 3-28 3/4 3-28 l

3/4 3-29 3/4 3-29*

l 3/4 3-30 3/4 3-30 l

3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32*

3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-39 3/4 3-39*

3/4 3-40 3/4 3-40 3/4 3-41 3/4 3-41 l

3/4 3-42 3/4 3-42*

. ATTACHMENT TO LICENSE AMENDMENT NO. 53 FACILITY OPERATING LICENSE NO. NPF-39 DOC,KE,T NO, 50-352 Remove Insert 3/4 3-53 3/4 3-53 3/4 3-54 3/4 3-54 3/4 3-55 3/4 3-55*

3/4 3-56 3/4 3-56 3/4 3-61 3/4 3-61 3/4 3-62 3/4 3-62*

B 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2 B 3/4 3-3 B 3/4 3-3 B 3/4 3-4 B 3/4 3-4 B 3/4 3-5 B 3/4 3-5 B 3/4 3-6 B 3/4 3-6*

-.i l

BASES SECTION 1

PAGE INSTRUMENTATIONt(Continued)

Seismic Monitoring _ Instrumentation..........................

B 3/4 3-5 l

(0eleted)...................................................

B 3/4 3-5 1 j Remote Shutdown System ?nstrumentation and Controls.........

B 3/4 3-5 Accident Monitoring Instrumentation.........................

B 3/4 3 Source Range Monitors.......................................

B 3/4 3-5 Traversing-In-Core _ Probe System.............................

B 3/4 3 Chlorine and Toxic Gas Detection Systems....................

B 3/4 3-6

' Fire Detection Instrumentation..............................

B 3/4 3-6

Loose-Part Detection System.................................

B 3/4 3-6 s -

-(0eleted)...................................................

B 3/4 3-6 Offgas Monitoring Instrumentation...........................

B 3/4 3-7 3/4.3.B -TURBINE OVERSPEED PROTECTION SYSTEM.........................

-B 3/4 3-7 3/4.3.9-FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.............................................

_B 3/4.3-7 I

Bases Figure'B 3/4.3-1 Reactor Vessel Water leve1...........................

B-3/4 3-B 3f4.4 REACTOR @0LANT SYSTEM-3/4.4.1 RECIRCULATION SYSTEM........................................

B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES........................................

B 3/4 4-2 I

L3/4.4.3 _ REACTOR' COOLANT SYSTEM LEAKAGE Leakage Detection Systems...................................

B 3/4 4-3 i-l-

. Operational Leakage.........................................

B 3/4 4 3/4.4.4 ~ CHEMISTRY...................................................

B 3/4 4-3 LIMERICK - UNIT 1 xix Amendment No. 48, 53,

INDEX BASES

~SECTION P60.E REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY.......................................

B 3/4 4-4

-3/4.4.6 PRESSURE / TEMPERATURE LIMITS......................

8 3/4 4-4 Bases Table B 3/4.4.6-1 Reactor Vessel l

Toughness.................

B 3/4 4-7 Bases Figure 8 3/4.4.6-1 Fast Neutron Fluence L

(E>l MeV) At 1/4 T As A Function of Service t

Life......................

B 3/4 4-8 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................

B 3/4 4-6

.3/4.4.8 STRUCTURAL INTEGRITY....................................

B 3/4 4-6

(

3/4.4.9 RESIOUAL HEAT REM 0 VAL................................... 8 3/4 4-6 3/4.5 EMERGENCY CORE COOLING SYSTEMS p

3/4.5.1.and 3/4.5.2-ECCS - OPERATING and SHUTDOWN............

B 3/4 5-1 3/4.5.3 SUPPRESSION CHAMBER................................

o L

8 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS

3/4.6.-

PRIMARY CONTAINMENT-b j.

Prima ry Contai nment Integri ty...................... -

L

_B 3/4 6-1 Pri mary Containment-Leakage........................

B 3/4 6-1 Primary Containment Ai r Lock.......................

B 3/4 6-1 l

MSIV Leakage Control System........................

B 3/4 6-1 Primary Containment Structural Integri ty...........

B 3/4 6-2 l~

Drywell and Suppression Chamber Internal Pressure.........................................

B 3/4 6-2 p

Drywell Average Air Temperature....................

B 3/4 6-2 o

i_

Drywell'and Suppression Chamber Purge System......,

8 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS...........................

B 3/4 6-3 LIMER OK - UNIT 1 xx Amendment No. 33 DCT 3 0 E69 m

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

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

l APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

a.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The provisions of Specification 3.0.4 are l

not applicable.

b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and take the ACTION required by Table 3.3.1-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every h times 18 months where N is the total number of redundant channels in a specific reactor trip system.

An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, the inoperable channel shall be restored to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or the l

ACTION required by Table 3.3.1 1 for that Trip Function shall be taken.

- The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.

LIMERICK - UNIT 1 3/4 3-1 Amendment No. 53

TABLE 3.3.1-1 REACTOR PROTECTION' SYSTEM INSTRUMENTATION n

n APPLICABLE-MINIMUM g

OPERATIONAL OPEPABLE CHANNELS q FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)

ACTION 1.

Intermediate Range Monitors (b),

a.

Neutron Flux - High 2

3 1

3, 4 3

2 5(c) 3(d) 3 b.

Inoperative 2

3 1

2 3, 4 3

2 5

3(d) 3 2.

Average Power Range Monitor ('):

b a.

Neutron Flux - Upscale,.Setdown 2

2 1

3 2

2 5(c)(1) 2(d) 3 l

p b.

Neutron Flux - Upscale N

1) Flow Biased 1

2 4

c_. E

2) High Flow Clamped

.1 2

4 r2 c

a c.

Inoperative 1, 2 2

1 3

2 2

g 5(c)(1) 2(d) 3 l

h d.

Downscale 1(g) 2 4

3.

Reactor Vessel Steam Dome Pressure - High 1,2(f) 2 1

4.

Reactor Vessel Water Level - Low, Level 3 1, 2 2

1 5.

Main Steam Line Isolation Valve -

Closure 1(g) 1/ valve 4

~

i

- ~ _.

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a);-

A channel may be placed in an inoperable status for up to 0 hours0 days 0 hours 0 weeks 0 months for l-required surveillance without placing the trip system in the tripped condition _provided at least one OPERABLE channel in the same_ trip system is monitoring that' parameter..

~(b)

This. function shall'be automatically bypassed when the reactor mode switch is'in the Run position and the associated APRM is not downscale.

(c)

The " shorting links" shall be removed from the RPS circuitry prior to and

during the time any control rod is withdrawn

and shutdown margin demonstrations performed per Specification _3.10.3.

(d)

The noncoincident NMS reactor trip function logic is such-that all channels go to both trip systems. Therefore, when the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMs, 6 IRMs and 2 SRMs.

(e)

.An APRM' channel is inoperable if there are less than 2 LPRM inputt per-level or-less than 14 LPRM inputs to an APRM channel.

(f)

This function is not required-to be OPERABLE when the reactor pressure vessel-head is removed per Specification 3.10.1.

.(g)

This function shall be' automatically bypassed when the reactor mode switch

.is not in the Run position.

(h)

This function is-not required to be OPERABLE when PRIMARY CONTAINMENT

-INTEGRITY is not required.

_(i)

With:any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j)

This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.

(k)'

Also actuates the EOC-RPT system.

(1)-

Required-to be OPERABLE only prior to and during s.utdown margin demonstrations as performed per Specification 3.10.3.

Not-required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

LIMERICK-- UNIT 1 3/4 3-5 Amendment No. JI,53

TABLE 3.3.1-2 r-

$f REACTOR PROTECTION SYSTEM RESPONSE TIMES E.U.

52 RESPONSE TIME FUNCTIONAL UNIT (Seconds)

E 1.

Intermediate Ranga Monitors:

-4 a.

Neutron Flux - High N.A.

w b.

Inoperative N.A.

2.

Average Power Range Monitor *:

a.

Neutron Flux - Upscale, Setdodi N.A.

b.

Neutron Flux - Upscale

1) Flow Biased 10.09

2) High Flow Clamped 10.09 jy c.

Inoperative N. A.

ts d.

Downscale N.A.

NN 3.

Reactor Vessel Steam Dome Pressure - High 5 0.55 4.

Reactor Vessel Water Level - Low, Level 3

$ 1.05 5.

Main Steam Line Isolation Valve - Closure 5 0.06 6.

Main Steam Line Radiation - High N.A.

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 5 0.06 10.

Turbine Control Valve Fast Closure, Trip Dil Pressure - Low

$ 0.08**

11.

Reactor Mode Switch Shutdown Position N.A.

12. Manual Scram M.A.

Neutron detectors are exempt from response time testing.

Response time shall be ceasured from the detector output or from the input of the first electronic component in the channel.

- Measured from start of turbine control valve fast closure.

I

TABLE 4e3.1.1-1 c-g:

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS x

CHANNEL OPERATIONAL M

CHANNEL FUNCTIONAL CHANNEL CONDITIONS;FOR.WHICH-f' FUNCTIONAL UZIT CHECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED' I

E 1.

Intermediate Range Monitors:

4 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 (0 :

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

1) Flow' Biased-S.D(g)

S/U(c),Q W(d)(e),SA 1

l j

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) l l ]'

R 1

d.

Downscale S

Q SA 1

y 3.

Reactor Vessel Stens Dome Pressure - High S

Q R

1,2(h) 4.

Reactor Vessel Water Level -

Low,' Level 3 S

Q R

1, 2 l

2-l 5.

Main Steam Line Isolation o-

' Valve - Closure N.A.

Q R

1 l'

d 5

6.

Main Steam Line Radiation -

g High S

Q R

1,2(h)

I.

7.

Drywell Pressure - High S

Q R

1, 2 l

p 8.

Scram Discharge Volume Water Level - High 1, 2, SI}

1,2,5(l) a.

Level Transmitter S

Q R

l b.

Float Switch N.A.

Q R

9 v

TABLE 4.3.1.1-1 (Continued) r.

k REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

=

E CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDIlIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

!E 9.

Turbine Stop Valve - Closure N.A.

Q R

1 l,

10. Turbine Control Valve Fast Closure. Trip Oil Pressure - Low N.A.

Q R

1 l

11.

Reactor Mode Switch Shutdown Position N.A.

R N.A.

1,2,3,4,5

12. Manual Scram N.A.

W H.A.

1.2,3,4,5 l

(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 u3 decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

v (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25'i of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Specification 3.2.2 Gall not be included in determining the absolute difference.

(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.

k (9) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing loop flow (APRM % flow). During the startup test program, data shall be recorded for the parameters listed to g

provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with g

the criteria listed shall commence upon the conclusion of the startup test program.

g (h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

f

((i)

With any control rod withdrawn. Not applicable to control rods removed per Spe

ication 3.9.10.1 or 3.9.10.2.

j)

If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 my hours fnr required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.

,p (k) Required to be OPERABLE only prior to and during shutdown margin cemanstrations as performed per Specification 3.10.3.

y e

'3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.

APPLICABILITY: As shown in Table 3.3.2-1.

ACTION:

a.

With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system:

1.

If placing the inoperable channels (s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to OPERABLE status witt.in:

a) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for trip functions not common

to the Reactor Protection System (RPS) and/or Emergercy Core Cooling System (ECCS) Actuation Instrumentation, or b) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for trip functions comon* to RPS and/or ECCS Actuation Instrumentation.

If this cannot be accomplished, the ACTION required by Table 3.3.2-1 for the affected trip function shall be taken, or the channel shall be placed in the tripped condition.

2.

If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed in the tripped condition within:

a)

I hour for trip functions not comon* to the RPS and/or ECCS Actuation Instrumentation, b) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common

to RPS Instrumentation, c) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions comon* to ECCS Actuation Instrumentation, and d) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions comon* to RPS and ECCS Actuation Instrumentation.

The provisions of Specification 3.0.4 are not applicable.

I o Trip functions comon to RPS and/or ECCS Actuation Instrumentation are shown i

in Table 4.3.2.1-1.

LIMERICK - UNIT 1 3/4 3-9 Amendment No. 53

'fNSTRUMENTATION LIMITING CONDITION FOR OPERATION _(Continued)

ACTION:.-(Continued)

With the number of OPERABLE channels less thdn required by the Minimum c.

0PERABLE Channels per Trip System requirement for both trip systems, T

~

place at least one trip system ** in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and take the ACTION required by Table 3.3.2-1, t

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated

-OPERABLE by the performance of the CHANNEL CHECK. CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations.for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulatad automatic operation of all channels-shall be performed at least once per 18 months.

4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each' isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at 1rast once per 18-months.

Each test shall include at least one channel per trip system such that all channele are tested at'least once every N times 18 months, where N is the total nur.l.

~ redundant channels in a specific isolation trip system.

4 a

- .The-trip system need not be placed in the' tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in-the tripped coadition.

LIMLAICK - UNIT 1 3/4 3-10 Amendmenc No. 53 i

1 TABLE 3.3.2-1 (Continued)

E 1SOLAIK44 A'.'TtiETT5N INSTRUMENTATION 9

M MINI ^rRM APPli:ABLE i

150LATIgI*IC) OPERABLE CHANNELS ) OPERATIONAL

$1GNAL PER TRIP SYSTEM tb TRIP FUNCTION COfWITION ACTION E

3 r

M 7.

SECONDARY CONTAIIstMT ISOLATION l

y

- m.

Reacter'.*r'tel Water Level i

Lou, Low - Level 2 8

2 1,2,3 25 I

b.

Dry s11 Pressure - High H

2 i, 2. 3 25 c.1.ReboelingAreaUnit1 Ventilation Exhaust Suct Radiation - High R

2 25

[

2. Refuel'ing Area Unit 2 Venti?ation w

25 l

I 2

Er%nust Duct Radiation - High R

2 d.

4*wter Enclosure Ventilatico Es3aust s ut Radiation - High 5

2 1,2,3 25 i

e.

Outside Atmosphere to Reactor Enclosure a Pressure - Low t'

1 1,2,3 25 f.

Outside Atmosphere To Refueling Area A Pressure - Lou T

1 25 i

g.

Reactor Enclosure Manual Initiation NA 1

1,2,3 24 f

h.

Refueling Area Manual Initiation MA 1

25 r

ma ca

?

18

=

~m-

iSOLAi b ACTUAi1W ANSWOMtn

..u.

ACTION STATEMENTS ACTION 20 - Be in at least HOT SHUiDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 21 - Be in at least STARTUP with the associated isolaticn valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUT 00WN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 22 - Be in at least cTARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 23 - In OPER* %

'91T10N 1 or 2. verify the affected system isolation valves are clos, Ith:

. hour and declare the affected system inoperable, in OPERA 110N, h #10N 3 be in at least COLD SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTICN 24 - Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at lear.t HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWd within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within I hour.

ACTION 26 - Close the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

TABLE NOTATIONS 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, May be bypassed under administrative control, with all turbine stop valves closed.

During operation of the associated Unit 1 or Unit 2 ventilstion exhaust system.

(a)

See Specification 3.6.3, Table 3.6.3-1 for primary containment isolation valves which are actuated by these isolation signals.

(b) A channel may be placed in an inoperable status for up to:

a) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for trip functions not common the the Reactor Protection System (RPS) and/or Emergency Core Cooling System (ECCS) Actuation Instrumentation, er b) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for trip functions common to RPS and/or ECCS Actuation Instrumentation for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter. Trip functions common to RPS and/or ECCS Actuation Instrumentation are shown in Table 4.S.2.1-1.

In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, M oard or cutboard, as applicable, in each line is OPERABLE and all required a.tvat4n instrumentation for that valve is OPERABLE, one channel may be placed in an inoprre.nle status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for required surveillance without placing the channel or trip system in the tripped condition.

LIHERICK - UNIT 1 3/4 3-16 Amendment No. 23. H, 53

S TABLE 3.3.2-1 (Continued)

TABLE NOTATIONS (c) Actuateslecondary containment isolation valves shown in Table 3.6.5.2.1-1 and/or 3.6.5.2.2-1 and signals B. H. S. U, R and T also start the standby gas treatment system.

(d)

RWCU system inlet outboard isolation valve closes on SLCS "B" initiation.

RWCU system inlet inboard isolation valve closes on SLCS "A" or SLCS "C" initiation.

(e) Manual initiat1on isolates the steam supply line outboard isolation valve and only following manual or automatic initiation of the system.

(f)

In the event of a loss of ventilation the temperature - high setpoint may be 0

raised by 50 F for a period not to exceed 30 minutes to permit restoration of the ventilation flow without a spurious trip.

During the 30 minute period, an operator, or other qualified member of tha technical staff, shall observe the temperature indications continuously, so that, in the event of rapid increases in temperature, the main steam lines shall be manually isolated.

(g) Wide rang'e accident monitor per Specification 3.3.7.5.

LIHERICK - UNIT 1 3/4 3-17 Amendment No. 28, 53

TABLE 3.3.2-2

_ ISOLATION ACTUATION INSTRUE NTATION SETPOINTS M

n ALLOWA8LE TRIP FUNCTION TRIP SETPOINT VALUE

[

1.

MAIN STEAM LINE ISOLATION i

a.

Reactor Vessel Water Level

1) Low, Low - Level 2

> - 38 inches *

> - 45 inches

2) Low, Low, Low - Level 1

[-129 inches *

[-136 inches b.

hin Steam Line

-< 3.0 x Full Power

-< 3.6 x Full Power Radiation - High

Background

Background c.

Main Steam Line

{

Pressure - Low 1 756 psig 1 736 psig Y

d.

h in Steam Line E$

Flow - High i 108.7 psid i 111.7 psid e.

Condenser Vacuum - Low 10.5 psia 1 10.1 psia /$ 10.9 psia f.

Outboard MSIV Room l

j Temperature - High

$ 192*F

$ 200*F g

g.

Turbine Enclosure - h in Steam i

g, Line Tunnel Temperature - High 1 165'F

$ 175'F

e al h.

h aual' Initiation N.A.

M.A.

w o

- z 2.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION i

8P e

~

a.

Reactor Vessel Water Level Low - Level 3 1 12.5 inches

1 11.0 inches

{

b.

Reactor Vessel (RHR Cut-in j

Permissive) Pressure - High i 75 psig 1 95 psig c.

Manual Initiation N.A.

N.A.

[

l l

' TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS e

y 9

CHANNEL GPERATION%L 1

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHIC!

]

]-

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIR) i j;

1.

' MAIN STEAM LINE ISOLATION U

I

1. ~

a.

Reactor Vessel Water Level

(

j.

L 1)' Low, Low, Level 2 S

Q R,

1, 2, 3 1

i l

2) Low, Low, Low - Level 1 5

Q R

1, 2, 3 i

i b.

MainSteagne

'l 2

j Radiation

--High 5

Q.

R 1, 2, 3 i

l

.c.

Main Steam Line-I Pressure - Low S

M R

1 l

l d.

Main Steam Line i

flow - High 5

M-R 1, 2, 3 -

1 t

i' w

e.

Condenser Vacuum - Low S

M R

1, 2**, 3**

3 t

i f.

Outboard MSIV Room l

a l

Temperature - High 5

M' R

1, 2, 3 i

g.

Turbine-Enclosure - Main Steam I

Line Tunnel Temperature - High 5

M R

1, 2, 3 i

h.

Manual Initiation N.A.

R N.A.

1, 2, 3 j

2.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION i

l Reactor Vessel Water Level a.

y Low - Level 3 S

Q R

1, 2, 3

{

o i

+

b.

ReactorVessel(RHRgt-In S

Q R

1, 2, 3 E

Permissive) Pressure. - High

(

l

[

c.

Manual Initiation-N.A.

R N.A.

1, 2, 3 l

a l

s IL

TABLE 4.3.2.1-1 (Continued)-

IS0tATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS h

2 5

CHANNEL OPERATIONAL Q

CHANCEL FUNCTIONAL CHANNEL CONDITIONS FOR lAtICH 4

1 RIP FUNCTION CHECK TEST CALIBRATION SUJRittANCE RfQUIRED i

C

' y 3.

REACTOR WATER CLEAMUP SYSTEM IS0!ATION l

1, 2, 3 a.

RWCS A Flow - High S.

M R

b.

RWCS Area Temperature - High S

M R

1, 2, 3 I

c.

RWCS Area Ventilation A Temperature - High 5

M R

1, 2, 3 d.

SLCS Initiation M.A.

R N.A.

1, 2, 3 e.

Reactor Vessel Water Leve1 Low, Low. - Level 2 5

Q R

1, 2, 3

\\

f.

Manual Initiation M.A.

R N.A.

1, 2, 3 i

I 4.

HIGH PRESSURE C00 TANT INJECTION SYSTEM ISOLATION l

a.

HPCI Steam Line i

t' A Pressure - High 5

M R

1, 2, 3 s

Y b.

HPCI Steam Supply g

Pressure - Low S

M R

1, 2, 3 c.

HPCI Turbine Exhaust Olaphrage Pressure - High 5

M R

1, 2, 3

{

d.

HPCI Equipment Room Temperature - High S

M R

1, 2, 3 g

R i

s e.

HPCI Equipment Room

[

A Temperature - High S

M R

1, 2, 3 o

f.

HPCI Pipe Routing Area U

Temperature - High S

M R

1, 2, 3 g.

Manual Initiation N.A.

R N.A.

1, 2, 3 i

h.

HPCI Steam Line A Pressure Timer N. A. -

M R

1, 2, 3

j TABLE 4.3.2.1-1 (Continued)

CM ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 59 CHANNEL OPERATIONAL 4

CHANNEL e

TRIP FUNCTION CHECK '

FUNCTIONAL CHANNEL CONDITIONS FOR 14tICH i

TEST CALIBRATION SURVEILLANCE REQUIRE 0 l

5

[

S.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.

RCIC Steam Line A Pressure - High S

M R

1, 2, 3 i

j b.

RCIC Steam Supply l

Pressure - Low S

M R

1,2,3 c.

RCIC Turbine Exhaust Diaphragm Pressure - High S

M R

1,2,3 f

R d.

RCIC Equipment Room l

T Temperature - High 5

M R

1, 2, 3 I

U j

e.

RCIC Equipment Room l

A Temperature - High S

M R

1, 2, 3 i

f.

RCIC Pipe Routing Area j.

Temperature - High 5

M R

1, 2, 3 f

g.

Manusi Initiation N.A.

R N.A.

1,2,3 i

i h.

RCIC Steam Line l

a Pressure Timer N.A.

M R

1, 2, 3 l

oR I

QE u 3 a

o k.5 t

i

]

TABLE 4.3.2.1-1,(Continued)

ISOLATION ACTUATION INSTRlMENTATION SURVEILLANCE REQUIREMENTS c-m$

CHANNEL-OP[ RAT 10NAL

-Q CHANNEL FUNCTIONAL CilANNEL CONDill0NS FOR WHICH TRIP FUNCTION CHECK TEST CAllBRATION SURVEILLANCE REQUla:

e E

6.

PRIMARY CONTAINMENT ISOLATION

I a.

Reactor Vessel Water Leve1 1)

Low,' Low - Level 2 5

Q R

1, 2, 3

2) Low, Low, Low - Level 1 5

Q R

1, 2, 3

'Drywell Pressure ### - High S

Q R

1, 2, 3 l

b.

c.

North Stack Effluent Radiation - High S

Q R

1, 2, 3 d.

Deleted v2 e.

Reactor Enclosure Ventilation w

Exhaust Duct - Radiation - High 5

M R

1, 2, 3 o

f.

Outside Atmosphere to Reactor Enclosure A Pressure - Low-N.A.

M Q

1, 2, 3 g.

Deleted h.

Drywell Pressure

- High/

Reactor Pressure

- Low S

Q R

1, 2, 3 2

1.

Primary Containment Instrument R

Gas to Drywell A Pressure - Low N.A.

M Q

1, 2, 3 8

5 j.

Manual Initiation N.A.

R N.A.

1, 2, 3 y

G

I

-1ABLE'4.3.2.1-1'(Continued) t.

. ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

e.

CHANNEL

. CONDITIONS FOR WHICH r

OPERATIONAL

{

~

CHANNEL.

FUNCTIONAL CHANNEL I

h CHECK TEST CALIBRATION SURVEILLANCE REQUIRED TRIP FUNCTION I

M 7.

SECONDARY CONTAIMMENT ISOLATION ##

a.

Reactor Vessel Water Leve1 y

Low, Low - Level 2-S Q

R 1,2,3 b.

Drywell Pressure ### - High 5

Q R

1, 2, 3

c. 1. Refueling Area Unit 1 Ventilation j

Exhaust Duct Radiation - High S

M R

4 g

2. Refueling Area' Unit 2 Ventilation-I j

Exhaust Duct Radiation -;High

.S M

R

I l

4 r

I d.

Reactor Enclosure Ventilation i

j Exhaust Duct Radiation - High S

M R

1, 2, 3 p

l' e.

Outside Atmosphere To Reactor j

g Enclosure A Pressure - Low N.A.

M Q

1, 2, 3 y

l Y

f.

Outside Atmosphere To Refueling M

Area A Pressure - Low N.A.

M Q

3 i

i g.

Reactor Enclosure j

Manual Initiation N.A.

R N.A.

1, 2, 3 i

L h.

Refueling Area

[

Manual Initiation M.A.

R N.A.

o

Required when (1) handling irradiated fuel in the refueling area secondary containment, or (2) during CORE g

ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel

}

head removed and fuel in the vessel.

f m**

i j

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

y

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

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

[

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

u l

i j.

fffThis trip function (Ib) is common to the RPS trip function.

i t

j I

1.

i m

, +

i

!*SteuwfNtar!cN 3/a 3 3 EutRGENCY CCRE C00 LING SYST!W aCTUAf!CN IN5teUw(N?Af!ON

AY

\\

L w:*:NG CON 0! TION FOR OD(Sat;ON l

2.3 3 fee e+e ;enty : ore cooling system (ECCS) actsation instrumentation esa-aeis sno.e in Tacle 3. 3.31 small ee OPERABLE ita tnote trio setcoinst set ::as t steat its the va16es sno.a in tes Tric Setootnt column of Tao?e 1.3 J 2 anc =stn EMER3ENCr CCRE COOLING SYSTEN RESPONSE TIME as sn A9o(!CABILITY:

As snown in Table 3.3.3+1.

AC*10N; Witn an ECCS actuation instrumentation channel trio setootnt less 4.

conservative than the value shown in the Allowacle Values column of Table 3.3.3-2. ceclare the channel incoerable until the channel is restorea to OPERA 8LE status with its trip setooint aajustec consistees witn the-Trip 5etcoint value.

With one or more ECCS actuation instrumentation channels inoperstle, D.

take the ACTION reQuirec Dy Table 3.3.3 1.

With either A05 trip system suosystem inoperacle, restore the c.

inoperaole trip system to CPERA8LE status within:

1.

7 days, p oviced that the HPCI and RCIC systems are OPERA 8LE.

2.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Otr.orwise, be in at least NOT SHUTDOWN within the nest 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam come pressure to less than or equal to 100 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVE!L'ANCC RE00!REWENTS 4.3.3.1 Each ECC5 actuation instrumentation channel shall be demonstrated CPERA8LE by the performance of the CHANNEL CNECK, CHANNEL FUNCTIONAL TEST and CHANNE. 4Al.! BRAT 10N operations for the OPERATIONAL CON 0! TION 5 and at the frequencies shoun in Tatie 4.3.3.1 1.

'4. 3. 3. 2 f.0GIC WSTEM FUNCTIONAL TEST 5 and simulated automatic operation of all channels shall be perfereed at least once per 18 months.

4.3.3.3 The ECCS RE5o0NSE TIME of each ECCS trip function shown in Table 3.3.3 3 shall bei demonstrated to be within the Ifnit et least once per la months. Each test shall include at least one channel per trip systes such that all channels are tested at least once every N times 18 months where N is the total number of reovncant channels in a specific ECC5 trip systas.

LIMERICK - UNIT 1 3/4 3 32

} } -} y-[ ;t e..: c'g s-':q.* M f :.

c.~

~.

.j <

< v. n

meLE 3.3.3-1 (Continued)

FMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE NOTATIONS

{

(a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for l

required surveillance without p'. acing the trip system in the tripped conditjon provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) Also provides input to actuttien logic for the associated emergency diesel generators.

(c) One trip system.

Provides signal to HPCI pump suction valves only.

(d) On 1 out of 2 taken twice logic, provides a signal to trip the HPCI pump turbine only.

(e)

The manual initiation push buttons start the respective core spray pump and diesel generator. The "A"

and "B" logic manual push buttons also actuate an initiation permissive in the injection valve opening logic.

(f) A channel as used here is defined as the 127 bus relay for Item 1 and the 127, 127Y, and 127Z feeder relays with their associated time delay relays taken together for item 2.

When the system is required to be OPERABLE per Specification 3.5.2.

I Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

Required when ESF equipment is required to be OPERABLE.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.

LIMERICK - UNIT 1 3/4 3-35 Amendment No. 53

t 7ABLE 3.3.3-1 U.ontinued)

EMERGENCY CORE COOLING SY5 TEM ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 30 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

'a.~

'i. one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the associated l

system inoperable.

b.

With more than one channel inoperable, declare the associated system inoperable.

ACTION 31 -

With the number of OPE *ABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ECCS inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ACTION 32 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the.

inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ACTION 33 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the l

associated ECCS inoperabla.

ACTION 34 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a.

For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the HPCI system j

inoperable.

b.

With more than one channel inoperable, declare the HPCI system inoperable.

ACTION 35 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable chan.9el in the tripped condition within 24 l

hours or declare the HPCI system inoperable.

ACTION 36 -

With the number of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator inoperable and take the ACTION required by 5pecification 3.8.1.1 or 3.8.1.2, as appropriate.

ACTION 37 -

With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; operation may then continue until performance of the next required CHANNEL FUNCTIONAL TEST.

LIMERICK - UNIT 1 3/4 3-36 Amendment No. !!, 53

TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ECCS RESPONSE TIME (Seconds) 1.

CORE SPRAY SYSTEM

$ 27 2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

$ 40 3.

AUTOMATIC DEPRESSURIZATION SYSTEM N.A.

4.

HIGH PRESSURE COOLANT INJECTION SYSTEM

$ 30 5.

LOSS OF POWER N.A.

l i

l LIMERICK - UNIT 1 3/4 3-39

TABLE 4.'3.3.1-1 e-EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERAT10NAL-ft CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH

[

TRIP FUNCTION CHECK TEST

-CALIBRATION SURVEILLANCE REQUIRED l

s 4

1.

CORE SPRAY SYSTEM i

a.

Re.,ctor Vessel Water Level -

t Low Low Low, Level 1 S

Q R

1, 2, 3, 4*, 5*

[

b.

Drywell Pressure - High 5

Q R

1, 2, 3 c.

Reactor Vessel Pressure - Low 5

Q R

1, 2, 3, 4*, 5*

d.

Manual. Initiation N. A..

R N.A.

1, 2, 3, 4*, 5*

i 2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM a.

Reactor Vessel Water Level -

I Low Low Low, Level 1 5

.Q R

1, 2, 3, 4*, 5*

I b.

Drywell Pressure'.- High S

Q R

1, 2, 3 U

c.

Reactor Vessel Pressure - Low S

Q R

1, 2, 3 l

i d.

Injection Valve Differential Pressure Low (Permissive)

S Q

R 1, 2, 3, 4*, 5*

I

[

i Y

I' e.

Manual Initiation N.A.

R N.A.

1, 2, 3, 4*, 5*

l

.i 3.

HIGH PRESSURE COOLANT INJECTION SYSTEM ***

Reactor Vessel Water Level -

a.

Low Low Level 2 5

Q R

1, 2, 3 l

l b.

Drywell Pressure - High 5

Q R

1, 2, 3 l

c.

Condensate Storage Tank Level -

S Q

R 1, 2, 3 l

Low d.-

Suppression. Pool Water Level -

.High S

Q R

1, 2, 3 I

i 7

e.

Reactor Vessel Water Level -

I i

High, Level 8~

S.

Q R

1, 2, 3 f.

Manual Initiation N.A.

R H.A.

1, 2, 3 I

I w

i s

L i

TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS e.

k,,

CHANNEL OPERATIONAL F;

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED I

4.

AUTOMATIC DEPRESSURIZATION SYSTEM U

a.

Reactor Vessel Water Level -

Low Low Low, level 1 5

Q R

1, 2, 3 b.

Drywell Pressure - High S

Q R

1, 2, 3 c.

ADS Timer N.A.

Q Q

1, 2, 3 d.

Core Spray Pump Discharge Pressure

.High S

Q R

1, 2, 3 I*

e.

RHR LPCI Hode Pump Discharge Pressure - High 5

Q R

1, 2, 3 f.

Reactor Vessel Water Level - Low, Level 3 S

Q R

1, 2, 3 g.

Manual Initiation N.A.

R N.A.

1, 2, 3 h.

ADS Drywell Pressure Bypass Timer N.A.

Q Q

1, 2, 3 l

t'.

5.

LOSE OF POWER z-4.16 kV Emergency Cas Undery, Y

a.

O vc:tage (Loss of Voltage)

N.A.

R N.A.

1, 2, 3, 4**, 5**

b.

4.16 kV Emergency Bus Under-I voltage (Degraded Voltage)

S M

R 1, 2, 3, 4**, 5**

. o j{

When the system is required to be OPERABLE per Specification 3.5.2.

Required OPERABLE when ESF equipment is required to be OPERABLE.

5

,!?

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

tg

1. Loss of Voltage Relay 127-IlX is not field setable.

e

INSTRUMENTATION

-3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.1 The anticipated transient wittnut scram recirculation pump trip (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall be OPERABLE with their trip setpoints set consistent with values shown in the Trip Setpoint column of Table 3.3.4.1-2.

APPLICABILITY:

OPERATIONAL CONDIf!ON 1.

ACTION:

a.

With an ATWS recirculation pump trip system instrumentatte., channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel trip setpoint adjusted consistent with the Trip Setpoint value, b.

With tha number of OPERABLF. channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within I hour, c.

With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:

1.

If the inoperable channels consist of one reactor vessel water level channel and one reactor vessel pressure channel, place both inoperable channels in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or, if this action will initiate a pump trip, declare the trip system inoperable.

2.

If the inoperable channels include two reactor vessel water level channels or two reactor vessel pressure channels, declare the trip system inoperable.

d.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months nr be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e.

With both trip systems inoperable, restore at least one trip system to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.4.1.1.

Each ATWS recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.1-1.

4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

LIMERICK - UNIT 1 3/4 3-42

~ <.-

... a. s REACTOR CORF ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTA710N MINIMUM OPERABLE CHANNELS FUNCTIONAL UNITS PER TRIP FUNCTION

ACTION o.

Reactoc. Vessel Water Level -

Low Low. Level 2 4#

50 b.

Reactor Vessel Water Level -

High, level 8 4#

51 c.

Condensate Storage Tank Water Level - Low 2**

52 d.

Manual Initiation 1/ system ***

53

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l

for required surveillance without placing the trip system in the tripped condition provided all other channels monitoring that pera.neter are OPERABLE.

- 0ne trip system with one-out-of-two logic.

- - 0ne trip system with one channel.

0ne trip system with one-out-of-two twice logic.

LIMERICK - UNIT 1 3/4 3-53 Amendment No. 53

TABLE 3.3.5-1 (Continued)

REACTOR CORE ISOLA?!ON COOLING SYSTEM ACTION STATEATNTS ACTION 50 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

3."

With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the RCIC l

system inoperable.

b.

With more than one channel inoperable, declare the RCIC system inoperable.

ACTION 51 -

With the number of OPERABLE channels less than required by the minimum OPERABLE channels per Trip Systei requirement, declare the RCIC system inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ACTION 52 -

With the number of OPERACLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, place at leest one inoperable channel in the tripped cone'ition within 24 l

hours ur declare the RCIC system inoperable.

ACTION 53 -

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channel to OPERABLE statur within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the l

RCIC system inoperable.

LIMERICK - UNIT 1 3/4 3-54 Amendment No. 53

l t

TABLE 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOf 1

ALLOWABLE FUNCTIONAL UNITS TRIP SETPOINT VALUE-a.

Reactor Vessel Water Level -

I Low Low, Level 2

>-38 inches *

>-45 inches b.

Reactor Vessel Water Level -

High, Level 8 1 54 inches 1 60 inches _

l c.

Condensate Storage Tank Level -

Low

> 135.8** inches

> 132.3 inches d.

Manual Initiation N.A.

N. A.

3 L

"See Bases Figure-B 3/4.3-1.

- Corresponds to 2.3 feet indicated.

?

l.IMERICK - UNIT 1 3/4 3-55 Amendment No. 33 00T 3 0 889

ixott e.a.5.1-1

+

REACTOR CORE ISOLATION SYSVEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL FUNCTIONAL UNITS CHECK TEST CAllBRATION a.

Reactor Vessel Water level -

Low Lod, level 2 S

Q R

\\

b.

Reactor Vessel Water Level -

High, level B S

Q R

l c.

Condensate Storage Tank Level - Low 5

Q R

g d.

Manual Initiation N.A.

R N.A.

1 LIHERICK - UNIT 1 3/4 3-56 Amendment No. 53 i

I

TABLE 4.3.6-1 7

i Q

CONTROL' ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS j$

CiANNEL OPERATIONAL g

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH Q

TRIP FUNCTION CHECK-TEST CALIBRATION (a)

SURVEILLANCE REQUIRED h

1.

ROD BLOCK MONITOR z

5 S/U(D)(C)

I (C) a.

Upsc.nle N.A.

SA 1*

S/U b)(c),(c)

I b.

Inoperative N.A.

~

H.A.

1*

l c.

Downscale N.A.

S/U(b)(c),(c)

SA 1*

j.=

2.

APRM a.

Flow Biased Neutron Flux -

S/U((b),0 b) Q SA 1

I Upscale N.A.

b.

Incperative N.A.

S/U(b),

N.A.

1' 2* 5***

i i

c.

Downsca!e N.A.

S/U SA l

S/U(b),Q d.

Neutron Flux - Upscale, Startup N.A.

,Q SA 2, 5***

3.

SOURCE RANGE MONITORS S/U((b)*W b)

R a.

Detector not full in N.A.

NA 2, 5 S/U y

gj

2, 5

.b.

Upscale N.A.

Y c.

Inoperative N.A.

S/U(b)*g d.

Downscale N.A.

S/U(b) y Sj-4.

INTERMEDIATE RANGE MONITORS y

S/U(b)

N.A.

2, 5 a.

Detector not full 1i N.A.

S/U(b),W l

b.

Upscale N.A.

S/U(b),W SA 2, 5 c.

Inoperative N.A.

S/U(b),W N.A.

2, 5 i

d.

Downscale N.A.

,W SA 2, 5 i

l

[

5.

SCRAM DISCHARGE VOLUME i

a.

Water Level-High N.A.

Q R

1, 2. 5**

l

[

6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW 4

O f

a.

Upscale N.A.

S/U(b) Q SA 1

l

~

S/U(b),

N.A.

1 E

b.

Inoperative N.A.

S/U(b),Q

~

I c.

Comparator N.A.

,Q SA-1 w

l u

7.

REACTOR MODE SWITCH SHUTDOWN POSITION N.A.

R N.A.

3, 4 i

~.

TABLE 4.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

(c) Includes reactor manual control multiplexing system input.

With THERMAL POWER > 30% of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods, removed per Specification 3.9.10.1 or 3.9.10.2.

Reaufaed to be OPERABLE only prior to and during shutdown margin der -

7tions as performed per Specification 3.10.3.

I l

l t.

LIMERICK - UNIT 1 3/4 3-62 Amendment N 41

3/4,3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION-SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel. cladding, b.

Preserve-the integrity of the reactor coolant system.

c.

Minimize _the. energy which must be absorbed following a 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 perfort 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 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.

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 t.nalyses. No credit was takea 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 (2) utilizing replacement sensors with certified response times.

L LIMERICK - UNIT-1 B 3/4 3-1 Amendment No. 53

w.. a u wn. w wu 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 iesponse 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).

Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial offect on safety. The setpoints of other instrumentation, where only the high or low end of tb setting have a direct bearing on safety, are established at a level away from the iormal 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 serind delay is as' sed before the valve starts to move.

For A.C. operated valves.

t 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 valve is assumed; thus the signal delay (sensor response) is concurrent with the 10-second diesel startup and the 3 second load center loa;ing delay.

The saft:ty analysis ccesiders 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.

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 Valuc 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 l

initiate actions to mitigate the consequences of accidents that are beyond the l

ability of the operator to control. This specification provides the OPERADILITY j

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.

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Speicification Improvement Methodology (with Demonstration for BWR ECCS l

LIMERICK - UNIT 1 B 3/4 3-2 Amendment No. 33. 53 l

l

th5 a g.a 10N

.SASES I

3/4.3.3 EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION (Continued)

Actuation instrumentation)f as approved by the NRC and documented in the SER (letter to D. N. Grace from A. C. Thadant dated December 9, 1988 (Part 1) and letter to D. N. Grace from C. E. Rossi dated December 3, 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 analyses.

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient.

The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NEDO-10349, dated March 1971, NED0-24222, dated December 1979, and Section 15.8 of the FSAR.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a supplement to the reactor trip.

During turbine trip and generator load rejection events, the EOC-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 EOC-RPT protective feature till function are closure of the turbine stop valves and fast closure of the turbine control valves.

A f ast closure sensor from each of two L:rbinS control valves provides input to the EOC-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 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 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 room.

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 eithin 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 l

analyses.

l LIMERICK UNIT 1 8 3/4 3-3 Amendment No. 53

(

INSTRUMENTATION BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. This instrumentation does not provide actuation of any of the emergency core cooling equipment.

Specified surveillance intervals and maintenance outage times rave been specified in accordance with recommendations made by GE in their letter to the BWR Owner's Group dated August 7, 1989,

SUBJECT:

" Clarification of Technical Specification changes given in ECCS Actuation Instrumentation Analysis."

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the oasis that the difference between each Trip Setpoint and the Allowable Value is an ellowance

-for instrument drift specifically allocated for each trip in the safety analyses.

3/4.3.6 CONTROL'R00 BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 1. " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation "

as approved by the NRC and documented in the SER (letter to D. N. Grace from C.

E. Rossi dated September 22,1988).

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.

3/4.3.7-MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that;

-(1)-the radiation levels are continually meamred in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation ~ level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an' accident. This capability is consistent with 10 CFR Part 50, Appendix-A, General Design Criteria 19, 41 -60, 61, 63, and 64.

LIMERICK - UNIT 1 B'3/4 3-4 Amendment No. 48,53

BASES l

3/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a i

seismic event and evaluate the response of those features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for the unit.

3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCH.

3/4.3.7.4 REMOTE SHUTOOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdawn and maintenance of HOT SHUTDOWN of the unit from locations outside of the contrei room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50, Appendix a.

3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.

This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"

December 1975 and NUREG-0737, " Clarification of TM1 Action Plan Requirements,"

November 1980.

3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with infarmation of the status of the neutron level in the core at very low power levels during startup and shutdown.

At these power levels, reactivity additions shall not be made without this flux level information available to the operator. When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.

3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obiained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor Core.

The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e.,

detectors) prior to performing an LPRM calibration function. Monitoring core thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE. The OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output in the resultant heat balance calculation (P-1) with data obtained during a previous heat balance calculation (P-1).

LIMERICK - UNIT 1 8 3/4 3-5 Amendment No. 48,53

e.

INSTRUMENTATION BASE 5 3/4.3.7.8 CHLORINE AND T0XIC GAS DETECTION SYSTEMS The OPERABILITY of the chlorine and toxic gas detection systems ensures that an accidental chlorine and/or toxic gas release will be detected promptly and the necessary protective actions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for control room personnel.

Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically be placed in the chlorine isolation mode of operation to provide the required protection.

Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of operation to provide the required protection.

The detection systems required by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95 " Pro-tection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975.

3/4.3.7.9 FIRE DETECTION INSTRUMENTATION

~

OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that fite suppression systems, that are actuated by fire detectors, will discharge extin-guishing agent in a timely manner.

Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.

Fire detectors that are used to actuate fire suppression systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification.

i Consequently, the minimum number of OPERABLE fire detectors must be greater.

l The loss of detection capability for fire suppression systems, actuated l

by fire detectors, represents a significant degradation of fire protection for l

any area.

As a result, the establishment of a fire watch patrol must be initi-ated at an earlier stage than would be warranted for the loss of detectcrs that provide only early fire warning.

The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.7.10 LOOSE-PART DETECTION SYSTEM l

The OPERASILITY of the loose part detection systen ensures that sufficient l

capability is available to detect loose metallic parts in the primary system I

and avoid or mitigate damage to primary system components.

The allowable out-l of-service times and surveillance requirements are consistent with the recoa-l mendations of Regulatory Guide 1.133, " Loose-Part Detection Program for tha l

Primary System of Light-Water-Cooled Reactors," May 1981.

3/4.3.7.11 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.

i 1

l LIMERICK - UNIT 1 B 3/4 3-6 Amendment No.48

% 48f/

-~

. - ~ _ -..

.~

c

~ UNITED STATES -

- [

v. ~ ' p, TJUCLEAR REGULATORY COMMISSION aE.

W ASHING TON, D. C, 2055$

'%, ;, g PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION, UNIT,2, AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 17 License No. NPF-85 1.

The Nuclear Regulatory Comission (the Comission) has found that A.

The application for amendment by Philadelphia Electric Company (the licensee) dated April 26, 1990, couplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and-the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules.and regulations of the Comission; C.

Thereisreasonableasseance(i)thattheactivitiesauthorizedby this amendment can be r :drL W without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

Thel issuance-of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The' issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable _ requirements have been ictisfied.

2.

Accordingly, the license is amended by changes to the Technical Specificatie.s as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-85 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.17., __are her<eby incorporated into this license. Philadelphia Electric Corpany shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

l 3.

THs license anendnent is effective fif teen (15) days af ter date of

issuance, FOR THE NUCLEAR REGULATORY COMMISSICH

$s

.:r, Dir,ecter Charles L.c rro;ect Directorate 1-2 Division of Reactor Prcjects - 1/11 t,t ta chttent :

Clarges to the Technical Specifications Date of Issuance: December 2,1991

4 8:

ATTACHMENT TO LICENSF AMENDMENT N0. 17 FACILITY OPERATING LICENSE NO. NPF-85 POCKET NO. 50-353 Replace the following pages of the Appendix A Technical Specificatiens with the attached pages. The revised pages are identified by Amendment nuriner and contain verticai lines indicating the area of change. Overleaf pages are provided to maintain' document completeness.*

Remove Insert xix xix xx xX*

3/4 3-1 3/4 3-1 3/4 3-2 3/4 3-2*

3/4 3-5 3/4 3-5 3/4 3-0 3/4 3-6*

3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10 3/4 3-15 3/4 3-15*

3/4 3-16 3/4 3-16 3/4 3-17 3/4 3-17 3/4 3-18 3/4 3-18*

3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 3-29 3/4 3-29*

3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32*

3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-39 3/4 3-39*

3/4 3-40 3/4 3-40 3/4 3-41 3/4 3-41 3/4 3-42 3/4 3-42*

- ATTACHMENT,T0__ LICENSE AMENDMENT NO.17 FACILITY OPERATING LICEN_SE_NO._NPF-85 DOCKET NO. 50-353 I

Remove Insert 3/4 3-53 3/4 3-53 3/4 3-54 3/4 3-54 3/4 3-55 3/4 3-55*

3/4 3-56 3/4 3-56 3/4 3-61 3/4 3-61 3/4 3-62 3/4 3-62*

B 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2 B 3/4 3-3 B 3/4 3-3 8 3/4 3-4 B 3/4 3-4 B 3/4 3-5 B 3/4 3-5 B 3/4 3-6 B 3/4 3-6*

.t.J

o l

BASES SECTION PAGE INSTRUMENTATION (Continued)

Seismic Monitoring Instrumentation..........................

B 3/4 3-5

( D e l e t e d ).................................................. B 3 / 4 3-5 l

Remote Shutdown System Instrumentation and Controls......... B 3/4 3-5 Accident Monitoring Instrumentation......................... B 3/4 3-5 Source Range Monitors....................................... B 1/4 3-5 g

Traversing In-Core Probe System............................. B 3/4 3-5 Chlorine and Toxic Gas Detection Systems.................... B 3/4 3-6 Fire Detection Instrumentation.............................. B 3/4 3-6 Loose-Part Detection System................................. B 3/4 3-6 (Deleted)..................................................B3/43-6 Offgas Monitoring Instrumentation...........................

B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM.........................

B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION............................................. B 3/4 3-7 Bases Figure B 3/4.3 1 Reactor Vessel Water Level........................... B 3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM........................................ B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES........................................ B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................... B 3/4 4-3 Operational Letkage......................................... B 3/4 4-3 3/4.4.4 CHEMISTRY................................................... B 3/4 4-3a LIHERICK - UNIT 2 xix Amenon ent No. II,12,17 1

\\

i

IEEX BASES SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY.......................................

B 3/4 4-4 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.............................

B 3/4 4-4 Eases Table B 3/4.4.6-1 Reactor Vessel Toughness.................

B 3/4 4-7 Bases Figure B 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) At 1/4 T As A Function of Service Life......................

B 3/4 4-8 3/4.4.7 MAIN STEAM LINE ISOLATION VAi.VES........................

B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY....................................

B 3/4 4-6 3/4.4.9 RESIDUAL HEAT RFM0 VAL...................................

B 3/4 4-6 3/4.5 -EMERGENCY CORE C0OLING SYSTEMS 3/4.5.1 cnd 3/4.5.2 ECCS - OPERATING and SHUTDOWN............

B 3/4 5-1 3/4.5.3 SUPPRESSION CHAMIER................................

B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAIMENT Primary Contairunent Integrity......................

B 3/4 6-1 Primary Containment Leakage........................

B 3/4 6-1 Primary Containment Air Lock.......................

B 3/4 6-1 MSIV Leakage Control Systes........................

B 3/4 6-1 Primary Containment Structural Integrity..........._

B 3/4 6-2 Drywell and Suppression Chamber Internal Pressure.........,...............................

B 3/4 6-2 Drywell Average Ai r Temperature....................

B 3/4 6-2 Drywell and Suppression Chamber Purge System.......

B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS...........................

B 3/4 6-3

' **DTCK - UNIT 2 xx

-3/5.3 INSTRUMt'NTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With the number of OPERABLE channels less than required by the Minimum a.

OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The provisions of Specification 3.0.4 are l

not applicable.

b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and take the ACTION required by Table 3.3.1-1.

SURVEILLANCE RrQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.

An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, the inoperable channel shall be restored to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or the ACTION required by l

Table 3.3.1-1 for that Trip Function shall be taken.

- The trip system need not he placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system r

in the tripped condition.

LIMERICK - UNIT 2 3/4 3-1 Amendment No. 17

TABLE 3.3.1-1 Ey REACTOR PROT 2CTION SYSTEM INSTRUMENTATION' M

n APPLICABLE MINIMUM E

OPERATIONAL'.

OPERABLE CHANNELS Z FUNCTIONAL UNIT-CONDITIONS

_PER TRIP SYSTEM (a)

ACTION IDI-1.

Intermediate Range Monitors a.

Neutron Flux - High 2

3 1

3, 4 3

2 5(c) 3(d) 3 b.

Inoperative 2

3 1

3, 4 3

2 5

3(d) 3 2.

Average Power Range Monitor '):

I Y

N a.

Neutron Flux - Upscale, Setdown 2

2 1

3 2

2 5(c)(1) 2(d) 3 l

b.

Neutren Flux - Upscale

1) Flow Blased 1

2 4

2) High Flow Clamped 1

2 4

<-[

c.

Inoperative 1, 2 2-I-

Eg 3

2 2

c.a g 5(c)(1) 2(d) 3 l

+

5 d.

Downscale 1(g) 2 4

l 5

3.

Reactor Vessel Steam Dome Pressure - High 1,2(f) 2 1

4.

Reactor Vessel Water Level - Low, Level 3 1, 2 2

1 5.

Main Steam Line Isolation Valve -

Closure 1(g) 1/ valve 4

e

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)

_A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for l

required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b)

This function shall be automatically bypassed when the reactor mode switch is in the Run position and the associated APRM is not downscale.

(c)

The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn

and shutdown margin demonstrations performed per Specification 3.10.3.

(d)

The noncoincident NHS reactor-trip function logic is such that all channels go to both trip systems.

Therefore, when the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMs, 6 IRMs and 2 SRMs.

(e)

An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.

(f)

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(g)

This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h)

This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(1)

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(j)

This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.

(k)

Also actuates the E0C-RPT system.

(1)

Required to be 0Pi~;ABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.

Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

LIMERICK - UNIT 2 3/4 3-5 Amendment No. 7,17 a

y TABLE 3.3.1-2 E

m5 REACTOR PROTECTION SYSTEM RESPONSE TIMES E

RESPONSE TIME FuhCT10NAL UNIT c

(Seconds)

[*--

1.

' Intermediate Range Monitors:

to a.

Neutron Flux - High

-N.A.

b.

Inoperative M.A.

2.

Average Pow:.r Range Monitor *:

a.

Neutron Flux - Upscale, Setdown N.A.

b.

Neutron Flux - Upscale 1)- Flow Biased

<0.09

2) High Flow Clamped

}0.09

{

c.

Inoperative M.A.

y d.

Downscale M.A.

3.

Reactor Vessel Steam Dome Pressure - High.

~< 0.55 4.

Reactor Vessel Water Level - Low, Level 3

~< 1.05 5.

Main Steam Line Isolation Valve - Closure

< 0.06 6.

Main Steam Line Radiation - High N.A.

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 St8p Valve - Closure

_ 0.06 10.

Turbine Control Valve Fast Closure, Trip 011 Pressure - Lcw

_ 0.08**

11.

Reactor Mode Switch Shutdown Position M.A.

12. Manual Scran N.A.

" Neutron detectors are exempt from response time testing.

Response time shall be measured from *he detector output or from the input of the first electronic component in the channel.

- Measured from start of turbine control valve fast closure.

O

TABLE 4.3.1.1-1 b

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS h

CHANNEL OPERATIONAL M

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH j

FUNCTIONAL UNIT CHECK TEST CALIBRAlION(a)

SURVEILLANCE REQUIRE i

l f

E 1.

Intermediate Range Monitors:

l Q

a.

Neutron Flux - High S/U.S(b)

S/U(c). W R

2 1

S W(j)

R 3,4,5 u

b.

Inoperative N.A.

W(j)

N.A.

2,3,4,5 Average Power Range MonitorII):

2.

a.

Neutron Flux -

S/U.S(b)

S/U(c). W SA 2

Upscale Setdown S

W(j)

SA 3,5(k) i b.

Neutron Flux - Upscale

1) Flow Biased 5,D(g)

S/U(c),Q W(d)(e),SA 1

l

2) High Flow Clamped S

S/U(c),Q W(d)(e),SA 1

l c.

Inoperative N. #..

Q(j)

N.A.

1,2,3,5(k) g w

l 5

d.

Downscale S

Q SA 1

l m

3.

Reactor Vessel Steam Dome Pressure - High S

Q R

1,2(h) l 4.

Reactor Vessel Water Level -

Low, Level 3 S

Q R

1, 2 l

S.

Main Steam Line Isolation Valve - Closure N.A.

Q R

1 R

E 6.

Main Steam Line Radiation -

2 High S

Q R

1,2(h) l 7.

Drywell Pressure - High 5

Q R

1, 2 l

g l

h 8.

Scram Discharge Volume Water 1,2,5(i)I)

Level - High C

a.

Level Transmitter S

Q R

1 2, 5(

b.

Float Switch N.A.

Q R

TABLE 4.3.1.1-1 (Continued) e k

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

E CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH

[

FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED z

4 9.

Turbine Stop Valve - Closure N.A.

Q R

1 l

w

10. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low N.A Q

R 1

l 11.

Reactor Mode Switch Shutdown Position N.A.

R N.A.

1,2,3,4,5

12. Manual Scram H.A.

W N.A.

1,, 2, 3, 4, 5 l

(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 u3 decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

u 5

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat bahnce during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER. Any'APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be included in determining the absolute difference.

(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the JIP system.

(g) Verify measured core flow (trtal core flow) to be greater than or equal to established core flov at the existing loop flow (APRM % flow). During the startup test program, data shall be recorded for the p>;6 meters listed to provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with the criteria listed shall commence upon the conclusion of the startup test program.

al (h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

g (i) With any control rod withdravn. Not applicable to control rods removed per Specification 3.9.10.I or 3.9.10.2.

(j)

If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 y

hours for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.

" (k) Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.

(

l

J/4.3.2 ISOLdTION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.

APPLICABILITY: As shown in Table 3.3.2-1.

ACTION:

a.

With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allorable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Seticint value, b.

With the number of OPERABLE channe's less than required by the Minimum GPERABLE Channels per Trip System requirement for one trip system:

1.

If placing the inoperable (.hannels(s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to OPERABLE status within:

a) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for trip functions not comon* to the Reactor Protection System (RPS) and/or Emergency Core Cooling System (ECCS) Actuation Instrumentation, or b) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for trip functions common

to RPS and/or ECCS Actuation Instrumentation.

If this cannot be accomplished, the ACTION required by Table 3.3.2-1 for the affected trip function shall be taken, or the channel shall be placed in the tripped condition.

2.

If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed ir the tripped condition within:

l a)

I hour for trip functions not comon* to the RPS and/or ECCS Actuation t

Instrumentation, b) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functiors common

to RPS Instrumentation, l

c) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions common

to ECCS Actuation Instrumentation, and d) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions comon* to RPS and ECCS Actuation Instrumentation.

The provisions of Specification 3.0.4 are not applicable, Trip functions common to RPS and/or ECCS Actuation Instrumentation are shown o

in Table 4.3.2.1-1.

l LIMERICK - UNIT 2 3/4 3-9 Amendment No. 17 l

-n

1, e

INSTRUMENTATION LIMITING CONDITION FOR OPERATION (Continued)

L, ACTION:

(Continued)

With.the number of OPERABLE channels less than required by the Minimum c.

OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ati take the ACTION required by Table 3.3.2-1.

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel snail b2 demon cratad OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST,'and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.2.3 The ISOLATION SYSTEM P.ESPONSE TIM' of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within-its limit at least once per 18 months.

Each test shall include at least one channel per trip system such that all channels-are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific isolation trip system.

The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur. When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.

LikiRICK - UNIT 2 3/4 3-10 Amendment No. 17

TABLE _3.3.2-1 (Continued)

ISOLATION ACTUATTUN INSTRUMENTATION MINIMUN APPLICABLE SIGNAL {g)'gc)OPERABLECHANNELg) OPERATIONAL ISOLATI n

PER TRIP SYSTEM CONDITION ACTION s

TRIP FUNCTION 7.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level Low, low - Level 2 B

2 1,2,3 25 b.

Drywell Pressure - High H

2 1,2,3 25 c.1. Refueling Area Unit 1 Ventilation 25 l

Exhaust Duct Radiation - liigh R

2 l

l

2. Refueling Area Unit 2 Ventilation 25 l

Exhausi. Duct Radiation - High R

2 g

+

w d.

Reactor Enclosure Ventilation Exhaust h

Duct Radiation - High 5

2 1, 2, 3 25 e.

Outside Atmosphere To Reactor Encidsure a Pressure - Low U

1 1,2,3 25 f.

Outside Atmosphere To Refueling 25 Area A Pressure - Low T

1 g.

Reactor Enclosure Manual Initistion NA 1

1,2,3 24 h.

Refueling Area Manual Initiation NA 1

25

TABLE J.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACY10N STATEMENTS ACTION 20 - Be in at least HOT SHUTOOWN witin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 21 - Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTOOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT 00WN within -the-next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 22 - Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 23 - in OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valves are closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

In OPERATIONAL CONDITION 3, be in at least COLD SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 24 - Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT 00WN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 25 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 26 - Close the affected system isvlation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

TABLE NOTATIONS Required when (1) handling irradiated fuel in the refueling area secondary containment, or (2) during CORE ALTERATIONS, or (3) during operations vith a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel, o*

May be bypassed under administrative control, with all turbine stop valves closed.

0 During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.

(a)

See Specification 3.6.3. Table 3.6.3-1 for primary containment isolation valves which are actuated by these isolation signals.

(b) A channel may be placed in an inoperable status for up to:

a) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for trip functions not common the the Reactor Protection System (RPS) and/or Emergency Core Cooling System (ECCS) Actu: tion Instrumentation, or b) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for trip functions common to RPS and/or ECCS Actuation Instrumentation for required surveillance without placing the trip system in the trippeo condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

Trip functions common to RPS and/or ECCS Actuation Instrumentation are shovm in Table 4.3.2.1-1.

In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is OPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for required surveillance without placing the channel or trip system in the tripped condition.

LIMERICK - UNIT 2 3/4 3-16 Amendment No.17

3, c

TABLE 3.3.2-1 (Continued).

TABLE NOTATIONS (c)

Actuates secondary containment isolation valves shown in Table 3.6.5.2.1-1 and/or 3.6.5.2.2-1 and signals B, H S. U, R and T also start the standby gas treatment system.

(d)

RWCU system inlet outboard isolation valve closes on SLCS "B" initiation.

RWCU system inlet inboard isolation valve closes on SLCS "A" or SLCS "C" initiation.

(e)

Manual initiation isolates the steam supply line outboard isolation valve and

.only following manual or automatic initiation of the system.

(f)

In the event of a loss of ventilation the temperature - high setpoint may be 0

raised by 50 F for a period not to exceed 30 minutes'to permit restoration of the ventilation flow without a spurious trip.

During the 30 minute period,-an operator, or other qualified member of the technical staff, shall observe the temperature indications continuously, so that, in the event of_ rapid increases in temperature, the main steam lines shall be manually isolated.

(g)

Wide range accident monitor per Specification 3.3.7.5.

LIMERIC< - UNIT 2 3/4 3-17 Amendment No.17

~

TABLE 3.3.2_2-ISOLATION ACTUATION INSTR M NTATION SETPOINTS M

n i

.ALLOWA8LE-c TRIP FUNCTION TRIP SETPOINT VALUE. _'.

2

' 1.

M4ll JTEAM LINE ISOLATION a.

Reactor Vessel Water Level.

1) Low, Low -' Level 2

> - 38 inches *

> - 45 inches 2)

Low,' Low,. Low.- Level 1 1:-'129, inches *

[-'136 inches

~

b.

Main. Steam Line

< 3.0 x Full' Power

< 3.6 x Full Pouer 4

Radiation - High

Background

Background c.

Main' Steam Line Pressure - Low'

-3 756 psig 1 736 psig T

d.

Main Steam Line-17.

Flow - High 5 108.7 psid 5 111.7 psid i

e.

Condenser Vacuum - Low 10.5 psia 1 10.1~ psia /5 14.9 psia' f.

Outboard MSIV Room i

Temperature - High 1 192*F

$ 200*F l

g.

Turbine Enclosure - Main Steam Line Tunnel Temperature - High i 165'F

$ 175*F h.

Manual Initiation N.A.

N.A.

2.

RHR SYSTEM SHUTDOWN COOLING MDDE ISOLATION L

a.

Reactor Vessel Water Level-Low

' Level 3 1 12.5 inches *-

1 11.0'inclus a

b.

Reactor Vessel (RHR Cut-in Permissive) Pressure - High 5'75 psig

--$ 95 psso c.

Manual Initiation N.A.

M.A.

i l

i

s TABLE 4.3.2.1-1 ISOLATION ACTUATION !NSTRUMENTATION SURVEILLANCE REQUIREMENTS M

m CHANNEL OPERATIONAL d

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICF S

TRIP FUNCll0N CHECK TEST CAllBRAll0N SURVEILLANCE REQUIRt i

g 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.

MainSteaggyne Radiation High S

Q R

1,2,3 c.

Main Steam Line Pressure - Low S

M R

1 d.

Main Steam Line Flow - High S

H R

1, 2, 3 e.

Condenser Vacuum - L,<

S M

R 1,

2**, 3**

O 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 Ini*iation H.A.

R N.A.

1, 2, 3 E

E 2.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION R

~ Reactor Vessel Water Level ###

E a.

Low - level 3 S

Q R

1,2,3 z

?

O b.

Reactor Vessel (RHRgt-In S

Q R

1, 2, 3 Permissive) Pressure

- High c.

Manual Initiation H.A.

R N.A.

1, 2, 3

. TABLE'4.3.2.1-1.(Continued)

ISGlIJION ACTUATION INSTRUMENTATION SURVElLLANCE~ REQUIREMENTS c.

w CHANNEL OPEPAT10NAL.

~

M CHANNEL' FUNCT10NAL CHANNEL CONDITIONS FOR WHICH 3.i

. TRIP FUNCTION CHECK-TEST CALIBRATION ' SURVEILIANCE REQUIRR i

- ' 3. -

REACTOR WATER CLEANUP SYSTEM ISOLATION w

a.

RWCS A' flow - High 5

M-R 1, 2, 3 b.

RWCS Area' Temperature - High S

M-R l

1, 2, 3 c.

RWCS Area Ventilation A Temperature:.High

'S M'

R 1, 2, 3 d.

SLCS Initiation N.A.

R:

. N.A.

1, 2, 3 II e.

Reactor Vessel Water Level

. Low, Low.

. Level.'2 S

'Q R

1, 2, 3' I

l f.

Manual Initiation-N.A.

R N.A.

1, 2, 3

{

4.

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION Y

a.

HPCI Steam Line M

A Pressure - High '

S

'M R

1. : 2, 3 '

b.

HPCI Steam Supply Pressure - Low 5

M R

1, 2, 3 c.

-HPCI Turbine Exhaust Diaphragm Pressure - High S

M R.

1, 2, 3

[

R R

d.

HPCI Equipment Room

[.

Temperature - High 5

M R

1, 2, 3 z

e.

HPCI Equipment Room

?

A Temperature - High S

M-R 1, 2, 3 '

~

f.

HPCI Pipe Routin) Area i.

Temperature - High S

M R

1, 2, 3 g.

Manual Initiation N.A.

R N. A..

1, 2, 3

~ >

h.

HPCI Steam Line A Pressure Timer

'N.A.

M R

1,.'2, 3

Li

~.

TABLE 4.3.'2.1-1-(Continued)-

C L%

ISOLATION ACTUATION INSTRUMENTATIOP SURVEILLANCE REQUIREMENTS i';

CHANNEL OPERATIONAL'

'Q CHANNEL-FUNCTIONAL CHANNEL CONDITIONS FOR 14flCH.

TRIP FUNCTION CHECK TEST-CAllBRATION-SURVEILLANCE REQUIRED c-5 5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION m

a.

RCIC Steam Line A Pressure - High-5 M~

R 1, 2. - 3 b.

RCIC Steam Supply Pressure -' Low 5

M R

1,:2, 3 c.

RCIC Turbine Exhaust Diaphragm Pressure - High

'S M

R 1, 2, 3.

R d.

RCIC Equipment Roon Y

Temperature

.High S

M R

1,2,3 U$

e.

RCIC Equipment Room A Temperature - High S'

M R

1,2,3 f.

RCIC Pipe Routing Area Temperature

.High S

M R

1, 2, 3 Manual' Initiation N.A.

R N.A.

1, 2, 3 h.

RCIC Steam Line A Pressure' Timer.

N.A.

M R

1, 2, 3 I

TABLE 4.3.2.1-l'(Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILIANCE REQUIREMENTS E"

CHANNEL OPERAT10NAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICI TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRL:

n l

6.

PRIMARY CONTAINMENT' ISOLATION a.

Reactor Vessel Water Leve1

1) Low, low - Level:2 S

Q R

1, 2, 3 2)

Low, Low, Low

. Level 1 S

Q R

1, 2, 3 Drywell Pressure ### - High S

Q R

1, 2, 3..

b.

c.

North Stack Effluent Radiation - High S

Q R

1,.' 2. 3 R

d.

Deleted v

Reactor Enclosure Ventilation Y

e.

'8 Exhaust Duct - Radiation - High S

M R

1, 2, 3 f.

Outside Atmosphere to Reactor Enclosure A Pressure - Low N.A.

M Q

1, 2, 3 g.

Deleted Drywell Pressure ## - High/

h.

[

Reactor Pressure

- Low S

Q R

1, 2, 3 II

s

[

i.

Primary Containment Instrument.

Gas to Drywell A Pressure - Low N.A.

M Q

I, 2, 3 g

,5 j.

Manual Initiation N.A.

R N.A.

1, 2, 3 C

e T

TABLE'4.3.2.1-1 (Continued)'

-ISOLATION ACTUATION INSTRUMEN1AlION SURVEILLANCE REQUIREMENTS-

'r-CHANNEL OPERATIONAL h

' CHANNEL FUNCTIONAL

CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST-CALIBRATION SURVEILLANCE REQUIRED.

A 7.

SECONDARY CONTAINMENT' ISOLATION h

a.

Reactor 4 Vessel Water. Level ##

g Low, Low ~~-Level 2 5

Q R

1~,

2, 3 III b.

Drywell Pressure

- High S

Q R-1,2,3

c. 1. Recueling Area Unft 1 Ventilation Exhaust. Duct Radiation.- High S'

M R

I

2. Refueling Area Unit 2 Ventilation.

.M' R

Exhaust Duct Radiation - High S

d.

-Reactor Enclosure Ventilation Exhaust Duct-Radiation - High S

M R

1,2,3 j

u e.

Outside Atmosphere To Reactor Enclosure A Pressure Low N.A.

-M Q

1, 2, 3 1

w 0

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 l

Manual Initiation N.A.

R N.A.

R l

5

Required when (1) handling irradiated fuel in the refueling area secondary containment, or (2) during CORE l l ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel ~with the vessel head removed and fuel in the vessel.

<+

l z

P

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

i G

i

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

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

l

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

i

1. 1. 1. This trip function-(Ib) is common to the-RPS' trip function.

INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITIuN FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 sha.ll be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3.

APPLICABILITY:

As shown in Table 3.3.3-1.

ACTION:

a.

With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.

c.-

With either ADS trip system subsystem inoperable, restor'e the

, inoperable trip system to OPERABLE status within:

1.

7 days, provided that the HPCI and RCIC systems are OPERABLE.

2.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 100 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations.for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3.1-1.

4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3 shall be demonst~.ated to be within the limit at least once per 18 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ECCS trip system.

LIMERICK - UNIT 2 3/4 3-32

7 5)[

TABLE 3...F t 4 ton.

EMERGENCY CORE COOLING ~ SYSTEM ACTUATION INSTRUMENTATION-TABLE NOTATIONS (a) channel may be placed-in an' inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for

.[

requiredzsurveillance without placing'the trip system in the tripped condition'provided at least one OPERABLE channel-in-the same trip system _is monitoring that parameter.

(b)_-Also provides input to actuation logic for'the associated emergency diesel generators.

(c) One trip system.. Provides' signal __to HPCI pump suction valves only.

(d)' On~1'out of_2 taken twice logic,-provides a signal to trip the'HPCI pump turbine only.

(e) The manual, initiation push buttons start the respective core spray pump

_and diesel generator. The "A" and "B" logic' manual push buttons'also actuate an initiation permissive in the' injection valve opening logic.

(f) -A:_ channel as used here is defined as the 127-bus relay for Item-1 and the-127,-127Y, and 127Z feeder relays with their_associats' time delay relays taken together for Item 2.

When the system is required to be OPERABLE per Specification 3.5.2.

.Not required ~to be OPERABLE:when reactor steam dome pressure is less than or' equal to 100 psig.

Required when ESF equipment is required to be OPERABLE.

Not required-to be OPERABLE when reactor steam dome pressure is less than or equal.to 200 psig, e

f n

' LIMERICK - UNIT'2 3/4 3-35 Amendment No.17

_l TABLE 3;3[3-1 (Continued)-

~

. EMERGENCY CORE COOLING SYSTEM ACTUATION' INSTRUMENTATION ACTION STATEMENTS ACTION 30.-

-With-the-number of.0PERABLE channels less:than required by the Minimum-0PERABLE Channels per Trip Function requirement:

a. M i'th one channel' inoperable, place the inoperable channel in the-tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare _the associated l

system i_noperable.

'b..-

With more than one channel inoperable, declare the associated system _ inoperable.

' ACTION 31 -

With the number of OPERABLE channels less than required by the.

Minimum OPERABLE Channels per Trip Function requirement, declare the associated ECCS inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l

-ACTION 32 -

With-the number cf OPERABLE channels less than required by the Mirimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

\\

' ACTION 33 -

With_.the number of OPERABLE channels less than required by the Minimum OPERABLE-Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within-24 hours or declare the

[

= associated ECCS inoperable.

~ ACTION 34 -

With the-number of OPERABLE channels less'than required by the Minimum OPERABLE Channels per Trip Function requirement:

a.

For one channel inoperable, place the inoperable chaanel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the HPCI system

)

inoperable.-

b.

. With more than one channel inoperable, declare the HPCI system inoperable.

ACTION 35 -

lWith the' number of-0PERABLE channels'less.than required by the M4nimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 24 g

r.'.urs.or declare the HPCI system inoperable.

ACTION'36 -

With the number of OPERABLE che.nnels less than the Total Numeer of Channels, declare the associated-emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2, as appropriate.

ACTION 37 -

With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channel in the tripped condi" on t

within-1 hour; operat lon may then continue until performance or the next required CHANNEL FUNCTIONAL TEST.

-LIMERICK - UNIT 2 3/4 3-36 Amendment No.17

- t e

TABLE 3.3.3-3 EMERGENCY 0%E COOLING SYSTEM RESPONSE TIMES ECCS RESP 0 HSE TIME (Seconds) 1.

CORE SPRAY SYSTEM i 27 2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM i 40 3.

AUTOMATIC DEPRESSURIZATION SYSTEM N.A.

4.

HIGH PRESSURE COOLANT INJECTION SYSTEM 1 30 5.

LOSS OF POWER N.A.

4-LIMERICK - UNIT 2 3/4 3-39

~

o

/ %.

TABLE-4.3.3.1-1 lh

~ EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE ~ REQUIREMENTS is CHANNEL OPERATIONAL' p;

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH:

W 8

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED Ey 1.

CORE SPRAY SYSTEM I

w Reactor Vessel Water Level -

Low' Low Low, Level I S

Q R

1, 2, 3, 4*, 5*

a.

b.

Drywell Pressure - High 5

Q 1, 2, 3 a

Reactor Vessel Pressure - Low S

Q R

'1, 2, 3, 4*, 5*

d.

Manual Initiation N.A.

R N.A.

1, 2, 3, 4*, 5*

c.

2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM Reactor Vessel Water Level -

Low Low ~ Low, Level 1 S

Q R

1, 2, 3, 4*, 5*

a.

ti

b..

Drywell Pressure - High 5

Q R

1, 2, 3 Reactor Vessel Pressure - Low S-Q R

1, 2, 3 e

c.

y d.

Injection Valve Differential Pressure - Low (Permissive)

S Q

R 1, 2, 3, 4*, 5*

i N.A.

R N.A.

1, 2, 3, 4*, 5*

g e.

Manual Initiation 3.

HIGH PRESSURE COOLANT INJECTION SYSTEM ***

Reactor Vessel. Water Level -

Low Low, Level 2 S

Q R

1,2,3 a.

1 b.

Drywell Pressure - High 5

Q R

1, 2, 3 Condensate Storage Tank Level -

I S

Q R

1, 2, 3 c.

Low d.

. Suppression Pool Water Level -

g S

Q R

1,2,3 High g

Reactor Vessel Water Level -

I 5

Q R

1,2,3 g

e.

High, level 8 g

Manual Initiation N.A.

R N.A.

1, 2, 3 f.

5 d

4 '

TABLE 4.3.3.1-I_ (Continued)

EMERGENCY EORE COOLING SYSTEM ACTUATION INSTRUMENTATION SULEILLANCE REQUIPEfCNTS r-5!

I E

GlANNEL OPERAll0NAL E

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHlCH 7

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED I

k 4.

AU10MATIC DEPRESS'JRIZAlION SYSTEM 4

j Reactor Vessel Water level -

I a.

Low Low Low, level 1 5

Q R

1, 2, 3 b.

Drywell Pressure - High 5

Q R

1, 2, 3 I c.

ADS Timer N.A.

Q Q

1, 2, 3 d.

Core Spray Pump Discharge Pressure - tilgh 5

Q R

I, 2, 3 RHR LPCI Mode Pump Discharge e.

S Q

R 1, 2, 3 Pressure - High f.

Reactor ^1essel Water Leve! - Low, S

Q R

1, 2, 3 I

level 3 g.

Manual Initiation N.A.

R N.A.

1, 2, 3 h.

ADS Drywell Pressure Bypass Timer h.A.

Q Q

1, 2, 3 l

I

(

y 5.

LOSS OF POWER l

l 4.16 kV Erergency Bus Underg voltage (Loss of Voltage)

N.A.

R N.A.

1, 2, 3, 4**, 5**

j a.

b.

4.16 kV Emergericy Bes Under-voltage (Degraded Voltage)

S M

R 1, 2, 3, 4'*, 5**

8 E

4

]

When the system is required to be OPERABLE per Specification 3.5.2.

s Required OPERABLE when ESF equipment is required to be OPERABLE.

Hot required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.

C No* required to be OPERABLE when rea(.cr steam dome pressure is lo s than or equal (o 100 psig.

'd Loss of Voltage Relay 127-llX is not field setable.

INSTRUMENTATION 3/4.3.4 RECIRCULAf!ON PUMP TRIP ACTUATION INi.iUMENTATION

., ? RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION

p;

'9G CONDITION FOR OPERATION 3.3.4.1 The anticipated transient without scram recirculation pump trip (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall be OPERABLE with their trip setpoints set consistent with values shown in the Trip Setpoint column of Table 3.3.4.1-2.

APPLICABILITY:

OPERATIONAL CONDITION 1.

ACTION:

a.

With an ATWS recirculation pump trip system instrumentation channel trin srtpoint less conservative than the value'shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , c.

With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:

1.

If the inoperable channels consist of one reactor vessel water level channel and one reactor vessel pressure ti nnel, place both inoperable channels in the tripped condition wi

o I hour, or, if this action will initiate a pump trip, declare the trip system inoperable.

2.

If the inoperable chtanels include two reactor vessel water level channels or two reactor vessel pressure channels, declare the trip system inoperable.

l d.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e.

With both trip systems inoperable, - store at least one trip system to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or L in it 14ast STARTUP within the ne> t 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS.

4.3.4.1.1.

Each ATVS recirculation pump trip system instrumentation' channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.1-1.

4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of

all channels shall be performed at least once per 18 months.

LIHERICK - UNIT 2 3/4 3-42 l

REALTOR CORE !$0LAT10N COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM l

OPERABLE CHANNELS l

FUNCTf g l UNITS PER TRIP FUNCTION

ACTION l

a.

Raactor Vessel Water Level -

Low CUJ. Level 2 4#

50 b.

Reactor Vessel Water level -

High, level 8 4#

51 J

0 c.

Condensate Storage Tank Water Level - Low 2**

52 d.

Manual Initiation 1/ system ***

53 C

1

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

}

a for required surveillance without placing the trip system in the tripped condition provided all other channels monitoring that parameter are OPERABLE.

- 0ne trip system with one-out-of-two logic.

- - 0ne trip system with one channel.

0ne trip system with or,e-out-of-two twice logic.

LIMERICK - UNIT 2 3/4 3-53 Amendment No. 17 3

t REACTOR CORE ISOLAi!0h COOLING SYSTEM ACTION STATEMENTS ACTION 50 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip function requirement:

-0. - With one channel inoperable, place the inoperable channel in i

thi tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the RCIC

\\'

system inoperable, b.

With more than one channel inoperable, declare the RCIC system inoperable.

ACTION With the number of 0PERABLE channels less than required by the minimum OPERABLE channels per Trip System requirement, declare the RCIC system inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l-

' ACTION 52 -

With the number of OPERADLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, place at least one inoperable channel in the tripped condition within 24

\\

hours or declare the RCIC system inoperable.

ACTION 53 -

With the nember of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the l

RCIC system inoperable.

i I

L i-u L

LIMERICK - UNIT 2 3/4 3-54 Amendment No. 17-D E,

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

.. ~.... _

i TABLF 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNITS TRIP SETPOINT VALUE a.

Reactor Vessel Water Level -

Low Low, Level 2 1-38 inches

1-45 inches b.

Reactor Vessel Water Level -

High, level 8 1 54 inches 1 60 inches c.

Condensate Storage Tank Leuel -

Low 1 135.8** inches 1 132.3 inches d.

Manual Initiation N.A.

N.A.

See Bases Figure B 3/4.3-1.

- Corresponds to 2.3 feet indicated.

I LIMERICK - UNIT 2 3/4 3-55

REACTOR CORE ISOLATION SYSTEM ACTUATION INSTRUMEN7ATION SURVEILLANCE REQUIREMENYS CHANNEL CHANNEL FUNCTIONAL CHANNEL FUNCTIONAL UNITS CHECK TEST CAllBRATION a.

Reactor Vessel Water Level -

Low Low. Level 2 5

Q R

l b.

Reactor Vessel Water Level -

High, Level 8 S

Q R

l c.

Condensatt Storage Tank Level - Low S

Q R

l d.

Manual Initiation N.A.

R N.A.

LIMERICK - UNIT 2 3/4 3-56 Amendment No.17

TABLE'4.3.6-1

[

.{

CONTROL R00 BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

CHANNEL OPERATIONAL 5

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH Q TRIP FUNCTION CHECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED h

1.

R00 BLOCK MONITOR i

i S/U(b)(c)

(c S/U(b)(c)*(c})

a.

Upscale N.A.

SA 1*

b.

Inoperative N.A.

N.A.

1*

c.

Downscale N.A.

S/U(b)(c)

(c)

SA 1*

2.

APRM j.

a.

Flow Biased Neutron Flux -

i S/U(b)

SA 1

Upscale N.A.

S/U(b),Q b.

Inope N.A.

Q N.A.

1, 2, 5***

S/U(ID),0 i

c.

Downscane N.A.

SA 1

S/U D),Q SA 2, 5***

d.

Neutron Flux - Upscale, Startup N.A.

f 3.

SOURCE RANGE MONITORS t

a.

Detector not full in N.A.

S/U(b) W N.A.

2, 5 S/U(D),W b),W SA 2, 5 b.

Upscale.

N.A.

I N.A.

2, 5 Y

c.

Inoperative N.A.

S/U(b),W d.

Downscale N.A.

S/U SA 2, 5 4.

INTERMEDIATE RANGE MONITORS

[

.S/U(D),Wb) W N.A.

2, 5 a.

Detector not full in N.A.

S/U(I b.

Upscale N.A.

SA 2, 5 S/fl b),

c.

Inoperative N.A.

S/U(b),W N.A.

2, 5

,g.

SA 2, 5 d.

Downscale N.A.

[

5.

SCRAM DISCHARGE VOLUME f

s a.

Water Level-High N A.

Q R

1, 2, 5**

l

[

[ 6.

REACTOR COOLANT SYSTEM RECIRCul_ATION FLL.J O

S/U(b),Q b)

SA 1

a.

Upscale N.A.

S/U(b),Q N.A.

I b.

Inoperative N.A.

S/U(

,Q SA 1

C; c.

Comparator N.A.

7.

REACTOR MODE SWITCH SHUTDOWN POSITIOf.

M.A.

R N.A.

3, 4 I

l TABLE 4.3.6-1 (Continued)

CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

+

(c) Includes reactor manual control multiplexing system input, With THERMAL POWER > 30% of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.

a c

LIMERICK - UNIT 2 3/4 3-62 Amendment No. 7

{

JUL 3 019g0 -

{.N

~.

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be absorbed following a 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 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 neclear power plant protection systems.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NE0C-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.

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.

Response time may be demonstrated by any series of sequential, overlapping or tot:1 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.

LIHERICK - UNIT 2 B 3/4 3-1 Amendment No.17

BASES 3/4.3.2 ISOLATIONACTUATIONINSTRUME$NTATION 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 NE00-30851P, Supplement 2 " Technical Specification Improvement Analysis for BKR 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).

Some of the trip settings may have toleranres 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 indivicual sensor response times or the response times of the logic systems to which the sensors are connected.

For 0.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 valve is assumed; thus the signal delay (sensor response) is concurrent with the 10-second diesel startup and the 3 second load center loading delay.

The safety l

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 itolation functions.

l l

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

within its specified Allowable Value is accepteble 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 l

analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentstion is provided to initiate actions to mitigate the consequences of accidents that are beyond the i

ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiver.ess of'the systems to provide the design protection. Although the instruments are li::ted by system, in some cases the same instrument may be used to send the l

actuation signal to more than one system at the same time.

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 l

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. H.

Grace from C. E. Rossi dated December 9, 1988 (Part 2)).

I WERICK - UNIT ?

8 3/.1 3-2 Amendment No.17

~

.,u BASES 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 RECIRCU8.ATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequench of the unlikely occurrence of a failure to scram during an anticipated transient.

The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NE00-10349, dated March 1971. NE00-24222, dated December 1979, and Section 15.8 of the FSAR.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a supplement to the reactor trip. During turbine trip and generator load rejection events, the EOC-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 EOC-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 EOC-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 EOC-RPT system; a position switch from each of the other two stop valves provides input to the other E0C-RPI system.

For each E0C-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 E0C-RPT system and trip both recirculation pumps.

Each FOC-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 THERHAL 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 i

i for instrument drift specifically allocated for each trip in the safety analyses.

LIMERICK - UNIT 2 B 3/4 3-3 Amendment No.17

BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel.

This instrumentation does not provide actuation of any of the emergence core cooling equipment.

Specified surveillance intervals and maintenance outage times have been specified in accordance with recommendations made by GE in their letter to the BWR Owner's Group dated August 7, 1989,

SUBJECT:

" Clarification of Technical Specification changes given in ECCS Actuation Instrumentation Analysis."

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 s M cifically allocated for each trip in the safety analyses.

3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4 Control Rod Program Controls and Section 3/4.2 Power Distribution Limits and Section 3/4.3 Instrumentation. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P, Supplement 1 " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

as approved by the NRC and documented in the SER (letter to 0. N. Grace from C.

E. Rossi dated September 22,1988).

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 are allowance for instrument drift specifically allocated for each trip in the safety analyses.

3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORlhG INSTRUMENTATION

. The OPERABILITY of-the radiation monitoring instrumentation ensures thatt (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A. General Design Criteria 19, 41, 60, 61, 63, and 64.

LIMERICK - UNIT 2 8 3/4 3-4 Amendment No. II,17

BASES 3/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a

' seismic event and evaluate the response of those features important to safety.

This capability is required to permit comparison of the measured response to tnat used in the design basis for the unit.

3/4.3.7.3 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE 00CM.

3/4.3.7.4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS

The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTOOWN of the unit from locations outside of the control

room, lhis capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50 Appendix A.

The Unit 1 RHR transfer switches-are included only due to their potential impact on the RHRSW system, which is common to both units.

3/4.3.7.5 ACCIDENT MONITORING INSTRUME' R ION The OPERABILITY of the accident monn. ring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an. accident. This capability is consistent with the recommendations of Regulatory Guide 1.97. " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown.- At these power levels, reactivity additions shall not be made witnout this flex level information available to the operator. When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.

3/4.3.7.7 TRAVERSING IN-CO$E PROBE SYSTEM l

The 0PERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures t b t the measurements obtained from use l

of this equipment accurately represent the spatial neutron flux distribution of the reactor core.

l

_The TIP system OPERABILITY is demonstrated by normalizing all probes (i.e.,

l detectors) prior to performing an LPRM calibration function. Monitoring core l

thermal limits may involve utilizing individual detectors to monitor selected areas of the reactor core, thus all detectors may not be required to be OPERABLE. We OPERABILITY of individual detectors to be used for monitoring is demonstrated by comparing the detector (s) output in the resultant heat balance calculation (P-1) with data obtained during a previous heat balance calculation (P-1).

LIMERICK - UNIT 2 B 3/4 3-5 Amendment No. II,17

~.

_.. _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _. _ _ _ _ _ _ _ _ _ ~ _ _ _ _ _ _ _ _.

INSTRUMENTATION BASES 3/4.3.7.8 CHLORINE AND 70XIC GAS DETECTION SYSTEMS The OPERA 81LITY of the chlorine and toxic gar @tection systems ensures that an accidental chlorine and/or toxic gas releaw will be detected promptly and the necessary protective actions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for control room personnel.

Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically be placed in the chlorine isolation mode of operation to provide the required protection.

l'pon detection of a high concentration of toxic gas, the control roce emergency ventilation system will manually be placed in the chlorine isolation mode of operation to provide the required protection.

The detection systems required by this s ficationareconsistentwiththerecommendationsofRegulatoryGuide1.95geci-Pro-taction of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release," February 1975.

3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERA 8ILITY of the detection instrumentation ensures that-both adequate l

warning capability is available for prompt detection of fires and that fire suppression systems, that are actuated by fire detectors, will. discharge extin-guishing agent in a timely manner.

Proept detection and suppression of fires will reduce the potential-for damage to safety-related equipment and is an

. integral element in the overall facility fire protection program.

Fire detectors that are used to actuate fire suppression-systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification.-

Consequently, the minimum number of OPERA 8LE fire detectors must be greater.

The loss of detection capability for fire suppression systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. As a result, the establisheept of a fire watch patrol must be initi-i ated at an. earlier stage than would be warranted for the loss of detectors that L

provide only early fire warning.

The establishesnt of frequent fire patrols

in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERASILITY.

3/4.3.7.10 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose-part detection systes ensures that sufficient capability is available to detect loose estallic parts in the primary system-and avoid or mitigate daeane to primary system components.

The allowable out-of-cervice times and surve'11ance requirements are consistent with the recom-mandations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.

3/4.3.7.11 (Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO TH2 00CM.

i LIERICK - UNIT 2 8 3/4 3-6 Amendment No.11 4

S/

1AhML QMMy c?e

- - - - - -

ML20086K543

Contents

Text

Attachment:

Background

SUBJECT:

Background

SUBJECT:

Navigation menu

ML20086K543

Text

Attachment:

Background

SUBJECT:

Background

SUBJECT:

Navigation menu

Search