Text

Amends 70 & 33 to Licenses NPF-39 & NPF-85,respectively, Providing Extension of Surveillance Test Intervals & AOT for Selected Actuation Instrumentation TS & Make Editorial Changes
	ML20070N100
Person / Time
Site:	Limerick
Issue date:	04/26/1994
From:	Chris Miller; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20070N103	List: ML20070N100; ML20070N105;
References
	NUDOCS 9405050086
	Download: ML20070N100 (76)
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UNITED STATES Qg i

NUCLEAR REGULATORY COMMISSION gv,/

WASHINGTON, D.C. 20555-0001 PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. NPF-39 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Philadelphia Electric Company (the licensee) dated May 6,1993, as supplemented by letter dated April 18, 1994, 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; 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; D.

The issuance of this amendment will not be inimical to the common 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 j

satisfied.

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i 9405050086 940426 i

yDR ADOCK 05000352 PDR I

l:

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-l-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 Facility Operating License No. NPF-39 is hereby

<.nended 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.

70, are hereby incorporated into this license.

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

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation l

Attachment:

Changes to the l

Technical Specific'ations Date of Issuance: April 26,1994 1

-o ATTACHMENT TO LICENSE AMENDMENT N0. 70 FACILITY OPERATING LICENSE N0. NPF-39 QDCKET NO. 50-352.

Replace the following pages of the Appendix A Technical Specifications with the attached 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 j

3/4 1-5 3/4 1-5 l

3/4 1-6 3/4 1-6*

3/4 3-41 3/4 3-41*

3/4 3-42 3/4 3-42 3/4 3-43 3/4 3-43 l

3/4 3-44 3/4 3-44*

l 3/4 3-45 3/4 3-45 3/4 3-46 3/4 3-46 3/4 3-47 3/4 3-47*

l 3/4 3-48 3/4 3-48 3/4 3-51 3/4 3-51 3/4 3-52 3/4 3-52*

3/4 3-57 3/4 3-57 1

3/4 3-58 3/4 3-58*

1 3/4 3-59 3/4 3-59 3/4 3-60 3/4 3-60*

3/4 3-65 3/4 3-65*

- ~ ~ ~

l 3/4 3-66 3/4 3-66 l

3/4 3-89 3/4 3-89*

3/4 3-90 3/4 3-90 l

3/4 3-111 3/4 3-111*

3/4 3-112 3/4 3-112 3/4 3-115 3/4 3-115 3/4 4-7 3/4 4-7 3/4 4-8 3/4 4-8*

..m e-e 5-ys y

-WMa9-q-e-my*

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ATTACHMENT T0 LICENSE AMENDMENT NO. 70 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 i

Replace the following pages of the Appendix A Technical Specifications with the attached 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 3/4 6-13 3/4 6 13*

3/4 6-14 3/4 6-14 8 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 8 3/4 3-7 B 3/4 3-7 B 3/4 3-8 B 3/4 3-8*

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i REACTIVITY CONTROL SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) 4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by demonstrating:

a.

The scram discharge volume drain and vent valves OPERABLE, when control rods are scram tested from a normal control rod configura-tion of less than or equal to 50% R00 DENSITY at least once per 18 months, by verifying that the drain and vent valves:

1.

Close within 30 seconds after receipt of a signal for control rods to scram, and 2.

Open when the scram signal is reset.

b.

Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod block level instrumentation at least once per 92 days.

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LIMERICK - UNIT 1 3/4 1-5 Amendment' No 70 l

REACTIVITY CONTROL SYSTEMS CONTROL R00 MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 5, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

l ACTION:

a.

With the maximum scram insertion time of one or more control rods exceeding 7 seconds:

1.

Declare the control rod (s) with the slow insertion time inoperable, and 2.

Perform the Surveillance Requirements of Specification 4.1.3.2c.

at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.

i Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demon-strated through measurement with reactor coolant pressure greater than or i

equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:

For all control rods prior to THERMAL POWER exceeding 40% of RATED a.

THERMAL POWER following CORE ALTERATIONS or after a reactor shutdown that is greater than 120 days.

b.

For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods, and c.

For at least 10% of the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION.

LIMERICK - UNIT 1 3/4 1-6

IABLE 4.1 3 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREM r

h,,,

CHANNEL OPERA 110NAL M

CHMNEL FUNCTIONAL CHANNEL CONDIIIONS FOR WHICH TRIP FUNCTION CHECK TEST CAllBRATION SURVEILLANEE REQUIRED E

4.

AUTOMATIC DEPRESSURIZATION SYSTEMI li' a.

Reactor Vessel Water Level -

Low Low Low, level 1 5

0 R

1, 2, 3 b.

Drywell Pressure - High 5

Q R

1, 2, 3 c.

ADS Timer N.A.

Q Q

1, 2, 3 d.

Core Spray Pump Discharge Pressure - High 5

Q R

1, 2, 3 l

e.

RHR LPCI Mode Pump Discharge Pressure - High S

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 M.A.

Q Q

1, 2, 3 4

5.

LOSS OF POWER o-Y a.

4.16 kV Emergency Bus underg voltage (Loss of Voltage)

N.A.

R N.A.

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

.s b.

4.16 kV Emergency Bus Under-voltage (Degraded Voltage)

S M

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

t

(

g When the system is required to be OPERA 8LE per Specification 3.5.2.

L $

Required OPERABLE when ESF equipment is required to be OPERABLE.

5 g

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

h i Not required to be OPERA 8LE when reactor steam done pressure is less than or. equal to 100 psig.

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

NNs

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INSTRUMENTATION I

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION I

ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION j

LIMITING CONDITION FOR OPERATION l

l 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 i

j OPERABLE with their trip setpoints set consistent with values shown in the Trip i

Setpoint column of Table 3.3.4.1-2.

I APPLICABILITY: OPERATIONAL CONDITION 1.

I ACTION:

i a.

With an ATWS recirculation pump trip system instrumentation 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 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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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 if this action will initiate a pump trip, declare the trip system inoperable channels in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or, j

4 inoperable.

I 2.

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

i 4

i 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 .

Withbothtripsystemsinoperable,restoreatIe[sIonetripsystem e.

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 .

i SURVEILLANCE REOUIREMENTS I

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 Amendment No. 70

TABLE 3.3.4.1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUM OPERABLE CHANNELS PER TRIP FUNCTION TRIP SYSTEM

1.

Reactor Vessel Water Level -

Low Low, Level 2 2

2.

Reactor Vessel Pressure - High 2

.=:-

One 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 provided the other channel is OPERABLE.

1 LIMERICK - UNIT 1 3/4 3-43 Amendment No. 70

I TABLE 3.3.4.1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE TRIP FUNCTION SETPOINT VALUE 1.

Reactor-Vessel, Water Level -

Low Low, Level 2 1 -38 inches

2 -45 inches 2.

Reactor Vessel Pressure - High

< 1093 psig 5 1108 psig l

l

See Bases Figure B3/4 3-1.

LIMERICK - UNIT 1 3/4 3-44

. -. - ~. -.. -.

4 TABLE 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION

~

SURVEILLANCE REOUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION CHECK TEST CALIBRATION I.

Reactor Vessel Water Level -

l Low Low, Level 2 S

Q R

l 2.

Reactor Vessel Pressure -

High S

Q R

k 4

i 4

i

.1

,m~.

3 l

LIMERICK - UNIT 1 3/4 3-45 Amendment No. 70

INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2

- The end-of-cycle recirculation pump trip (EOC-RPT) system instrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.

ACTION:

a.

With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

i b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both 2

l!

tripsystems,placetheinoperablechannel(s)inthetrippedcondition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

With the number of OPERABLE channels two or more less than required i

by the Minimum OPERABLE Channels per Trip System requirement for one trip system and-l 1.

If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l) 2.

If the inoperable channels include two turbine control valve channels or two turbine stop valve 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 or take_the ACTION required by 1

Specification 3.2.3.

e.

With both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or take the ACTION required by Specification 3.2.3.

LIMERICK - UNIT 1 3/4 3-46 Amendment No. 70

INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.4.2.1 Each end-of-cycle recirculation pump trip system instrumentation channel.shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies _shown in j

Table 4.3.4.2.1-1.

4.3.4.2.2.

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

4.3.4.2.3 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function shown in Table'3.3.4.2-3 shall be demonstrated to'be within its limit at least once per 18 months.- Each test shall include at least the logic of one type of channel input, turbine control valve fast closure or turbine stop valve closure, such that both types of channel inputs are tested at least once per 36 months.

The measured time shall be added to the most recent breaker arc suppression time and the resulting END-OF-CYCLE-RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be verified to be within its limit.

4.3.4.2.4 The time interval necessary for breaker arc suppression from energi-zation of the recirculation pump circuit breaker trip coil.shall be measured at least once per 60 months.

mm.%

LIMERICK - UNIT 1 3/4 3-47

TABLE 3.3.4.2-1

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END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUM OPERABLE CHANNELS TRIP FUNCTION PER TRIP SYSTEM

1.

Turbine Stop Valve - Closure 2**

2.

Turbine Control Valve-Fast Closure 2**

8 P

)

.5

==

A trip system 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 provided that the other trip system is OPERABLE.

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

LIMERICK UNIT 1 3/4 3-48 Amendment No. 70

TABLE 4.3.4.2.1-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE RE0VIREMENTS CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION TEST CALIBRATION I.

Turbine Stop Valve-Closure Q*

R 2.

Turbine Control Valve-Fast closure Q*

R 4

'~_

i

Including trip system logic testing.

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e LIMERICK - UNIT 1 3/4 3-51 Amendment No. 70

INSTRUMENTATION 3/4.3.5 REACTOR CORE' ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 4

3.3.5 The _ reactor core isolation cooling (RCIC) system actuation instrumentation channels shown in Table 3.3.5-1 shall be OPERABLE with their trip setpoints-set consistent with the values shown in the Trip Setpoint column of Table 3.3.5-2.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig.

ACTION:

With a RCIC system actuation instrumentation channel trip setpoint a.

less conservative than the value shown.in the Allowable Values column of Table 3.3.5-2, 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 RCIC system actuation instrumentation channels inoperable,.take the ACTION required by Table 3.3.5-1.

SURVEILLANCE REQUIREMENTS

~

4.3.5.1 Each RCIC system actuation instrumentation channel shal.1 be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies'shown in Table 4.3.5.1-1.

4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed'at least once per 18 month's.

I LIMERICK - UNIT 1 3/4 3-52

INSTRUMENTATION 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.

The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.

APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

a.

With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, 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 Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE RE0VIREMENTS 4.6.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE

by the performance of l

the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.

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

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LIMERICK - UNIT 1 3/43-57 Amendment No. 70

i TABLE 3.3.6-1 5E CONTROL ROD BLOCK INSTRUMENTATION MINI Mt APPLICABLE R

4 OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION E

1.

ROD BLOCK MDNITOR(a) i Q

a.

Upscale 2

1*

60 b.

Inoperative 2

1*

60

c. 'Downscale 2

1*

60 2.

APRM a.

Flow Biased Neutron Flux -

Upscale 4

1 61 II) l l

b.

Inoperative 4

1, 2, S 61 c.

Downscale 4

1 61 d.

Neutron Flux - Upscale, Startup 4

2, 5 61 l

ff)

R 3.

SOURCE RANGE MONITORS ***

b Detector not full in(b) 3 2

61 a.

J, 2

5 61 b.

Upscale (c) 3 2

61 2

5 61 Inoperative (C) f f

h c.

d.

Downscale(d) g 4.

INTERMEDIATE RANGE MONITORS c_ a a.

Detector not full in 6

2,. 5 61

p jg b.

Upscale 6

2, 5 61 Inoperati 6

2, 5 61 w%

c.

Downscale{g) l, 6

2, 5 61 l e, 2 d.

l 5.

SCRAM DISCHARGE VOLUME l

a.

Water Level-High 2

1, 2, 5**

62

=

l 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW

~

a.

Upscale 2

1 62 i

b.

Inoperative 2

1 62 c.

Comparator 2

1 62 i

7.

' MTOR M00E SWITCH SHUT 00WN POSITION 2

3, 4 63 l

i.

TABLE 3.3.6-1 (Continued)

L CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION i

ACTION STATEMENTS i

ACTION 60 Declare the RBM inoperable and take the ACTION required

{

by Specification 3.1.4.3.

t.

l ACTION 61 With the number of OPERABLE Channels:

1 a.

One less-than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or place the inoperable channel in the tripped condition.

b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.

i With the number of OPERABLE channels less than required by the ACTION 62 Minimum OPERABLE Channels per Trip Function requirement, place I

the inoperable channel in.the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l With the number of OPERABLE channels less than required by the ACTION 63 Minimum OPERABLE Channels per Trip Function requirement, initiate j

a rod block.

[

For OPERATIONAL CONDITION of Specification 3.1.4.3.

j

~

With more than one control rod withdrawn.

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

i These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.

(a)

The RBH shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30%'of i

RATED THERMAL POWER.

. In 1

3

'(b)

This function shall be automatically bypassed if detector count rate is

> 100 cps or the IRM channels are on range 3 or higher.

(c)

This function is automatically bypassed when the associated IRM channels j

are on range 8 or higher.

4 (d)

This function is automatically bypassed when the IRM channels are on-range 3 or higher.

(e)

This function is automatically bypassed when the IRM channels are on range 1.

(f)

Required to be OPERABLE only prior to and during shutdown margin i

demonstrations as performed per Specification 3.10.3.

i 1

LIMERICK - UNIT 1 3/4 3-59 Amendment No. 4, 4I, 66, 70

h TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT All0WABLE VALUE 1.

R0D BLOCK MONITOR r-52 a.

Upscale

9 1)

Low Trip Setpoint (LTSP)

M 2)

Intermediate Trip Setpoint (ITSP) f

3) High Trip Setpoint (HTSP)

E b.

Inoperative N/A N/A Z

c.

Downscale (DTSP) d.

Power Range Setpoint

26% RATED THERMAL POWER

1) Low Power Setpoint (LPSP) 23% RATED THERMAL POWER 2)

Intermediate Power Setpoint (IPSP) 58% RATED THERMAL POWER 61% RATED THERMAL POWER

3) High Power Setpoint (HPSP) 78% RATED THERMAL POWER 81% RATED THERMAL POWER 2.

8PBS a.

Flow Blased Neutron Flux - Upscale

1) During two recirculation loop s 0.66 W + 59%

s 0.66 W + 63%

I operation

2) During single recirculation loop 5 0.66 W + 54%

s 0.66 W + 58%

1

.t' operation b.

Inoperative N.A.

N.A.

i' c.

Downscale 2 4% of RATED THERMAL POWER 2 3% of RATED THERMAL POWER g

d.

Neutron Flux - Upscale, Startup*

5 12% of RATED THERMAL POWER s 14% of RATED THERMAL POWER 3.

SOURCE RANGE MONITORS a.

Detector not full in

.N.A.

N.A.

g b.

Upscale s 1 x 10' cps 51.6 x 10' cps g

c.

Inoperative N.A.

N.A.

g d.

Downscale 2 3 cps **

21.8 cps **

e S

4 INTERMEDIATE RANGE MONITORS z

a.

Detector not full ini N.A.

N.A.

P b.

Upscale li s 108/125 divisions of s 110/125 divisions of full scale full scale

-4 c.

Inoperative N.A.

N.A.

~

3 d.

Downscale 2 5/125 divisions of full 2 3/125 divisions of full scale scale E

5.

SCRAM _ DISCHARGE VOLUME a.

Water Level-High s 257' 5 9/16" elevation ***

s 257' 7 9/16" elevation hy a.

Float Switch

4 J

TABLE 3.3.7.1-1 (Continued) i RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS

When irradiated fuel is being handled in the secondary containment.

J (a) With fuel in the spent fuel storage pool.

(b) Alarm only.

ACTION STATEMENTS ACTION 70 With one monitor inoperable, restore the inoperable monitor to the OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j

initiate and maintain operation of the control room emergency filtration system in the radiation-isolation mode of operation.

1 With two or more of the monitors inoperable, within one hour, initiate and maintain operation of the control'* room emergency filtration system in the radiation mode of operation.

ACTION 71 With one of the required monitor inoperable, ass ~ure a portable continuous monitor with the same alarm setpoint is OPERABLE in i

the vicinity of the installed monitor during any fuel movement.

If no-fuel movement is being made, perform area surveys of the monitored area with portable monitoring instrumentation at least

{

once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

)

ACTION 72 With the required monitor inoperable, obtain and analyze at least one grab sample of the monitored parameter at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

J ACTION 73 With the required monitor inoperable,' assure a portable alarming monitor is OPERABLE in the vicinity of. the installed monitor or i

perform area surveys of the monitored area with-portable monitor-ing instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

~.

~iT_

4 i

LIMERICK - UNIT 1 3/4 3-65

E:

TABLE 4.3.7.1-1 ll RADIATION HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS n

OPERATIONAL CHANNEL CONDITIONS FOR E5 CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE El INSTRUMENTATION CHECK TEST CALIBRATION IS REQUIRED 1.

Main Control Room Normal Frash Air Supply Radiation Monitor S

Q R

1, 2, 3, 5 and

l 2.

Area Monitors a.

Criticality Monitors 1)

Spent Fuel Storage S

M R

(a)

,,g Pool I'

b.

Control Room Direct S

M R

At All Times EI Radiation Monitor 3.

Reactor Enclosure Cooling Water Radiation Monitor S

M R(b)

At All Times ir g

e

(

4 INSTRUMENTATION TRAVERSING IN-CORE PROBE SYSTEM LIMITING CONDITION FOR OPERATION 1

{

3.3.7.7 The traversing in-core probe system shall be OPERABLE with:

Five movable detectors, drives and readout equipment to map.the core, a.

and b.

Indexing equipment to allow all five detectors to be calibrated in a common location.

l APPLICABILITY: When the' traversing in-core probenis used for:

i a.

Recalibration of the LPRM detectors, and l

b.

Monitoring the APLHGR, LHGR, MCPR, or MFLPD.

ACTION:

With the traversing in-core probe system inoperable, suspend use of the system for the above applicable monitoring or calibration functions. The provisions of Specification 3.0.3 are not applicable.

j e

i SURVEILLANCE REQUIREMENTS AP' 4.3.7.7 The traversing in core probe system shall be demonstrated OPERABLE by normalizing each of the above required detector outputs within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to use for the LPRM calibration function.

s

't

~.

4

0nly the detector (s) in the required measurement location (s) are required to be OPERABLE.

LIMERICK - UNIT 1 3/4 3-89 Amendment No. 11 NOV 7 1988

INSTRUMENTATION CHLORINE DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.8.1 Two independent chlorine detection system subsystems shall be OPERABLE with their alarm and trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 0.5 ppa APPLICABILITY:

All OPERATIONAL CONDITIONS.

ACTION:

a.

With one chlorine detection subsystem inoperable, restore the inoperable' detection system to OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , initiate and maintain operation of at least one control room emergency filtration system subsystem in the chlorine isolation mode of operation.

b.

With both chlorine detection subsystems inoperable, within I hour initiate and maintain operation of at least one control room emer-gency filtration system subsystem in the chlorine isolation mode of operation.

i SURVEILLANCE RE0VIREMENTS 4.3.7.8.1 Each of the above required chlorine detection system subsystems shall be demonstrated OPERABLE by performance of a:

a.

CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

CHANNEL FUNCTIONAL TEST at least once per 92 days, and l

c.

CHANNEL CALIBRATION at least once per 18 months.

. = _

l 8

LIMERICK - UNIT 1 3/4 3-90 Amendment No.11, 70 i

s

INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) l l

b.

At least once per 31 days by:

1.

Cycling each of the following valves through at least one complete cycle from the running position:

a)

For the overspeed protection control system; 1)

Four high pressure turbine control valves b)

For the electrical overspeed trip system and the mechanical overspeed trip system; 1)

Four high pressure turbine control valves j

c.

At least once per 18 months by performance of a CHANNEL CALIBRATION of the turbine overspeed protection instrumentation.

I i

d.

At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of all valve seats, disks and stems and verifying no unacceptable flaws or excessive corrosion.

If unacceptable flaws or excessive corrosion are found, all other valves of that type shall be inspected, a

l

. ru i

LIMERICK - UNIT l' 3/4 3-111.

Amendment No. 33

- - - - - - -00T 3 01989- - - = - - - - -

INSTRUMENTATION 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION i

LIMITING CONDITION FOR OPERATION 3.3.9 The feedwater/ main turbine trip system actuation instrumentation channels shown in the Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent j

with the values shown in the Trip Setpoint column of Table 3.3.9-2.

APPLICABILITY:

As shown in Table 3.3.9-1.

ACTION:

l a.

With a feedwater/ main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.9-2, declare the channel inoper-able and either place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip set-l point adjusted consistent.with the Trip Setpoint value, or declare the associated system inoperable.

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days 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 .

c.

With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels 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 .

SURVEILLANCE REOUIREMENTS 4.3.9.1 Each feedwater/ main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE

by the perfomance of the CHANNEL CHECK, CHANNEL l

FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.9.1-1.

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

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 required surveillance without placing the trip system in the tripped condition.

LIMERICK - UNIT 1 3/4 3-112 Amendment No.

70

TABLE 4.3.9.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE RE0VIREMENTS OPERATIONAL CONDITIONS CHANNEL FOR WHICH CHANNEL FUNCTIONAL CHANNEL SURVEILLANCE TRIP FUNCTION CHECK TEST CALIBRATION REOUIRED l

1.

Reactor Vessel Water D

Q R

1 Level-High, level 8

- =.

LIMERICK - UNIT 1 3/4 3-115 Amendment No. 70-

,- REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 11 of the following reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*#

4 safety / relief valves 01130 psig i 1%

5 safety / relief valves @ 1140 psig i 1%

5 safety / relief valves 91150 psig 1%

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With the safety valve function of one or more of the above required safety / relief valves inoperable, be in at least HOT SHUTDOWN 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 .

b.

With one or more safety / relief valves stuck open, provided that suppression pool average water temperature is less than 105'F, close the stuck open safety / relief valve (s); if unable to close the stuck open valve (s) within 2 minutes or if suppression pool average water temperature is 110*F or greater, place the reactor mode switch in the Shutdown position.

c.

With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE0VIREMENTS 4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.20 of the full open noise leve1## by performance of a:

a.

CHANNEL FUNCTIONAL TEST at least once per 92 days, and a j

b.

CHANNEL CALIBRATION at least once per 18 months **.

4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously, set _ pressure tested and stored in accordance with manufacturer's recommendations at least once per 24 months, and they shall be rotated such that all 14 safety relief valves are removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 54 months.

The lift setting pressure shall correspond to ambient conditions of the o

valves at nominal operating temperatures and pressures.

The provisions of Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling.

1. Initial setting shall be in accordance with the manufacturer's recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

LIHERICK - UNIT 1 3/4 4-7 Amendment No. 56, 70

l REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS l

LIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant sy.

i leakage detection systems shall be OPERABLE:

The primary containment atmosphere gaseous radioactivity monitoring a.

system, b.

The drywell floor drain sump and drywell equipment drain tank flow monitoring system, The drywell unit coolers condensate flow rate monitoring system, and c.

d.

The prin+y containment

,ure-and temperature monitoring system.

c APPLICABILITY:

0FtRATIONAL CONDi: 10NS 1, 2 and 3.*

ACTION:

With only three of the above required leakage detection systems 0PERABLE, operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous radioactive monitoring system, primary containment pressure and temperature monitoring system and/or the drywell unit coolers condensate flow rate monitoring system is inoperable; 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems'shall be demonstrated OPERABLE by-l Primary containment atmosphere gaseous radioactivity monitoring a.

systems performance of a CHANNEL CHECK at least once-per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

b.

The primary containment pressure shall be monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and the primary containment temperature shall be monitored at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , c.

Drywell floor drain sump and Drywell equipment drain tank flow monitor-ing system performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION TEST at least once per 18 months.

d.

Drywell unit coolers condensate flow rate monitoring system-performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

"The primary containment atmosphere gaseous radioactivity monitor is not required to be OPERABLE until OPERATIONAL CONDITION 2.

LIMERICK - UNIT 1 3/4 4-8

C CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued) 3.

With the suppression chamber average water temperature greater than 120*F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With only one suppression chamber water level indicator OPERABLE and/or c.

i with less than eight suppression pool water temperature indicators, one in each of the eight locations OPERABLE, restore the inoperable indicator (s) to OPERABLE status within 7 days or verify suppression i

chamber water level and/or temperature to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d.

With no suppression chamber water level indicators OPERABLE and/or with less than seven suppression pool water temperature indicators covering at least seven locations OPERABLE, restore at least one water level indicator and at least seven water temperature indicators to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next l

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1 With the drywell-to-suppression chamber bypass leakage in excess of

{

e.

the limit, restore the bypass leakage to within the limit prior to i

increasing reactor coolant temperature above 200*F.

SURVEILLANCE REQUIREMENTS 1

4.6.2.1 The suppression chamber shall be demonstrated OPERABLE:

By verifying the suppression chamber water volume to be within the a.

limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the suppression chamber average water temperature to be less than or equal to 95'F, except:

1.

At least once per 5 minutes during testing which adds heat to i

the suppression chamber, by verifying the suppression chamber average water temperature less than or equal to 105'F.

2.

At least once per hour when suppression chamber average water temperature is greater than or equal to 95'F,-by verifying:

a)

Suppression chamber average water temperature to be less than or equal to 110'F, and b)

THERMAL POWER to be less than or equal to 1% of RATED THERMAL POWER 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter suppression chamber average water temperature has exceeded 95'F for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to 95'F, by verifying suppression chamber average water temperature less than or equal to 120*F.

LIMERICK - UNIT 1 3/4 6-13

CONTAINMENT SYSTEMS l

SURVEILLANCE REOUIREMENTS (Continued) c.

By verifying at least 8 suppression pool water temperature' indicators in I

at least 8 locations, OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and 3.

CHANNEL CALIBRATION at least once per 18 months, with the temperature alarm setpoint for:

l 1.

High water temperature:

l a)

First setpoint s 95 F b)

Second setpoint 5 105 F i

c)

Third setpoint s 110*F d)

Fourth setpoint s 120*F d.

By verifying at least two suppression chamber water level indicators OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 2.

CHANNEL FUNCTIONAL TEST at least once per 92 days, and 3.

CHANNEL CALIBRATION at least once per 18 months, with the water level alarm setpoint for high water level 5 24'l-1/2"'

e.

Drywell-to-suppression chamber bypass leak tests shall be conducted at I

40 +/- 10 month intervals to coincide with the ILRT at an initial differential pressure of 4 psi and verifying that the A/(k calculated from the measured leakage is within the specified limit.

If any drywell-to-suppression chamber bypass leak test fails to meet the specified-limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission.

If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed.

f.

By conducting a leakage test on the drywell-to-suppression chamber vacuum i

breakers at a differential pressure of at least 4.0 psi and verifying that the total leakage area A//k contributed by all vacuum breakers is less than or equal to 24% of the specified limit and the leakage area for an individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the drywell-to-suppression chamber bypass

)

leak test in Specification 4.6.2.1.d is not conducted.

1 LIMERICK - UNIT 1 3/4 6-14 Amendment No 68, 70

INSTRUMEN'ATION BASES 3/4.3.3 EMERGENCY CORE COOLING ACTUATION INSTRUMENTATION -(Continued)

'Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts I and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D. N. Grace from A. C. Thadant dated December 9,1988 (Part 1) and letter to D. N. Grace from C. E. Rossi dated December 9, 1988 (Part 2)).

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety 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,' NEDO-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 l

to the E0C-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-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 EOC-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 EOC-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.

i LIMERICK - UNIT 1 B 3/4 3-3 Amendment No. 53. H, 70 g

I 1

INSTRUMENTATION BASES j

3/4.3.4 RECIRCULATION PUMP TRIP ACTVATION INSTRUMENTATION (Continued)

Specified surveillance intervals and maintenance outage times have been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D.

Binz, IV, from C.E. Rossi dated July 21,1992).

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.

3/4.3.5 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 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 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 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 analyses.

j LIMERICK - UNIT 1 B 3/4 3-4 Amendment No. 4. $3. 70

INSTRUMENTATION BASES 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 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.

The specified surveillance interval for the Main Control Room Normal Fresh Air Supply Radiation Monitor has been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D. Binz, IV, from C.E. Rossi dated July 21, 1992).

3/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION i

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 that used in the design basis for the unit.

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

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 SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habit;bility 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't5~ 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,"

i 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 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.

LlMERICK - UNIT I B 3/4 3-5 Amendment No. 48, 53, 70

INSTRUMENTATION BASES 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 obtained from use of this i

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).

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 emercency ventilation system will manually be placed in the chlorine isolation moce 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.

Specified surveillance intervals and maintenance outage times have been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D.

Binz, IV, from C.E. Rossi dated July 21,1992).

.n 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION i

OPERABILITY 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-quishing 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.

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

LIMERICK - UNIT 1 B 3/4 3-6 Amendment No. 48, 60, 70 j

t ',

INSTRUMENTATION BASES FIRE DETECTION INSTRUMENTATION (Continued) 4 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 establishment of a fire watch patrol must be initi-ated at an earlier stage than would be warranted for the loss of detectors that i

provide only early fire warning. The establishment of frequent fire patrols j

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

The surveillance requirements for demonstrating the OPERABILITY of the fire i

detectors are based on the recommendations of NFPA 72E - 1990 Edition.

3/4.3.7.10 LOOSE PART DETECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system j

and avoid or mitigate damage to primary system components.

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

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

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

3/4.3.7.12 0FFGAS MONITORING INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.

3/4.3.8.

TURBINE OVERSPEED PROTECTION SYSTEM i

This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.

Protection from turbine excessive overspeed is required since excessive overspeed of the i

turbine could generate potentially damaging missiles which could impact and i

damage safety related components, equipment or structures.

~

3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTVATION INSTRUMENTATION i

The feedwater/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system i

in the event of failure of feedwater controller under maximum demand.

i 1

LIMERICK - UNIT 1 B 3/4 3-7 Amendment No. 33, 43, 70

Wide Arnee Leval This indication is reactor coolant temperature sensitive.

g calibration is thus made at rated conditions.

The The level error at low pressures (temperatures) is bounded by the safety analysis g

which reflects the weight of-coolant above the lower tap, and not g

indicated level.

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E UNITED STATES 5'

NUCLEAR REGULATORY COMMISSION k.....,"/

WASHINGTON, D.C. 20555-0001 PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 33 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 May 6, 1993, as supplemented by letter dated April 18, 1994, 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; 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; D.

The issuance of this amendmer' will not be inimical to the common defense and security or to t'.

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 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 Facility 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.

33, are hereby incorporated into this license.

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

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION b

Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 26, 1994 i

e 9ffy.

.i.y-,.

, - y

ATTACHMENT TO LICENSE AMENDMENT NO. 33 l

5 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 Replace the following pages of the Appendix A Technical Specifications with the attached 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 3/4 1-5 3/4 1-5 3/4 1-6 3/4 1-6*

3/4 3-41 3/4 3-41*

3/4 3-42 3/4 3-42 3/4 3-43 3/4 3_-43 3/4 3-44 3/4 3-44" 3/4 3-45 3/4 3-45 3/4 3-46 3/4 3-46 3/4 3-47 3/4 3-47*

3/4 3-48 3/4 3-48 3/4 3-51 3/4 3-51 3/4 3-52 3/4 3-52*

3/4 3-57 3/4 3-57 3/4 3-58 3/4 3-58*

3/4 3-59 3/4 3-59 m

w 3/4 3-65 3/4 3-65*

3/4 3-66 3/4 3-66 3/4 3-89 3/4 3-89*

3/4 3-90 3/4 3-90 3/4 3-111 3/4 3-111*

3/4 3-112 3/4 3-112 3/4 3-115 3/4 3-115 3/4 4-7 3/4 4-7 3/4 4-8 3/4 4-8*

..I

l l

l ATTACHMENT TO LICENSE AMENDMENT N0. 33 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353.

Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified.by Amendment number and l

contain vertical lines indicating the area of change. Overleaf.pages are provided to maintain document completeness.*

1 l

l Remove Insert l

i 3/4 6-13 3/4 6-13*

j 3/4 6-14 3/4 6 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 8 3/4 3-5 B 3/4 3-6 B 3/4 3-6 8 3/4 3-7 B 3/4 3-7 B 3/4 3-8 8 3/4 3-8*

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REACTIVITY CONTROL SYSTEMS SURVEILLANCEREQUIREMENTSJContinued) 4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by demonstrating:

a.

The scram discharge volume drain and vent valves OPERABLE, when control rods are scram tested from a nonnal control rod configura-tion of less than or equal to 50% R00 DENSITY at least once per 18 months, by verifying that the drain and vent valves:

1.

Close within 30 seconds after receipt of a signal for control rods to scram, and 2.

Open when the scram signal is reset.

b.

Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod block level instrumentation at least once per 92 days.

l 1

l T.

i l

1

~

LIMERICK - UNIT 2 3/4 1-5 Amendment No.33

REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 5, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a.

With the maximum scram insertion time of one or more control rods exceeding 7 seconds:

1.

Declare the control rod (s) with the slow insertion time inoperable, and 2.

Perform the Surveillance Requirements of Specification 4.1.3.2c.

at least'once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demon-strated through measurement with reactor coolant pressure great'er than or equal to'950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:

4 a.

For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER following CORE ALTERATIONS or after a reactor shutdown that is greater than 120 days.

4 b.

For specifically affected individual control rods following maintenance on or modification to the control rod or' control rod drive system which*could affect the scram insertion time of those specific control rods, and c.

For at least 10% of the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION.

LIMERICK - UNIT 2 3/4 1-6

l I

1ABLE 4.5

.-l (Continued)

EM[RGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS e

h

=

CHANNEL OPERA 110NAL U

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 7

TRIP FUNCTION CHECK TEST.

CALIBRATION SURVEILLANCE REQUIRED l

I 4.

AUTOMATIC DEPRESSURIZATION SYSTEM a.

Reactor Vessel Water Level -

N Low Low Low, Level 1 S

Q R

1, 2, 3 l b.

Drywell Pressure - High 5

Q R

1, 2, 3 c.

ADS Timer N.A.

Q Q

1, 2, 3 d.

Core Spray Pump Discharge Pressure - High 5

Q R

1, 2, 3 e.

RHR LPCI Mode Pump Discharge Pressure - High S

Q R

1, 2, 3 f.

Reactor Vessel Water Level - Low, Level 3 S

Q R

1, 2, 3 I

g.

Manual Initiation M.A.

R N.A.

1, 2, 3 g

h.

ADS Drywell Pressure Bypass Timer M.A.

Q Q

1, 2, 3 s &

y 5.

LOSS OF POWER w

k~

4.16 kV Emergency Bus Undery, a.

q voltage (Loss of Voltage)

N.A.

R N.A.

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

b.

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

S M

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

6{ g 2

!l When the system is required to be OPERA 8LE per Specification 3.5.2.

Required OPERA 8LE when ESF equipment is required to be OPERA 8LE.

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

54

Not required to be OPERABLE when reactor steam done pressure is lu s than or equal to 100 psig.

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

?

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 without scram recirculation pump trip (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1 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 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 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l 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 bo inoperable channels in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 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 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, restore at liast one trip system to OPERABLE status within I hour 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 2 3/4 3-42 Amendment No.33

-=

TABLE 3.3.4.1-1 ATWS RECISsy!ATION PUMP TRIP SYSTEM INSTRUMENTATION f

MINIMUM OPERABLE CHANNELS PER TRIP FUNCTION TRIP SYSTEM

1.

Reactor Vessel Water Level -

Low Low, Level 2 2

2.

Reactor Vessel Pressure - High 2

1$

'kt f

e l

l

One 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 required surveillance provided the other channel is OPERABLE.

LIMERICK - UNIT 2 3/4 3-43 Amendment No. 33 l

TABLE 3.3.4.1-2 AWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE TRIP FUNCTION SETPOINT VALUE 1.

Reactor Vessel, Water Level -

^

Low Low, Level 2

> -38 inches *

> -45 inches 2.

Reactor Vessel Pressure - High

< 1093 psig 5,1108 psig

=.

i

See Bases Figure B3/4.3-1.

LIMERICK - UNIT 2 3/4 3-44

)

i 4

TABLE 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE R'EQUIREMENTS i

l CHANNEL-CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION..

CHECK TEST CALIBRATION 1.

Reactor Vessel Water Level -

Low Low, Level 2 S

Q R

l 2.

Reactor Vessel Pressure -

l High S

Q R

l

l 4

...e.

I l

.1_

1 i

4 1

LIMERICK - UNIT 2 3/4 3-45 Amendment No. 33 l

INSTRUMENTATION END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEN INSTRUMENTATION LIMITING CONDITION FOR OPERATION i

3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system-instrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.

ACTION:

With an end-of-cycle recirculation pump trip system instrumentation a.

channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel 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 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l With the number of OPERABLE channels two or more less than required c.

by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:

1.

If the -inoperable channels consist of one turbine control valve l channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l 2.

If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.

d.

With one trip system inoperable, restore the inoperable trip system to OPERAB E status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or take" the ACTION required by Specification 3.2.3.

e.

With both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or take the ACTION required by Specification 3.2.3.

5 LIMERICK - UNIT 2 3/4 3-46 Amendment No.33

j INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

4.3.4.2.1 Each end-of-cycle recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.2.1-1.

4.3.4.2.2.

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

4.3.4.2.3 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function shown in Table 3.3.4.2-3 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least the logic of one type of channel input, turbine control valve fast closure or turbine stop valve closure, such that both types of channel inputs are tested at least once per 36 months.

The measured time shall be added to the most recent breaker arc suppression time and the resulting END-OF-CYCLE-RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be verified to be within its, limit.

4.3.4.2.4 The time interval necessary for breaker arc suppression from energi-zation of the recirculation pump circuit breaker trip coil shall be measured at least once per 60 months.

LIMERICK - UNIT 2 3/4 3-47

TABLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATIO'N T

MINIMUM i

OPERABLE CHANNELS PER TRIP SYSTEM

TRIP FUNCTION 4

1.

Turbine Stop Valve - Closure 2**

l 2.

Turbine Control Valve-Fast Closure 2**

i l

l l

i

A trip system 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 required surveillance provided that the other trip system is OPERABLE.

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

i LIMERICK - UNIT 2 3/4 3-48 Amendment No. 33 I

TABLE 4.3.4.2.1-1 I

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REOUIREMENTS

~

CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION TEST CALIBRATION 1

l l

1.

Turbine Stop Valve-Closure Q*

R l

2.

Turbine Control Valve-Fast Closure Q*

R

?

4b I l

)

i

\\

Including trip system logic testing.

. Tz LIMERICK - UNIT 2 3/4 3-51 Amendment No. 33

4 INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5 The reactor core isolation cooling (RCIC) system actuation instrumentation channels shown in Table 3.3.5-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Satpoint.

column of Table 3.3.5-2.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam done pressure greater than 150 psig.

ACTION:

a.

With a RCIC system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.5-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

I b.

With one or more RCIC system actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.5-1.

SURVEILLANCE REQUIREMENTS 4.3.5.1 Each RCIC system actuation instrumentation channel shall be demonstrated SPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.5.1-1.

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

n 4

A LIMERICK - UNIT 2 3/4 3-52

~.

.= -

INSTRUMENTATION 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.

The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.

APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

a.

With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of-Table 3.3.6-2, 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 Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE RE0VIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE

by the performance of l

the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1;

- 22 i

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

I LIMERICK - UNIT 2 3/4 3-57 Amendment No. 33 i

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

TABLE 3.3.6-1 h

CONTROL R00 BLOCK INSTRUMENTATION MINIMUM APPLICABLE p

OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION g

1.

ROD BLOCK MONITOR (a) y a.

Upscale 2

1*

60 b.

Inoperative 2

1*

60 m

c.

Downscale 2

1*

60 2.

APRM a.

Flow Biased Neutron Flux -

Upscale 4

1 61 I7) b.

Inoperative 4

1, 2, 5 61 l

c.

Downscale 4

1 61 II) d.

Neutron Flux - Upscale, Startup 4

2, S 61 R 3.

SOURCE RANGE MONITORS ***

[

Detector not full in(b) 3 2

61 a.

J, 2

5 61 b.

Upscale (c) 3 2

61 2

5 61 Inoperative (c) j c.

d.

Downscale(d) g u l 4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in 6

2, 5 61 g(E b.

Upscale 6

2, 5 61 Inoperati 6

2, 5 61 c.

Downscale{g) l 6

2, 5 61 d.

5.

SCRAM DISCHARGE VOLUME a.

Water Level-High 2

1, 2, 5**

62 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale 2

1 62 b.

Inoperative 2

1 62 c.

Comparator 2

1 62

-l 7.

^ 7CTOR MODE SWITCN SHUTDOWN POSITION 2

3, 4 63

TABLE 3.3.6-1 (Ccntinued)

CONTROL ROD-WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS Declare the RBM inoperable. and take the ACTION required by ACTION 60 Specification 3.I.4.3.

ACTION 61 With the number of OPERABLE Channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE 4

status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or place the inoperable channel in the tripped condition.

b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in j

the tripped condition within one hour.

With the number of OPERABLE channels less than required by the

~

ACTION 62 Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l With the number of OPERABLE channels less than required by the ACTION 63 Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block.

HQlfl With THERMAL POWER 2: 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.

These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.

4 (a)

The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less tSn 30% of RATED THERMAL POWER.

(b)

This function shall be automatically bypassed if deteEtor count rate is

> 100 cps or the IRM channels are on range 3 or higher.

(c)

This function is automatically bypassed when the associated IRM channels are on range 8 or higher.

(d)

This function is automatically bypassed when the IRM channels are on range 3 or higher.

(e)

This function is automatically bypassed when the IRM channels are on range 1.

(f)

Required to be OPERABLE only prior to and during shutdown margin -

demonstrations as performed per Specification 3.10.3.

4 LIMERICK - UNIT 2 3/4 3-59 Amendment No. 7, 33

a_

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u a se-

,__J,xp.

J h

l 9

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+

1

-P i

I

tn.

1

%=

l l

l 1

4 I

I l

l l

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i k

TABLE 3.3.7.1-1 (Continued) i RADIATION MONITORING INSTRUMENTATION 4

TABLE NOTATIONS l

.a "When irradiated fuel is.being handled in the secondary containment.,-

(a) With fuel in the spent fuel storage pool.

j (b) Alarm only.

ACTION STATEMENTS l

ACTION 70 With one monitor inoperable, restore the inoperable monitor to the OPERA 8LE status within 7. days or, within the next,6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , l

initiate and maintain operation.of the,contrel room emergency.

filtration system in the radiation isolation mode of operation.

1 With two or more of the monitors _. inoperable, within one hour, initiate and maintain operation of. the control room emergency i

filtration system in the radiation mode of operation.

.-c ACTION 71 With one of the required monitor inoperable,. assure 's portable continuous monitor with the'same' alarm'setpoint is OPERA 8LE in the vicinity of the installed monitor during any fuel movement.

l If.noifdel movement is being made, perform area surveys.of the monitored area with portable monitoring instrumentation at least once per.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i ACTION 72 With the required monitor inoperable, obtain and analyze at least one grab sample of'the monitored parameter at least once j

per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3 r

i ACTION 73 With the required monitor inoperable, assure a portable alarming

{

monitor is OPERA 8LE in the ' vicinity of the installed monitor or perform area surveys of the monitored area with portable monitor-ing instrumentation at least~once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l

.<_m 4

e i

i t

4 LIMERICK - UNIT 2 3/4 3-65

TABLE 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS OPERATIONAL CHANNEL CONDITIONS FOR CHANNEL FUNCTIONAL CHANNEL WHICH SURVEILLANCE

' INSTRUMENTATION CHECK TEST CALIBRATION IS REQUIRED

1. Main Control Room Normal Fresh Air Supply Radiation S

Q R

1; 2, 3, 5 and

l Monitor

2. Area Monitors

a. Criticality Monitors S

M R

(a) 1)

Spent Fuel Storage Pool S

M R

At All Times

b. Control Room Direct Radiation Monitor

3. Reactor Enclosure Cooling S

M R(b)

~At All Times, W2tsr Radiation Monitor I I l

i b

e e

INSTRUMENTATION TRAVERSING IN-CORE PROBE SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.7 The traversing in-core probe system shall be OPERABLE with:

~

a.

Five movable detectors, drives and readout equipment to map the core, and b.

Indexing equipment to allow all five detectors to be calibrated in a common location.

APPLICABILITY: When the traversing in-core probe is used for:

a.

Recalibration of the LPRM detectors, ano b.*

Monitoring the APLHGR, LHGR, MCPR, or MFLPD.

ACTION:

With the traversing in-core probe system inoperable, suspend use of the system for the above applicable monitoring or calibration functions. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.7 The traversing in-core probe system shall be demonstrated OPERABLE by normalizing each of the above required detector outputs within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to use for the LPRM calibration function.

.1

0nly the detector (s) in the required reasurement location (s) are required to be OPERABLE.

LIMERICK - UNIT 2 3/4 3-89

INSTRUMENTATION CHLORINE DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.8.1 Two independent chlorine detection system subsystems shall be OPERABLE with their alarm and trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 0.5 ppm APPLICABILITY:

All OPERATIONAL CONDITIONS.

ACTION:

a.

With one chlorine detection subsystem inoperable,. restore the inoperable detection system to OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , initiate and maintain operation of at least one control room emergency filtration system subsystem in the chlorine isolation mode of operation.

b.

With both chlorine detection subsystems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of at least one control room emer-gency filtration system subsystem in the chlorine isolation mode of operation.

SURVEILLANCE REOUIREMENTS 4.3.7.8.1 Each of the above required chlorine detection system subsystems shall be demonstrated OPERABLE by performance of a:

a.

CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

CHANNEL FUNCTIONAL TEST at least once per 92 days, and c.

CHANNEL CALIBRATION at least once per 18 months.

LIMERICK - UNIT 2 3/4 3-90 Amendment No.33

.. - - =..

l INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days by:

1.

Cycling each of the'following valves through at least one complete cycle from the running position:

a)

For the overspeed protection control, system; 1)

Four high pressure turbine control valves b)

For the electrical overspeed trip system and the mechanical overspeed trip system; 1)

Four high~ pressure' turbine control valves

~

c.

At least once per 18 months by performance of'a CHANNEL CALIBRATION of the turbine overspeed protection instrumentation.

d.

At least once per 40 months by disassembling at least'one of each of the above valves and performing a visual and surface inspection of all valve seats, disks and stems and verifying no unacceptable flaws or excessive corrosion.

If unacceptable flaws or excessive corrosion are found, all.other valves of that type s, hall be. inspected.

"I!_

LIMERICK - UNIT 2 3/4 3-111 t

INSTRUMENTATION 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.9 The feedwater/ main turbine trip system actuation instrumentation channels shown in the Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.9-2.

APPLICABILUy;.

As shown in Table 3.3.9-1..

EllM.i.

a.

With a feedwater/ main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.9-2, declare the channel inoper-able and either place the inoperable channel in the tripped condition.

until the channel is restored.to OPERABLE status with its trip set-point adjusted consistent with the Trip Setpoint value, or declare the associated system inoperable.

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days 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 .

c.

With the number of OPERABLE channels two less than required by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to OPERABLE : stat.u: 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 .

SURVEILLANCE RE0VIREMENTS 4.3.9.1 Each feedwater/ main turbine trip system actuation instrumentation channel shall be demonstrated OPERABLE

by the performance of the CHANNEL CHECK, CHANNEL l

FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.9.1-1.

4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 monthst

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 required surveillance without placing the trip system in the tripped condition.

LIMERICK - UNIT 2 3/4 3-112 Amendment No. 33

TABLE 4.3.9.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE RE0VIREMENTS OPERATIONAL CONDITIONS CHANNEL FOR WHICH CHANNEL FUNCTIONAL CHANNEL SURVEILLANCE TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED 1.

Reactor Vessel Water D

Q R

1 Level-High. Level 8

~~

LIMERICK - UNIT 2 3/4 3-115 Amendment No. 33

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 11 of the following reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*f 4

safet relief valves 91130 g i 1%

5 safet relief valves 91140 g i 1%

i 5

safet relief valves 91150 g i 1%

)

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

4 ACTION:

a.

With the safety valvo function of one or more of the above required safetv/ relief valves inoperable, be in at least HOT SHUTDOWN 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 .

b.

With one or more safety / relief valves stuck open pool average water te erature is less than 105 f,provided that suppression close the stuck open safety / relief valve minutes or if su) pre (s) if unable to close the stuck open valve ss on pool average water temperature is 110(s) within 2

F or greater, place tie reactor mode switch in the Shutdown position.

c.

With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REOUIREMENTS 4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.20 of the full open noise leve1## by performance of a:

a.

CHANNEL FUNCTIONAL TEST at least once per 92 days, and a b.

CHANNEL CALIBRATION at least once per 18 months **.

l 4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 24 months, and they shall be rotated such that all 14 safety relief valves are removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 54 months.

The lift setting pressure shall correspond to ambient conditions of the o

valves at nominal operating temperatures and pressures.

The provisions of Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling.

ff Initial setting shall be in accordance with the manufacturer's recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

LIMERICK - UNIT 2 3/4 4-7 Amendment No. 21,33

^

{-

I 3

REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR C0OLANT SYSTEM LEAKAGE l

LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION i

3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE:

The primary containment atmosphere gaseous radioactivity monitoring a.

j

system, b.

The drywell floor drain sump and drywell equipment drain tank flow j

monitoring system, j

The drywell unit coolers condensate flow rate monitoring system, and c.

j d.

The primary cont.nment Wess'."e and temperature monitoring system.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.*

1 ACTION:

I With only three of the above required leakage detection systems OPERABLE, i

operation may continue for up to 30 days provided grab samples of the contain-

{

nent atmosphere are obta.ined and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the i

required gaseous radioactive monitoring system, primary containment pressure and j

temperature monitoring system and/or the drywell unit coolers condensate flow rate monitoring system is inoperable; otherwise, be in at least HOT SHUTDOWN l

within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

~

j SURVEILLANCE REQUIREMENTS s 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE t,y:

1 Primary containment atmosphere gaseous radioactivity monitoring a.

systems performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a 1

CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL j

CALIBRATION at least once per 18 months.

b.

The primary containment pressure shall be monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and the primary containment temperature shall be monitored at j

least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

Drywell floor drain sump and Drywell equipment drain tank flow monitor-ing system performance of a CHANNEL FUNCTIONAL TEST at least once per 1

31 days and a CHANNEL CALIBRATION TEST at least once per 18 months.

I d.

Drywell unit coolers condensate flow rate monitoring system,

j performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

l-

The primary containment atmosphere gaseous radioactivity monitor is not j

required to be OPERABLE until OPERATIONAL CONDITION 2.

LIMERICK - UNIT 2 3/4 4-8 i

I

0 1-4 1

i i

CONTAI N NT SYSTEMS i

LINITING CONDITION FOR OPERATION (Continued) i ACTION:

(Continued) 3.

With the suppression chamber average water temperature greater i

than 120'F, depressurize the reactor pressure vessel to less l

than 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

With only one suppression chamber water level indicator OPERABLE and/or with less than eight suppression pool water temperature indicators, I

one in each of the eight locations OPERABLE, restore the inoperable i

indicator (s) to OPERABLE status within 7 days or verify suppression chamber water level and/or temperature to be within the limits at least

{

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

1

.d.

With no suppression chamber water level indicators OPERABLE and/or with 1

less than seven suppression pool water temperature indicators covering i

at least seven locations OPERA 8LE, restore at least one water level j

indicator and at least seven water temperature indicators to OPERABLE i

status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i e.

With the drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to 4

increasing reactor coolant temperature above 200*F.

i SURVEILLANCE REQUIREMENTS t

4 4.6.2.1 The suppression chamber.shall be demonstrated OPERABLE:

a.

By verifying the suppression chamber water volume to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the suppression chamber j

average water temperature to be less than or equal to 95*F, except:

1.

At least once per 5 minutes during testing which adds heat to J

the suppression chamber, by verifying the suppression chamber j

average water temperature less than or equal to 105'F.

j 2.

At least once per hour when suppression chamber _ average water j

temperature is greater than or equal to 95'F, by verifying:

i a)

Suppression chamber average water temperature to be les than or equal to 110'F, and I

b)

THERMAL POWER to be less than or equal to 1% of RATED THERMAL POWER 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after suppression chamber average water j

temperature has exceeded 95'F for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l 3.

At least once per 30 minutes following a scram with suppression i

chamber average water temperature greater than or equal to 95'F.

by verifying suppression chamber average water temperature less than or equal to 120'F.

LIMERICK - UNIT 2 3/4 6-13

... ~. -.

CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) c.

By verifying at least 8 suppression pool water temperature indicators in j

at least 8 locations, OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and 3.

CHANNEL CALIBRATION at least once per 18 months, with the temperature alarm setpoint for:

l 1.

High water temperature:

l a)

First setpoint s 95*F b)

Second setpoint s 105'F c)

Third setpoint s 110*F d)

Fourth setpoint s 120*F d.

By verifying at least two suppression chamber water level indicators OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 2.

CHANNEL FUNCTIONAL TEST at least once per 92 days, and 3.

CHANNEL CALIBRATION at least once per 18 months, with the water level alarm setpoint for high water level s 24'l-1/2" e.

Drywell-to-suppression chamber bypass leak tests shall be conducted at l

40 +/- 10 month intervals to coincide with the ILRT at an initial differential I

pressure of 4 psi and verifying that the A/(k calculated from the measured leakage is within the specified limit.

If any drywell-to-suppression chamber bypass leak test fails to meet the specified limit.. the test schedule for subsequent tests shall be reviewed and approved by the Commission.

If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed.

f.

By conducting a leakage test on the drywell-to-suppression chamber l

vacuum breakers at a differential pressure of at least 4.0 psi and verifying that the total leakage area A/(k contributed by all vacuum breakers is less than or equal to 24% of the specified limit and the leakage area for an individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during i

each refueling outage for which the drywell-to-suppression chamber bypass leak test in Specification 4.6.2.1.d is not conducted.

LIMERICK - UNIT 2 3/4 6-14 Amendment No. 3I,33

INSTRUMENTATION l

BASES t

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION (Continued)_

Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30936P, Parts 1 and 2, " Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation)," as approved by the NRC and documented in the SER (letter to D.

N. Grace from A. C. Thadant dated December 9, 1988 (Part 1) and letter to D. N.

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

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety 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 NED0-10349, dated March 1971, NEDO-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 1

E0C-RPT will reduce the likelihood of reactor vessel level decreasing to level 2.

Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in 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 EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one E0C-RPT system; a position switch from each of the other two stop valves provides input to the other EOC-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.

LIMERICK - UNIT 2 B 3/4 3-3 Amendment No. 17, 32, 33

INSTRUMENTATION BASES Specified surveillance intervals and maintenance outage times have been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Ou' of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D. Binz, IV, from C.E. Rossi dated July 21,1992).

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.

3/4.3.5 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 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 ep?cifically allocated for each trip in the safety analyses.

3/4.3.6 CONTROL ROD 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 outahi time 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.

1 LIMERICK - UNIT 2 B 3/4 3-4 Amendment No.11, //,33

1

=

INSTRUMENTATION FASES

' /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 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 infomation is available on selected plant parameters to monitor and asses these variable following an accident. This capability _is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.

The specified surveillance interval for the Main Control Room Normal Fresh Air Supply Radiation Monitor has been determined in accordance with GENE-770-06-1,

" Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D. Binz, IV, from C.E. Rossi dated July 21,1992).

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 that used in the design basis for the unit.

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

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 SHUTDOWN of the unit from locations outside of the control t

)

This capability is required in the event control room habitability is room.

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 connon to both units.

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 asses important variables following an accident. this capability is consistent with the recommendations of Regulaton 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 THI Action Plan Requirements," November 1930.

3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with infomation 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 infomation is available without the SRMs and they can be retracted.

LIMERICK - UNIT 2 B 3/4 3-5 Amendment No. //.17,33

~

INSTRUMENTATION BASES 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 obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core.

j 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).

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 ventilating system will automatically be placed in the chlorine isolation mode of operation to provide the req'uired 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.

Specified surveillance intervals and maintenance outage times have been determined in accordance with GENE-770-06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," as approved by the NRC and documented in the SER (letter to R.D.

Binz, IV, from C.E. Rossi dated July 21,1992).

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 fire suppression systems, that are actuated by fire detectors, will discharge extin-quishing agent in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an ir.tegral 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 OPERABLE fire detectors must be greater.

LIMERICK - UNIT 2 B 3/4 3-6 Amendment No. 17, 25 33

. =~.

,- INSrTRUMENTATION 4

l BASES 1

l 3/4.3.7.9 FIRE DETECTION INSTRUMENTATION (Continued) 1 i

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 establishment of a fire watch patrol must be initi-ated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire-patrols j

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

1 1

The surveillance requirements for demonstrating the OPERABILITY of the fire detectors are based on the recommendations of NFPA 72E - 1990 Edition.

4 3/4.3.7.10 LOOSE PART DETECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient

~

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-of-service times and surveillance requirements are consistent with the recom-1 mendations 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 THE 00CM.

l 3/4.3.7.12 0FFGAS MONITORING INSTRUMENTATION 4

1 This instrumentation includes provisions for monitoring the concentrations

~

of potentially explosive gas mixtures and noble gases in the off-gas system.

3/4.3.8.

TURBINE OVERSPEED PROTECTION SYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the

[

turbine could generate potentially damaging missiles which could. impact and i

damage safety related components, equipment or structures.

~

3/4.3.9 FEE 0 WATER /MAINTURBINETRIPSYSTEMACTUATIONINSTRUMENTTI0k The feedwater/ main turbine trip system actuation instrumentation is provided to initiate action of the feedwater system / main turbine trip system in the event of failure of feedwater controller under maximum demand.

2 l

1 i

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l 1

LIMERICK - UNIT 2 83/43-7 Amendment No. II, 25,33

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