Text

Amend 40 to License NPF-39,making Several Administrative Changes to Unit 1 Tech Specs to Eliminate Differences Between Unit 1 & Unit 2 Tech Specs
	ML20043H874
Person / Time
Site:	Limerick
Issue date:	05/30/1990
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20043H875	List: ML20043H874; ML20043H876;
References
	NUDOCS 9006270004
	Download: ML20043H874 (44)
	v • d • e

-

h o

y* *eng'o UNITED $TATES

'g NUCLEAR REGULATORY COMMISSION n

i j

wAssiwovow. o. c. rosss

\\...,./

PHILADELPHIA ELECTRIC COMPANY DOCKET-NO. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT 10 FACILITY OPERATING LICENSE Amendment No. 40 License No. NPF-39 1.

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

A.

The application for amendment by Philadelphia Electric Company (thelicensee)datedMarch 20, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter 1; i

i B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; 7

C.

Thereisreasonableassurance(i)thattheactivitiesauthorizedby 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

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 l

of the Comission's refulations and all applicable requirements have been satisfied.

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. 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 L

through Amendment No. 40, are hereby incorporated into this license. Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

l Jgp2J8880 3888 ;g2 P

a o

4'4 2

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION f.( f b

o Walter R. Butler, Director Project Directorate 1 2 Division of Reactor Projects 1/11

Attachment:

Changes to the Technical Specifications Date of issuance: May 30, 1990 r

i

1 l

ATTACHMENT TO LICENSE AMENDMENT NO. 40 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 Replace the following pages of the Appendix A Technical Specifications with the attached page. 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 xiii xiii xiv xiv xxi xxi xxii xxii

3/4 3-15 3/4 3-15 3/4 3-16 3/4 3-16 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32*

3/4 3-67 3/4 3-67*

3/4 3-68 3/4 3-68 3/4 3-69 3/4 3 69*

3/4 3-70 3/4 3-70 3/4 3-71 3/4 3-71 3/4 3-72 3/4 3-72 3/4 6-49 3/4 6-49*

3/4 6-50 3/4 6-50 3/4 6-52 3/4 6-52 3/4 6-53 3/4 6-53*

3/4 6-54 3/4 6-54 3/4 7-1 3/4 7-1*

3/4 7-2 3/4 7-2 3/4 7-3 3/4 7-3 3/4 7-4 3/4 7-4 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6

l. ..

2-ATTACHMENT TO LICENSE AMENDMENT N0. 40, FACILITY OPERATING LICENSE NO NPF-39 DOCKET NO. 50-352 Remove Insert-3/4 7-7 3/4 7-7 3/4 7-8 3/4 7-8*

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

B 3/4 7-1 B 3/4 7-1*

B 3/4 7-la B 3/4 7-la B 3/4 8-1 B 3/4 B-1 B 3/4 B-2 B 3/4 8-2 6-15 6-15*

6-16 6-16 i

r l

1 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE CONTAINMENT SYSTEMS (Continued) 3/4.6.4 VACUUM RELIEF Suppression Chamber - Drywell Vacuum Breakers...........

3/4 6-44 3/4.6.5 SECONDARY CONTAINMENT Reactor Enclosure Secondary Containment Integrity.......

3/4 6-46 Refueling Area Secondary Containment Integrity..........

3/4 6-47 Reactor Enclosure Secondary Containment Automatic Isolation Va1ves........................................

3/4 6-48 Table 3.6.5.2.1-1 Reactor Enclosure Secondary Containment Ventilation System Automatic Isolation Va1ves..........................

3/4 6-49 Refueling Area Secondary Containment Automatic Isolation Va1ves........................................

3/4 6-50 Table 3.6.5.2.2-1 Refueling Area Secondary Contain-ment Ventilation System Automatic Isolation Valves................

3/4 6-51 Standby Gas Treatment System - Common System............

3/4 6-52 l

Reactor Enclosure Recirculation System..................

3/4 6-55 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Primary Containment Hydrogen Recombiner Systems.........

3/4 6-57 Drywell Hydrogen Mixing System..........................

3/4 6-58 Drywell and Suppression Chamber Oxygen Concentration....

3/4 6-59 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS Residual Heat Removal Service Water System - Common System..................................................

3/4 7-1 Emergency Service Water System - Common System..........

3/4 7-3 Ultimate Heat Sink......................................

3/4 7-5 LIMERICK - UNIT 1 xiii Amendment No. 27, 40 l'

t 4

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued) 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMHON SYSTEM..................................................

3/4 7-6 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM...................

3/4 7-9 3/4.7.4 SNUBBERS................................................

3/4 7-11 Figure 4.7.4 1 Sample Plan 2) For Snubber Functional Test....................

3/4 7 16 3/4.7.5 SEALED SOURCE CONTAMINATION.............................

3/4 7-17 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System...........................

3/4 7-19 Spray and/or Sprinkler Systems..........................

3/4 7-22 CO Systems.............................................

3/4 7-24 Halon Systems...........................................

3/4 7-25 l

Fire Hose Stations......................................

3/4 7-26 l

Table 3.7.6.5-1 Fire Hose Stations................

3/4 7-27 Yard Fire Hydrants and Hose Cart Houses.................

3/4 7-29 l

Table 3.7.6.6-1 Yard Fire Hydrants and Hose t

Cart Houses......................

3/4 7-30 l

3/4.7.7 FIRE RATED 4SSEMELIES...................................

3/4 7-31 3/4.8 ELECTRICAL P0hER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. Sources - Operating................................

3/4 8-1 Table 4.8.1.1.2-1 Diesel Generator Test Schedule........................

3/4 8-8 1

A.C. Sources -

Shutdown.................................

3/4 8-9 3/4.8.2 D.C. SOURCES 1

D.C. Sources - Operating................................

3/4 8-10 LIMERICK - UNIT 1 xiv Amendment No. 33, 40 l

1 INDEX BASES SECTION PAGE CONTAINMENT SYSTEMS (Continued) l t

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES...............

B 3/4 6-4

~

3/4.6.4 VACUUM RELIEF......................................

B 3/4 6-4 3/4.6.5 SECONDARY CONTAINMENT..............................

B 3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL.............

B 3/4 6-6 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS - COMMON SYSTEMS.............

B 3/4 7-1 3

3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -

COMMON SYSTEM......................................

B 3/4 7-la 1

3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM..............

B 3/4 7-la-l 3/4.7.4 SNUBBERS...........................................

B 3/4 7-2 1

3/4.7.5 SEALED SOURCE CONTAMINATION........................

B 3/4 7-3 i

l 3/4.7.6 FIRE SUPPRESSION SYSTEMS........................

B 3/4 7-4 3/4.7.7 FIRE RATED ASSEMBLIES..............................

B.3/4 7-4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and DNSITE POWER DISTRIBUTION SYSTEMS...............................

B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES............

B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH................................

B 3/4 9-1 3/4.9.2 INSTRUMENTATION....................................

B 3/4 9-1 3/4.9.3 CONTROL ROD P0SITION...............................

B 3/4 9-1 L

3/4.9.4 DECAY TIME.........................................

B 3/4 9-1 l

3/4.9.5 COMMUNICATIONS.....................................

B 3/4 9-1 LIMERICK - UNIT 1 xxi Amendment No. 27, 40 l~

~

1 l.

l

+

m t

ngs BASES SECTION g

REFUELING OPERATIONS (Continuto) 3/4.9.6

.RdFUELING PLATF0RM................................

B 3/4 9 2 3/4.9.7 CRANE TRAVEL ' SPENT FUEL STORAGE P00L............

B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE P00L.........

B 3/4 9-2 3/4.9.10 CONTROL R0D REM 0 VAL............................s,.

B 3/4 9-2 3/4.9.11 RESIOUAL NEAT REMOYAL AND COOLANT CIRCULATION.....

B 3/4 9-2 3/4.1C SPECIAL TEST EXCEPT!0Ni 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY.....................

B 3/4 10-1 3/4.'10.2 R00 WORTH MINIMIZER...............................

B 3/4 10-1 3/4.10.3' SHUTOOWN MARGIN DEMONSTRATIONS.................... B 3/4 10-1 3/4.10.4 RECIRCULATION 100PS...............................

B 3/4 10-1 3/4.10.5 0XYGEN CONCENTRATION..............................

B 3/4 10 1 3/4:10.5 TRAINING STARTUPS.................................

B 3/4 10 1 3/4.11 RADI0 ACTIVE EFFLUENTS-3/4.11.1 LIQUID EFFLUENTS Concentration.....................................

B 3/4 11-1

'00se..............................................

B 3/4 11-1 Liquid Radwaste Treatment Systes..................

B 3/4 11-2 Liquid Noldup Tanks...............................

B 3/4 11-2

-3/4.11.2 GALE 005 EFFLUENTS Dose Rate.........................................

S 3/4 11-2 Dose - Noble Gases................................

B 3/4 11-3 Dose - Icdine-131. Iodine-133. Tritius, and Radionuclides in Particulate Fora.................

B 3/4 11-3 Ventilation Exhaust Treatment Systes..............

B 3/4 11-4 i

LIMERICK - UNIT 1 xxii Amendment No. 17 MAR 2 y1989 l

l q

=-

y Jg

.A;

~_ ;- ~ '~-

_q-L -:-

.~

TABLE ~3.3.2-1:(Continued)

~

- r-J'_

E ISOLATION ACTUATION INSTRUNENTATION E

MININUM APPLICABLE.-

~

OPERATIONAL-

ISOLATIg)'()OPERABLECHANNELgy TRIP FUNCTION SIGNAL

PER TRIP SYSTEM CONDITION

-ACTION 5Q 7.

SECONDARY CONTAIIMENT ISOLATION -

9 a.

Reactor Vessel Water Level Low, Low - Level 2

'B 2

1, 2, 3

_25 b.

Drywell Pressure.- High H

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

[.

Exhaust Doct Radiation - High R-2 2.~ Refueling Area Unit 2 Ventilation 25

'l w1 Exhaust-Duct Radiation 'High R

2 d.

Reactor Enclosure Ventilation Exhaust Duct Radiation - High 5

2 1,2,3 25 e.

Outside Atmosphere To Reactor Enclosure a Pressure - Low U

1 1, 2, 3

25

' f.

Outside' Atmosphere To Refueling 25 Area A. Pressure - Low T

1 g.

Reactor Enclosure Manual Initiation MA 1

1, 2, 3 -

24 25-k h.

Refueling Area Manua1' Initiation MA 1

E 2a E.

8.*

l3

.~.-

~

L

P TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 -

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 SHUTbOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 21-Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT 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 .

ACTION 22 -

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

ACTION 23 -

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

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

ACTION 24 -

Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least HOT 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 .

ACTION 25 -

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

ACTION 26 -

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

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

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

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

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

(b) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the channel or trip system in the tripped condition provided at least one other OPERARLE G,iiniel hi the same trip system is monitoring that parameter.

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

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

(d) f(WCU systen inlet outboard isolation valve closes on SLCS "B" initiation.

RWCU syste.n inlet inboard isolation valve closes on SLCS "A" or SLCS "C" initiatirn.

LIMERICK - Oh!T 1 3/4 3-16 Amendment No. 23, 40 l

1.

'ABLE 4.3.2.1-1 (Continued) g g

9 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS M

CHANMEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E TRIP FUNCTION CHECK TEST CALIBRA]9N SURVEILLANCE REQUIRED e

7.

SECONDARY CONTAllWENT ISOLATION a.

Reactor Vessel Water Level Low, Low - Level 2 S

M R

1, 2, 3 b.

Drywell Pressure - High S

M R

1,2,3 c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High S

M R

2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S

M R

a d.

Reactor Enclosure Ventilation w

g Exhaust Duct Radiation - High 5

M R

1,2,3

=

e.

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

M Q

1, 2, 3 f.

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

M Q

g.

Reactor Enclosure Manual Initiation M.A.

R N.A.

1,2,3 h.

Refueling Area Manual Initiation N.A.

R N.A.

5

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

ALTERATIONS, or (3) during operations with a potential for draining the 7 actor vessel with the vessel head 6

removed and fuel in the vessel.

t(

m

- When not administrative 1y bypassed and/or when any turbine stop valve is open.

8

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

wI

'IIII

'J

!NATRUMENTA??0N 3/a.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION'!NSTRUMENTATION 1

ll

- w L!*!*!NG CONDITION FOR ODERATION 3: 3. 3 79e eme gency core cooling system (ECCS) actsation instrumentation enancels sno.n in Taole 3.3.3-1 snall be OPERABLE with their trio setDoints set coFsistent itn.the values snown-. in the Trip Setooint column of Taote 3.3.3 2 ano with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Taole 3.3.3 3.

APoLICABILITY:

As shown in Table 3.3.3 1.

ACTION:

With an ECCS actuation instrumentation channel trip netpoint less a.

conservative than the value shown in the Allowable Values column cf Table 3.3.3-2, declare the channel inoperable until the channel is restorea to OPERABLE status with its trio *ttpoint adjusted consistent with the Trip Setpoint value.

b; With one or more ECCS actuation instrumentation channels inoperable, toe the ACTION reovired by Table 3.3.3 1.

.With either ADS trip system subsystem inoperable, restore the c.

inoperable trip system to OPERABLE status within:

1.

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

2.

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

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

' SURVEILLANCC REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL-FUNCTIONAL TEST and CHANNEL'CALIBRAT!0N operations for the OPERATIONAL CONDITIONS and'at the frecuencies shown in Table 4.3.3.1-1.

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

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

Each test shall include at least one channel per trip system such that all channels

- are tested at least once every N times 18 months where N is the total number of reduncant channels in a specific ECCS trip system.

LIMERICK'- UNIT 1 3/4 3-32

_m.

TABLE 4.3.7.1 1 (Continued)

RADI ATION W)NITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS "When irradiated fuel is being handled in the secondary Containment.

(a) With f uel in the spent fuel storage pool, (b) The initial CHANNEL CALIBRATION shall be perforined using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.

These standards shall permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

LIMERICK - UNIT 1 3/4 3-67 j

1 I

d I

INSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION

' LIMITING CONDITION FOR OPERATION 3.3.7.2 The seismic monitoring instrumentation shown in Table 3.3.7.2-1*

shal1~be OPERABLE.

APPLICABILITY:

At-all times.

ACTION:

a.

With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next-10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.2.1 Each of the-above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNC-TIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.2-1.

-4.3.7.2.2 Each.of the above required seismic monitoring instruments which is accessible during power operation and which is actuated during a seismic event greater than or equal to 0.01g shall be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within 5 days following the seismic event.

Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion.

A'Special' Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum and resultant effect.upon unit features important to safety.

Each of the above seismic monitoring instruments which is actuated during a seismic event greatet than or equal to 0.01 g but is not accessible during power operation shall be restored to OPERABLE status and a CHANNEL CALIBRATION performed the next time Unit 1 enters OPERATIONAL CONDITION 4 or below... A supplemental report shall then be prepared and submitted to the Commission within 14 days pursuant to Specification 6.9.2 describing the additional data from these instruments.

Shared with Unit 2.

l LIMERICK - UNIT 1 3/4 3-68 Amendment No.11, 40 l

1

TABLE 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERA 8LE 1.

Triaxial Time History Accelerographs (T/A's) a.

Sensors 1)

XE-VA-102 Primary Containment 0 to 1 g i

Foundation (Loc.109-R15-177) 2)

XE-VA-103 containment Structure o to 1 g 1

(Diaphragm Slab) 3)

XE-VA-104 Reactor Enclosure O to 1 g 1

Foundation (Loc. 111-R11-177) 4)

XE-VA-105 Reactor Piping Support 0 to 1 g 1

(Mn. Sta. Line 'D', El 313',

in containment) 5)

XE-VA-106 Outside Containment 0 to 1 g 1

on Seismic Category I Equipment (RHR Heat Exchanger, Loc. 102-R15-177) 6)

XRSH-VA-1078 Foundation of an 0 to 1 g 1

Independent Seismic Category I Structure (Spray Pond Pump House, El 237')

b.

Recorders (Panel 000693) 1)

XR-VA-102 for XE-VA-102 N.A.

1 2)

XR-VA-103 for XE VA-103 N. A.

1 3)

XR-VA-104 for XE-VA-104 N. A.

1 4)

XR-VA-105 for XE-VA-105 N.A.

1 5)

XR-VA-106 for XE-VA-106 N.A.

I

" Includes sensor, trigger, recorder, and backup power supply.

LIMERICK - UNIT 1 3/4 3-69

TABLE 3.3.7.2-1 (Continued)

SEISMIC MONITORING INSTRUMENTATION

. MINIMUM.

MEASUREMENT-INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

(

c.

Triaxial: Seismic Trigger (S/T)

1)

XSH-VA-001 (Activates Items' NAt 1-1.b.1) thru 5) above);

(Loc. Area 16 El 177')

2.-

' Triaxial Peak Recording Accelerograph (P/A's)=

a.

XR-VA-151 Reactor Equipment 0-2g 1***

(Top of reactor l vessel head) b.

XR-VA-152 Reactor Piping 0-2g 1

(Mn. Stm. Line 'D,' El 313',

in containment) c.

XR-VA-153 Reactor Equipment Outside 0 -2g 1-Containment-(RHR Heat Exchanger, Loc. 203-R15-201) t 3.

Triaxial Seismic Switches a.

XSHH-VA-001 Primary Containment.

NAtt 1*

l Foundation (Loc. 118-R16-117)'

4.

Triaxial Response Spectrum Analyzer 1-33.5 Hz 1*, * *

(RSA); (Loc. Control Room)

[

t tThe triaxial-seismic trigger setpoint is 0.005 g ttThis switch triggers at 5,0.15 g horizontal and < 0.10 g vertical.

With reactor control room indication and annunciation.

- Receives signal from playback unit fed with data from the Triaxial Accelerographs, Item 1.a above.

- - Not required to be OPERABLE when reactor vessel head is removed.

l LIMERICK - UNIT 1 3/4 3-70 Amendment No. 14, 40 l

p'

19.,.

1, j,

.I

$.-TABLE 4.3.7 M-1--

"n

=

' SEISMIC' MONITORING INSTRUMENTATION SURVEILLANCE ~ REQUIREMENTS n

CHANNEL CHANNEL FUNCTIONAL CHANNEL i

INSTRUMENTS AND SENSOR' LOCATIONS

-CHFCK

-TEST CALIBRATION 1.

Triaxial Time-History Accelerographs (T/A's) 4 a.

Sensors

1) XE-VA-102. Primary Contain-N.A.

SA R

ment Foundation (Loc. 209-R15-177)

-m*

2)- XE-VA-103 Containment N.A.

SA R

3 Structure =(Diaphragm Slab) 3)-XE-VA-104 Reactor Enclosure N.A.

SA R

Foundation (Loc. 111-R11-177) l 2

4) XE-VA-105 Reactor Piping N.A.

SA R

=

Support (Mn. Stm Line 'D,'

El 313', in containment)

5) XE-VA-106 Outside Contain-N.A.

'SA R

ment on Seismic Category I i

Equipment, (RHR Heat Exchanger, Loc. 102-R15-177)

'I

6) XRSH-VA-107* Foundation of N.A.

SA R

.an Independent Seismic Category I' Structure (Spray Pond Pump. House, El 237')

b.

Recorders (Panel 000693) f.1 1)

XR-VA-102 for XE-VA-102 N.- A.

SA R

2)

XR-VA-103 for XE-VA-103 N.A.

SA R

q U

3).

XR-VA-104 for XE-VA-104 N.A.

SA R

\\'

4): XR-VA-105 for XE-VA-105 N.A.

SA R

5)'

XR-VA-106-for-XE-VA-106 N.A.

SA R

x> ;

Includes sensor, trigger, recorder, and backup power supply.

.i i

j

+

? LIMERICK - UNIT 1 3/4 3-71 Amendment No. 40 l

b;

i

,4 l'j TABLE 4.3.7.2-1 (Continued) o ll:. [

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL INSTRUMENTS AND SENSOR LOCATIONS CHECK TEST CALIBRATION c.

Triaxial Seismic Trigger (S/T) 1).

XSH-VA-001 (Activates N.A.

SA R'

l' Items 1.b.1) thru 5) above)

- 2.

Triaxial Peak Recording Accelerograph'(P/A's) a.

XR-VA-151 Reactor Equipment N.A.

N.A.

R (Top of reactor vessel head)

.b.-

XR-VA-152 ' Reactor Piping N. A.

N.A.

-R (Mn. Stm. Line ' D,' ' El 313',

in containment) m c.

-XR-VA-153 Reactor Equipment N.A.

N.A.

R Outside Containment (RHR Heat.

Exchanger, Loc. 203-R15-201)

- 3.

Triaxial Seismic Switches a.

XSHH-VA-001 Primary Containment N.A.

SA R

Foundation (Loc. 118-R16-177) 4.<

Triaxial Response Spectrum Analyzer N.A.

SA.

R (RSA) l i

LIMERICK - UNIT 1 3/4 3-72 Amendment No. 40 l

O

i r '~.s

' 's ?

~

7

(

TABLE 3.6.5.2.1 [,'I REACTOR ENCLOSURE SECONDARY CONTAINMENT VENTILATION' SYSTEM

+

. AUTOMATIC I5OLATION VALVE 5

.f

' REAATOR ENCLOSURE ~(ZONE I)

MAXIMUM" ISOLATION TIME-ISOLATI (Seconds)

SIGNALSg)-

' j(/;d TUNCTION

1.. Reactor Enclosure ventilation Supply Valve HV-76-107 5

-B,H,5,0 2.-

Reactor Enclosure Ventilation Supply Valve HV-76-108 5

B,H,5,0

-3.

Reactor Enclosure Ventilation Exhaust Valve HV-76-157-5 B,H,5,U 4.

Reactor Enclosure Ventilation Exhaust Valve HV-76-158 5

B,H.S U 5.

Reactor Enclosure Equipment Compartment Exhaust Valve HV-76-141 5

B,H,5,U 6.

Reactor Enclosure Equipment Compartment Exhaust Valve HV-76 142 5

B,H,5,U.

7.

Drywell Purge Exhaust Valve HV 76-030

_5 B,H,5,U R.T B.

Drywell Purge Exhaust valve HV-76-031 5

B,H,5,U,R,T

.9.

Drywell Purge Exhaust Inboard 5

B, H,5, U,)t, R, T Valve HV-57-214 (Unit 2) 10.

Dryuell Purge Exhaust Outboard 6

B,H,5,U,W,R,T Valve HV-57-215 (Unit 2) 11.

Suppression Pool Purge Exhaust 5

B, H, S, U,W, R, T:

Inboard. Valve HV-57-204 (Unit 2) 12.

Suppression Pool Purge Exhaust 6

B;H,5,U,W,R,i Outboard Valve HV-57-212 (Unit 2) s

(*)See Specification 3.3.2, Table 3.3.2-1, for isolation signals that operate each automatic valve.

LIMERICK - UN!T 1 3/4 6-49 Amendment No. 23 JUN 1 s 1989 Q(dst,det 0L.is tsswtA.tM 2

+

CONTAINMENT SYSTEMS e

REFUELING AREA SECONDARY CONTAINMENT! AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic

-isolation valves shown in Table 3.6.5.2.2-1 shall be OPERABLE with isolation times less thantor equal to the times shown in Table 3.6.5.2.2-1.

-APPLICABILITY:

OPERATIONAL CONDITION *.

ACTION:

'With one or more of the refueling area secondary containment ventilation system automatic isolation valves shown in Table 3.6.5.2.2-1 inoperable, maintain at least one isolation valva-OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

a.

Restore the inoperable valves to OPERABLE status, or b.

Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or c.

Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.

Otherwise, in OPERATIONAL CONDITION *, suspend handling of irradiated l

fuel in the refueling area secondary containment, CORE ALTERATIONS and operations with-a potential for draining the reactor vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shown in Table 3.6.5.2.2-1 shall be demonstrated OPERABLE:

a.

Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

b.

At least once per 18 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation

~

position.

c.

By verifying the isolation time to be within its limit at least once per 92 days.

\\

Required when (1) irradiated fuel is being handled in the refueling area secondary containment, or (2) during CORE ALTERATIONS, or (3) during opera-tions with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

LIMERICK - UNIT 1 3/4 6-50 Amendment No. 6, 40 l

,_.c e

- 1 CONTAINMENT SYSTEMS ~

-STANDBY GAS TREATMENT SYSTEM - COMMON SYSTEM l

LIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.

w APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and *,

ACTION:

a.-

With-one standby gas treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:

1.

In OPERATIONAL CONDITION 1, 2, or 3, 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 .

2.

In OPERATIONAL CONDITION *, suspend handling of irradiated fuel l

4 in the secondary containment,-CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

The provisions of Specification 3.0.3 are not applicable, b. --

With both standby gas treatment subsystems inoperable in OPERATIONAL CONDITION *, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS or operations with a potential for-draining the reactor vessel.

The provisions of Specification 3.0.3.

are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:

a.-

At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.

Required'when (1) irradiated fuel is being handled in the refueling area secondary containment, or (2) during CORE ALTERATIONS, or (3) during opera-tions with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.

l l

LIMERICK - UNIT 1 3/4 6-52 Amendment No. 29, 40

[

e e

B B

E

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

At least once per 18 months or (1) after any structural maintenance b.

on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:

1.

Verifying that the subsystem satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Positions C.S.a.

C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3000 cfm i 10%.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 0.175%; and 3.

Verify that when the fan is running the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones I and II) and 2200 cfm minimum from the refueling area (Zone III) when tested in accordance with ANSI N510-1980.

4.

Verify that the pressure drop across the refueling area to SGTS prefilter is less than 0.25 inches water gage while operating at a flow rate of 2400 cfm 2 10%.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying c.

within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 0.175%.

d.

At least once per 18 months by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 9.1 inches water gauge while operating the filter train at a flow rate of 8400 cfm t 10%.

LIMERICK - UNIT 1 3/4 6-53 AmendmentNo.k33 007 3 0 539

--i----

CONTA2NMENT SYSTEMS'-

c SURVEILLANCE REQUIREMENTS (Continued) 2.

Verifying that the fan starts and isolation valves necessary to draw a suction from the refueling area or the reactor enclosure recirculation discharge open on each of.the following test signals:

a)

Manual initiation fron the control room, and b)

Sikolated automatic initiation signal.

3.

Veriftin is > 15'g that the temperature differential across each heater F when-tested in accordance with ANSI N510-1980, e.

After each complete or partial replacement' of a HEPA' filter bank by' verifyirg.that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at'a flow rate of 3000'cfm i 10%.

f.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the'inplace penetration and leakage testing ' acceptance criteria of less than 0.05%

in accordance'_with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 3000 cfm i 10%.

g..

'After any major system alteration:

l l'

Verify that when the SGTS fan is running-the subsystem flowrate is 2800 cfm minimum from each reactor enclosure (Zones-I and II) and 2200 cfm minimum from the refueling area (Zone III).

2.

Verify that one standby gas treatment subsystem will drawdown reactor enclosure Zone I secondary containment to greater than or equal to 0.25 inch of vacuum water gauge in less than or equal to 121 seconds with the reactor enclosure recirculation system in operation and the adjacent reactor enclosure and refueling area zones are in their isolation modes.

LIMERICK - UNIT 1 3/4 6-54 Amendment No. 40 l

3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.1 At least the following independent residual heat removal service water (RHRSW) system subsystems, with each subsystem comprised of:

a.

Two OPERABLE RHRSW pumps, and E'

b.

An OPERABLE flow path capable of taking suction from the RHR service

[

water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water through one Unit 1 l

RHR heat exchanger, shall be OPERABLE:

In OPERATIONAL CONDITIONS 1, 2, and 3, two subsystems.

a.

b.

In OPERATIONAL CONDITIONS 4 and 5, the subsystem (s) associated with

=-

systems and components required OPERABLE by Specification 3.4.9.2, 3.9.11.1, and 3.9.11.2.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.

ACTION:

a.

In OPERATIONAL CONDITION 1, 2, or 3:

,1.

With one RHRSW pump inoperable, restore the inoperable pump to OPERABLE' status within 30 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 .

2.

With one RHRSW pump in each subsystem inoperable, restore at least one of the inoperable RHRSW pumps 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 .

3.

With one RHRSW subsystem otherwise inoperable, restore the inoperable subsystem to OPERABLE status with at least one OPERABLE RHRSW pump within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT L

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 .

4.

With both RHRSW subsystems otherwise inoperable, restore at least one subsystem to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 .

"Whenever both RHRSW subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as I

low as practical by use of alternate heat removal methods.

LIMERICK - UNIT 1 3/4 7-1 Amendment No. 27 M

2 0 1999 g@9 h OL is issmd.b M7-

.. El E I E

f b

3 y

i

~ PLANT-SYSTEMS

^

JF LIMITING CONDITION FOR OPERATION (Continued) w

' ACTION: -(Continued)-

5.

LWith'two RHRSW pump / diesel generator pairs

inoperable, restore at least one inoperable RHRSW pump / diesel generator pair

to j

OPERABLE status within 30 days, or. be in at least HOT, SHUTDOWN l1.

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

The provisions of specification.3.0.4 are not.

-applicable.

= n' n

6.

With three RHRSW pump / diesel generator pairs

inoperable, restore at least one inoperable RHRSW pump / diesel generator-pair

to OPERABLE status within 7 days, or 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 following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7.

With four RHRSW pump / diesel generator pairs

inoperable, restore at least one inoperable RHRSW pump / diesel generator pair

to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 , b.

In OPERATIONAL CONDITION 3 or 4 with the RHRSW subsystem (s), which is associated with an RHR loop required OPERABLE by Specification 3.4.9.1 or 3.4.9.2,11noperable, declare the associated RHR loop

. inoperable and take the ACTION required by Specification 3.4.9.1 or-3.4.9.2, as applicable.

- c.

In OPERATIONAL CONDITION 5 with the RHRSW subsystem (s), which 1.)

associated with an RHR loop required OPERABLE by Specification 3.9.11.1 or 3.9.11.2 inoperable, declare the associated RHR system inoperable and take the ACTION required by Specification'3.9.11.1

-or 3.9.11.2, as applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.11 At least the above required residual heat removal service water system subsystem (s) shall be demonstrated OPERAELE:

a.

At least once per 31 days by verifying that each valve in the flow path that is not locked, sealed, or otherwise secured in position, is' in its correct position.

t

A RHRSW pump / diesel generator pair consists of a RHRSW pump and its associated dieseligenerator.

If either a RHRSW pump or its associated diesel generator

-.becomes inoperable, then the RHRSW pump / diesel generator pair is inoperable.

l 2

LIMERICK - UNIT 1 3/4 7-2 Amendment No. 27, 40 l.

a

y

t. _-

l PLANT SYSTEMS 9

EMERGENCY = SERVICE WATER SYSTEM - CO R N SYSTEM l

LIMITING CONDITION FOR OPERATION i -

3.7.1'2 ~At least the following independent emergency service water system loops, with each loop comprised of:

a.

Two OPERABLE emergency service water pumps, and b.

An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray _ pond or the cooling tower basin and transferring the water to the associated Unit I and common safety-related equipment, shall be OPERABLE:

a.

In OPERATIONAL CONDITIONS 1, 2, and 3, two loops.

b.-

In OPERATIONAL CONDITIONS 4, 5, and *, one loop.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and *.

ACTION:

a.

In OPERATIONAL CONDITION 1, 2, or 3:

1.

With one emergency service water pump inoperable.. restore the inoperable pump to OPERABLE status within 45 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 .

2.

With one emergency service water pump in each loop inoperable, restore at-least one inoperable pump to OPERABLE status within 30-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 , 3.

With one emergency service water system loop otherwise inoperable, declare all equipment aligned to the inoperable loop inoperable **, restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-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 .

8

When handling irradiated fuel in the secondary containment.

- The diesel generators may be aligned to the OPERABLE emergency service water system loop provided confirmatory flow testing has been performed.

Those diesel generators not aligned to the OPERABLE emergency service water system loop shall be declared inoperable and the actions of 3.8.1.1 taken.

LIMERICK - UNIT 1 3/4 7-3 Amendment Nc. 27, 40 l

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued) l 4.

With three ESW pump / diesel generator pairs ** inoperable, restore at least one inoperable ESW pump / diesel generator pair ** 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 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 .

5.

With four ESW pump / diesel generator pairs ** inoperable, restore at least one inoperable ESW pump / diesel generator pair ** to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 , b.

In OPERATIONAL CONDITION 4 or 5:

1.

With only one emergency service water pump and its associated flowpath OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the associated safety related equipment inoperable and take the ACTION required by Specifications 3.5.2 and 3.8.1.2.

c.

In OPERATIONAL CONDITION

1.

With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at least one flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or verify adequate cooling remains available for the diesel generators required to be OPERABLE or declare the associated diesel generator (s) inoperable and take the ACTION required by Specification 3.8.1.2.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.2 At least the above required emergency service water system loop (s) shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) that is not locked, sealed, or otherwise secured in position, is in its correct position.

b.

At least once per 18 months by verifying that:

1.

Each automatic valve actuates to its correct position on its appropriate ESW pump start signal.

2.

Each pump starts automatically when its associated diesel generator starts.

"When handling irradiated fuel in the secondary containment.

- An ESW pump / diesel generator pair consists of an ESW pump and its associated diesel generator.

If either an ESW pump or its associated diesel generator becomes inoperable, then the ESW pump / diesel generator pair is inoperable.

l LIMERICK - UNIT 1 3/4 7-4 Amendment No. 29, 40 l

PLANT' SYSTEMS I

ULTIMATE HEAT SINK-LIMITING' CONDITION'FOR OPERATION 3.7.1.3 The spray pond shall be OPERABLE with:

a.

A minimum pond water level at or above elevation 250' 10" Mean Sea

. Level, and b.

A pond water temperature of less than or equal-to 88'F.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and *.

-ACTION:

With the requirements of the above specification not satisfied:

a.

In OPERATIONAL CONDITION 1, 2, or 3, 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.

In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and the emergency service water system inoperable end take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2.

c.

In OPERATIONAL CONDITION *, declare the emergency service water l

system inoperable and take the ACTION required by Specification

-3.7.1;2.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.3 'The spray pond shall be determined OPERABLE:

a.

By verifying the pond water level to be greater than its limit at least once per.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

By verifying the water surface temperature (within the upper two feet

-of the surface)-to be'less than or equal to 88'F:

1.

at least once -per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the spray pond temperature is' greater than or equal to 80'F; and

2. -

at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when the spray pond temperature is greater than or equal to 85'F; and 3.-

at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the spray pond temperature is greater than 32'F.

c.

By verifying all piping above the frost line is drained within I hour after being used.

When handling irradiated fuel in the secondary containment.

LIMERICK - U MT 1 3/4 7-5 Amendment No. 25, 40 l

o

e3 8

PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR-SUPPLY SYSTEM - COMMON SYSTEM

[

LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems shall be OPERABLE.

- APPLICABILITY:

All OPERATIONAL CONDITIONS and 4 ACTION:

a.

In OPERATIONAL CONDITION 1, 2, or 3 with one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem 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 , b.

In OPERATIONAL CONDITION 4, 5, or *:

1.

With one control room emergency fresh air supply subsystem inoperable, restore the inoperable subsystem to OPERABLE status within'7 days or initiate and maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.

2.

With both control room emergency fresh air supply subsystems

-inoperable, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

c.

The provisions of Specification 3.0.3 are not applicable'in OPERATIONAL CONDITION *.

l SURVEILLANCE REQUIREMENTS s 4.7.2 Each control room emergency fresh air supply subsystem shall be demonstrated OPERABLE:

Atleastonceper12hoursbyverifgingthecontrolroomairtempera-a.

ture to be less than or equal to 85 F effective temperature.

b.

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying-that the subsystem operates with the heaters OPERABLE.

c.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the subsystem by:

1.

Verifying that the subsystem satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.S.a. C.S.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3000 cfm + 10%.

When irradiated fuel is being handled in the secondary containment.

LIMERICK - UNIT 1 3/4 7-6 Amendment No. 40 l

4 1,ft, k

(,

L % 1/ PLANT SYSTEMS:

o c,..

n..-

j SURVEILLANCE REQUIREMENTS (Continued) 2.

Verifying within 31 days after removal that a laboratory analysis-of a representative ctrbon sample obtained in accordince with Regulatory Position t.6.b of Regulatory Guide 1.52,- Revision 2 March 1978, meets.tne laboratory testing criteria of Regulatory Position C.6.e of Aegulatory Guide 1.52, Revision 2 March 1978, for a methyl. iodide penetration of'less than 1%;_and 3.

Verifying a subsystem flow rate of 3000 cfm + 10% during ~

subsystem operation when tested in accordance with ANSI N510-1980.'

-d.

After every:720 hours of charcoal adsorber operation by verifying within 31 daysiafter removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%.

e.

At least.once per 18 months by:

1.

Verifying that the pressure-drop across the combined profilter, upstream and downstream HEPA filters, and charcoal adsorber banks is less than 6 inches-water gauge while operating the subsystem at a flow rate of 3000 cfm + 10%; verifying that the profilter pressure drop is less-than D 8 inch water gauge and that the pressure drop across each HEPA is less than 2 inches water gauge.

2.

'!erifying that or each of the below chlorine isolation mode actuation test signals, the subsystem automatically switches

to the chlorine isolation mode of operation and the isolation valves close within 5 seconds:

e a)

Outside air intake high chlorine,.and b)

Manual initiation from-the control room.

3.

Verifying that on each of the below. radiation isolation mode actuation test signals, the subsystem automatically switches-to the radiation isolation mode of operation and the control room is maintained at a positive pressure of at least 1/8 inch water gauge relative to the turbine encksure and auxiliary equipment room and outside atmosphere during subsystem operation with an outdoor air flow rate less than or equal to 525 cfm:

l a)

Outside air intake high radiation, and b)

Manual initiation from control room.

LIMERICK - UNIT 1 3/4 7-7 Amendment No. 5, 40 l

l 4

us

~

'Ot

+

t

' PLANT $YSTEMS

~

' $URVEILLANCE REQUIREMENTS (Continued) f.

. After each complete or partial replacement'of. a HEPA filter bank by verifying.that the HEPA filter bank satisfies-the inplace penetra-tion and bypass leakage testing acceptance criteria of less than 0.05%

o in accordance with ANSI N510-1980 while operating the system at a flow rate of 3000 cfm 1 10%.

5 g.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass' leakage-testing acceptance criteria

.of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refri0erant test gas while operating the system at a flow rate of-3000 cfm i 10%.

e I

LIMERICK - UNIT 1 3/4 7-8

w, f.',:

.(4 *

? -3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. - SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the.following A.C. electrical power sources shall be OPERABLE:

a.

Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and b.

_ Four separate and independent diesel generators, each with:

1.

A separate day tank containing a minimum of 200 gallons of fuel, 2.

A separate fuel storage system containing a minimum of 33,500-gallons of fuel, and 3.

A separate fuel transfer pump.

APPLICABILITY:

OPERATIONAL CONDITIONS 1,- 2, and 3.

ACTION:.

a.

With one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.

sources by performing Surveillance Requirement 4.8.1.1.la, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 7 days thereafter.

If the diesel

_ generator became inoperable for any reason other than preplanned preventative maintenance or testing, demonstrate,the 1-OPERABILITY of the remaining operable diesel generators by performing Surveillance Requirement 4.8.1.1.2a.4 for:one-diesel generator at a time, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

and at-least once per 7 days thereafter.

-Restore the inoperable diesel generator to OPERABLE status within 30 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 .

See also ACTION e.

LIMERICK - UNIT 1 3/4 8-2a Amendment No. 40 l

e

)

sc.

'.

REACTOR COOLANT SYSTEM BASES 3/4.4.3 REyTORCOOLANTSYSTEMLEAKAGE 3/4.4.3.1 L(AKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These detection systems are consistent with the recommendations of l

Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

In conformance with Regulatory Guide 1.45, the channel calibration tests will verify the ability to detect a 1 gpm leak ia less than I hour and an atmospheric gaseous radioactivity system sensitivit) of i

10 5 pC/cc.

3/4.4.3.2 OPERATIONAL LEAKAGE The allow:ble leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also con-sidered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located ed known to be PRESSURE BOUNDARY LEAKAGE, the reactor will N shutdown to allow Vether investigation and corrective action.

The ACTION requirements for pressure isolation valves (PIVs) are used in conjunction with the system specifications for which PIVs are listed in Table 3.4.3.2-1 and with primary containment isolation valve requirements to ensure s

that plant operation is appropriately limited.

The Surveillance Requirements for RCS press we isolation valven provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RC3 pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system arc istablished to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainicss steel.

The effect.of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

Conductivity measurements are required on a continuous basis since changes 1

in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affe.cting conductivity i

must also be within their acceptable limits. With the conductivity meter i

inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

LIMERICK UNIT 1 B 3/4 4-3 Amendment No. 40 l

1

- \\

i o

REACTOR COOLANT SYSTEM BASES 3/4.4.4 CHEMISTRY (Continued)

{

lhe surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective i

action, 1

3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed 1

small fractions of the dose guidelines of 10 CFR Part 100.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as SITE BOUNDARY location and meteoro-logical conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131, but less than or equal to 4 micro-curies per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THEPMAL POWER. Operation with specific activity levels exceeding 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4 microcuries per gram DOSE EQUIVALENT I-131 1

must be restricted since these activity levels increase the 2-hour thyroid dose at the SITE BOUNDARY following a postulated steam line rupture.

Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

3/4.4.6 PRES $URE/ TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation, n.

LIMERICK - UNIT 1 S 3/4 4-4 Amendment No. 20, 40

\\

i

. m -

2

t0NTAINMENT SYSTEMS BASES 3]4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident.

The Reactor Enclosure and associated structures provide secondary containment during normal operation i

when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.

Estabilshing and maintainirig a vacuum in thir reactor enclosure secondary containment with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERABILITY of the reactor enclosure recirculation system and the standby gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA or refueling accident (SGTS only).

The reduction in containment iodine inventory reduces the resulting SITE B0UNDARY radiation doses associated with containment leakage.

The operation of this.

system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA and refueling accident analyses.

Provisions have been made to continuously purge the filter plenums with instrument air when the filters are not in use to prevent buildup of moisture on the adsorbers and the HEPA filters.

Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 135 secor.;s, these surveillance require-ments specify a draw down time of 121 seconds.

This 14 second difference is due to the diesel generator starting and sequence loading delays which is not part of this surveillance requirement.

The reactor enclosure secondary containment draw down time analyses assumes a starting point of 0.25 inch of vacuum water gauge and worst case SGTS dirty filter flow rate of 2800 cfm.

The surveillance requirements satisfy this as-sumption by starting the drawdown from ambient conditions and connecting the adjacent reactor enclosure and refueling area to the SGTS to split the exhaust flow between the three zones and verifying a minimum flow rate of 2800 cfm from the test zone.

This simulates the worst case flow alignment and verifies ade-quate flow is available to drawdown the test zone within the required time.

The Technical Specification Surveillance Requirement 4.6.5.3.b.3 is intended to be a multi-zone air balance verification without isolating any test zone.

The SGTS fans are sized for three zones and therefore, when aligned to a single zone or two zones, will have excess capacity to more quickly drawdown the affected zones.

There is no maximum flow limit to individual zones or pairs of zones and the air balance and drawdown time are verified when all three zones are ennnected to the SGTS.

The three zone air balance verification and drawdown test will be done after any major system alteration, which is any modification which will have l

an effect on the SGTS flowrate such that the ability of the SGTS to drawdown the reactor enclosure to greater than or equal to 0.25 inch of vacuum water gage in less than or equal to 121 seconds could be affected.

LIMERICK - UNIT 1 B 3/4 6-5 Amendment No.6. 40 l

e

CONTAINMENT SYSTEMS SA$E$

3/a.6.5 $ECONDARY CONTAINMENT (Continued)

The field tests for bypass leakage across the SGTS charcoal adsorber and HEPA filter banks are perfomed at a flow rate of 3000 a 105 cfa.

i This flow rate corresponds to the maximum overall three zone inleakage rate of 3264 cfm.

The SGTS filter train pressure drop is a function of air flow rate and filter conditions.

Surveillance testing is perfomed using either the SGTS or drywell purge fans to provide operating convenience.

Each reactor enclosure secondary containment zone and refueling area secondary containment tone is tested independently to verify the design leak tightness.

A design leak tightness of 1250 cfm or less for each reactor enclosure and 764 cfm or less.for the refueling area at a 0.25 inch of vacuum i

water gage will ensure that containment integrity is maintained at an acceptable level if all zones are connected to the SGTS at the same time.

The post LOCA offsite dose analysis assumes a reactor enclosure secondary containment post-draw down leakage rate of 1250 efs and certain post accident X/Q values.

While the post accident X/Q values represent a statistical inter-l protation of historical meteorological data, the highest ground level wind speed which can be associated with these values is 7 mph (Pasquill Gifford stability Class G for a ground level release). Therefo-o, the surveillance requirement assures that the reactor enclosure secondary containment is verified i

t under meteorological conditions consistent with the assumptions utilized in the design basis analysis.

Reactor Enclosure Secondary containment leakage tests that are successfully performed at wind speeds in excess of 7 mph would also satisfy the leak rate surveillance requires 4nts, since it shows compliance i

with more conservative test conditions.

3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL The OPERA 8!LITY of the systems required for the detection and control of hydrogen combustible mixtures of hydrogen and oxygen ensures that these systems i

will be available to maintain the hydrogen concentration within the primary s

containment below the lower flammability limit during post-LOCA conditions.

The primary containment hydrogen recombiner is provided to maintain the oxygen concentration below the lower flammability limit.

The combustible gas analyzer is provided to continuously monitor, both during normal operations and post LOCA, the hydrogen and oxygen concentrations in the primary containment. The primary i

containment atmospheric mixing system is provided to ensure adequate mixing of l

the containment atmosphere to prevent localized accumulations of hydrogen and oxygen from exceeding the lower flammability limit.

The hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7 " Control of Combustible Gas Concentrations in Containment Following a LOCA." March 1971.

i LIMERICK - UNIT 1 8 3/4 6-6 Amendment No. 8

8 3/4.7 PLANT SYSTEMS R

BASES t

3/4.7.1 SERVICE WATER SYSTEMS - CO W N SYSTEMS 1

The OPERABILITY of the service water systems ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.

The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in the accident i

conditions within acceptable limits.

The RHR and ESW systems are common to Units 1 and 2 and consist of two independent subsystems each with two pumps.

One pump per subsystem (loop) is powered from a Unit 1 safeguard bus and the other pump is powered from a Unit 2 Safeguard bus.

In order to ensure adequate onsite power sources to the systems during a loss of offsite power event, the inoperability of these supplies are restricted in system ACTION statements.

RHRSW is a manually operated system used for core and containment heat removal.

Each of two RHRSW subsystems has one heat exchanger per unit.

Each RHRSW pump provides adequate cooling for one RHR heat exchanger.

By limiting operation with less than three OPERABLE RHRSW pumps with OPERABLE Diesel Generators, each unit is ensured adequate heat removal capability for the design scenario of LOCA/ LOOP on one unit and simultaneous safe shutdown of the other unit.

Each ESW pum associated loop. p provides adequate flow to the cooling loads in its With only two divisions of power required for LOCA mitigation of one unit and one division of power required for safe shutdown of the other unit, one ESW pump provides sufficient capacity to fulfill design requirements.

ESW pumps are automatically started upon start of the I

associated Diesel Generators.

Therefore, the allowable out of service times for OPERABLE ESW pumps and their associated Diesel Generators is limited to i

ensure adequate cooling during a loss of offsite power event.

1 1

LIMERICK - UNIT 1 B 3/4 7-1 Amgmgng g27 Qfg;x. h ota m d h w 1.

i PLANT SYSTEMS BASES 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM - COMMON SYSTEM l

1 The OPERABILITY of the control room emergency fresh air supply system ensures that the control room will remain habitable for operations personnel during and following all design basis accident conditions.

Constant purge of the system at I cfm is sufficient to reduce the bulidup of moisture on the adsorbers and HEPA filters.

The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.

This limitation is consistent with the requirements of General i

Design Criterion 19 of Appendix A, 10 CFR Part 50.

3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided 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 without requiring actuation of any of the amergency core cooling system equipment.

The RCIC system is conservatively required to be OPERABLE whenever reactor pressure ex-ceeds 150 psig.

This pressure is substantially below that for which low i

pressure core cooling systems can provide adequate core cooling.

The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2, and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized.

With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCI system and justifies the specified 14 day out-of-service L

period.

The surveillance requirements provide adequate assurance that RCIC will l

be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment.

LIMERICK - UNIT 1 B 3/4 7-la Amendment No. 27, 40 l

.' 3/4.8 ELECTRICAL POWER SYSTEMS l

l BASES 3/4.8.1, 3/4.8.2. and 3/4.8.3 A.C. SOURCES. D.C. SOURCES, and DNSITE POWER DI5TRIBU1 ION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility.

The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation.

The OPERABILITY of the power sources are con-sistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least two of the onsite A.C. and the corresponding D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. or D.C. source.

At least two onsite A.C. and their corresponding D.C. power sources and distribution systems providing power for at least two ECCS divisions (1 Core Spray loop,1 LPCI pump and 1 RHR pump in suppression pool coolin0) are required for design basis accident mitigation as discussed in FSAR Table 6.3-3.

Onsite A.C. operability rcquirements for common systems such as RHRSW and ESW are addressed in the appropriate system specification action statements.

The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources," December 1974. When one or more diesel generators are 1.1 operable, there is an addi-l tional ACTION requirement to verify that all required systems, susbsystems, trains, components, and devices, that depend on the remaining OPERABLE diesel generators as a source of emergency power, are also OPERABLE.

The LPCI mode of the RHR system is considered a four train system, of which only two trains are required.

The verification for LPCI is not required until two diesel generators are inoperable.

This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period when one or more of the diesel l

generators is inoperable.

The term verify as used in this context means to administrative 1y check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons.

It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of.the component.

The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that (1) the facility can be maintained in the shutdown or refueling condition for extended time periods and (2) sufficient instrumentation and control capability i

is available for monitoring and maintaining the unit status.

The surveillance requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971, Regulatory Guide 1.137 " Fuel-011 Systems for Standby Diesel Generators," Revision 1, October 1979 and Regulatory Guide 1.108, LIMERICK - UNIT 1 B 3/4 8-1 Amendment No. 32, 40 l

i

.,J, ELECTRICAL POWER SYSTEMS BASES l

A.C. SOURCES. 0.C. SOURCES. and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

" Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977 except for paragraphs C.2.a(3), C.2.c(1), C.2.c(2), C.2.d(3) and C.2.d(4).

The l

exceptions to Regulatory Guide 1.108 allow for gradual loading of diesel generators during testing and decreased surveillance test frequencies (in response to Generic Letter 84 15).

The surveillance requirements for demonstrating the OPERA 81LITY of the unit batteries are in accordance with the recommendations of Regulatory Guide l

1.129 "Maintenarce Testing and Replacement of Large Lead Storage Batteries for l

Nuclear Power Plants," February 1978 and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Catteries for Generating Stations and Substations."

Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests l

ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity.

Table 4.8.2.11 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity.

The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer s full charge specificgravitywithanaveragespecificgravityofalltheconnectedcells not more than 0.010 below the manufacturer s full charge specific gravity, 4

ensures the OPERABILITY and capability of the battery.

l Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Trible 4.8.2.1-1 is permitted for up to

)

7 days.

During this 7-day period:

(1) the allowable value for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual 1

cell's specific gravity, ensures that an individual cell's specific gravity i

will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual l

cell's float voltage, greater than 2.07 volts, ensures the battery's capa-bility to perform its design function.

LIMERICK - UNIT 1 B 3/4 8-2 Amendment No. 40 l

l

1 1

..-f

\\

ADMINISTRATIVE CONTROLS

6. 9 REPORTING RE0VIREMENTs l

i 00VTINE REPORTS 6.9.1 In accition to the applicable reporting reavirements of Title 10. Coce Acministrator of the Regional Office of the NRC unless

)

i STARTUP REPORT

6. 9.1.1 be submitted following (1) receipt of an Operating License the license involving a planned increase in power level, (3) installation of to s

fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclea i

thermal, or hydraulic performance of the unit.

6. 9.1. 2 The startup report shall addt tss each of the tests identified in sub-tion of the measured values of the operating conditions o i

l obtained during the test program and a comparison of these values with design predictions.and specifications.

obtain satisfactory operation shall also be described.Any corrective actions thf details required in license conditions based on other commitments shall beAny l

included in this report.

6.9.1.3' Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or commence-i ment of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption i

or commencement of commercial operation) supplementary reports shall be submitted-at least every 3 months until all three events have been completed.

ANNUAL REPORTS

5 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following initial criticality.-

6.9.1.5 Reports required on an annual basis shall include:

A tabulation on an annual basis of the number of station, utility, a.

and other personnel (including contractors) receiving exposures greater than 100 mrom/yr and their associated man-rom exposure according to work and job functions ** (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing and refueling).

The done assignments to various duty functions may, be estimated based on pocket

A single submittal may be made for a multiple unit station.

- This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.

LIMERICK - UNIT 1 6 15

- -. -,. - +

~, - -,

e.-

b

-)

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) dosimeter, thermoluminescent dosimeter (TLD), or film badge measure-ments.

Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole-body dose received from external sources should

{

be assigned to specific major work functions; j

b.

Documentation of all challenges to safety / relief valves; and c.

Any other unit unique reports required on an annual basis, d.

The results of specific activity analysis in which the primary coolent exceeded the limits.of Specification 3.4.5.

The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed

)

prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radiciodine concentrations; (3) Cleanup system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the. steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

MONTHLY OPERATING REPORTS

l 6.9.1.6 Routine reports of operating statistics and shutdown experience, includ-ing documentation of all challenges to the the main steam system safety / relief valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC no later than the 15th of each month following the calendar month covered by the report.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6.9.1.7 Routine Annual Radiological Environmental Operating Reports covering i

the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Reports shall incluoe summaries, interpretations, and an analysis of trends of the results of the radiological

A single submittal may be made for a multiple unit station.

LIMERICK - UNIT 1 6-16 Amendment No. 29, 40 l

x

ML20043H874

Text

Attachment:

Navigation menu

Search