Text

Proposed Tech Specs,Relocating Turbine Overspeed Protection Sys Requirements,Relocating Primary Containment Conductor Protection Device Requirements & Revising Fw/Main Turbine Trip Sys Actuation Requirements
	ML20072Q958
Person / Time
Site:	Limerick
Issue date:	08/31/1994
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20072Q953	List: ML20072Q950; ML20072Q958;
References
	NUDOCS 9409120308
	Download: ML20072Q958 (38)
	v • d • e

_

ATTACHMENT 2 LIMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos.

50-352 50-353 License Nos.

NPF-39 NPF-85 TSCR 94-32-0 ' Relocation of Turbine Overspeed Protection System Requirements

List of Affected Pages UnN1 UnN2 x

x 3/4 3-110 3/43110 3/4 3-111 (Deleted) 3/4 3-111 (Deleted)

B 3/4 3-7 B 3/4 3-7 TSCR 94-36-0 ' Relocation of Primary Containment Conductor Protection Device Requirements" List of Affected Pages UnM 1 UnR2 xv xv 3/4 8-21 3/4 8-21 3/4 8-22 (Deleted) 3/4 8-22 (Deleted) 3/4 8-23 (Deleted) 3/4 8-23 (Deleted) 3/4 8-24 (Deleted) 3/4 6-24 (Deleted) 3/4 8-25 (Deleted) 3/4 8-25 (Deleted) 3/4 6-26 (Deleted) 3/4 8-26 (Deleted)

.........................../.4 8-3........................../.4 8-3 B3 B3 TSCR 94-40-0 'Feedwater/ Main Turbine Trip System Actuation Instrumentation Requirements" l

List of Affected Pages Unit 1 Unit 2 3/4 3-113 3/4 3-113 3/4 3-115 3/4 3-115 TSCR 94-42-0

Permit Operability of Low Pressure Coolant Injection While Aligned to Shutdown Cooling" List of Affected Pages UnN1 Unh2 3/45-6 3/45-6

......................./.4 5 7........................... /4 5-7 3

3 9409120308 940831 PDR ADOCK 05000352 P

PDR

TSCR 94-44-0 ' Rem;,ve Temperature Requirement for Operational Condition 5" i

i List of Affected Pages e

1

_Un!L1 Unit 2 1-10 1-10 B 3/4 9-2 B 3/4 9-2 j

TSCR 94-45-0 ' Reduce Frequency of Alternate Decay Heat Demonstration" 1

l List of Affected Pages UnN1 UnN2 3/4 4-25 3/4425 3/4 4-26 3/4 4-2$

l 3/4 9-17 3/4 9-17 1

3/4 9-18 3/4 9-18 l

B 3/4 4-8 B 3/4 4-6 B 3/4 9-2 B 3/4 9-2 i

l 1

1 I

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)

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4 JN_0_EX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE INSTRUMENTATION (Continued)

Table 3.3.7.9-1 Fire Detection Instrumentation............

3/4 3-93 Loose-Part Detection System...............................

3/4 3-97 The information from pages 3/4 3-98 through 3/4 3-101 has been intentionally omitted. Refer to note on page 3/4 3-98..................

3/4 3-98 Of fgas Monitoring Instrumentation.........................

3/4 3-103 Table 3.3.7.12-1 Offgas Monitoring Instrumentation...............

3/4 3-104 Table 4.3.7.12-1 Offgas Monitoring Instrumentation Surveillance Requirements................

3/4 3-107 3/4.3.8 (Deleted) The information on pages 3/4 3-110 and 3/4 3-111 has been intentionally omitted.

Refer to note on page 3/4 3-110.................

3/4 3-110 l

l 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.......................................

3/4 3-112 Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation.........

3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumen-tation Setpoints.........................

3/4 3-114 l

Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumenta-l tion Surveillance Require-ments.....................................

3/4 3-115 l

l 3/4.4 REACTOR COOLANT SYSTEM l

3/4.4.1 RECIRCULATION SYSTEM t

Recirculation Loops.......................................

3/4 4-1 I

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l LIMERICK - UNIT 1 x

i Section 3/4.3.8 (Deleted) i l

' l 4

THE INFORMATION FROM THIS TECHNICAL l

SPECIFICATIONS SECTION HAS BEEN RELOCATED TO Tile UFSAR.

TECHNICAL SPECIFICATIONS PAGE 3/4 3-111 HAS BEEN INTENTIONALLY OMITTED.

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LIMERICK - UNIT 1 3/4 3-110 m

B

-F

INSTRUMENTATION BASES FIRE DETECTION INSTRUMENTATION (Continued) j The loss of detection capability for fire suppression systems, actuated by fire detectors, represents a significant degradation of fire protection for j

any area. As a result, the establishment of a fire watch patrol must be initi-i 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 in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

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

1/4.3.8.

(Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.

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

LIMERICK - UNIT 1 B 3/4 3-7

l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) j Table 3.3.7.9-1 Fire Detection Instrumentation............

3/4 3-93 Loose-Part Detection System...............................

3/4 3-97 The information from pages 3/4 3-98 through 3/4 3-101 has been intentionally omitted. Refer to note on page 3/4 3-98..................

3/4 3-98 Offgas Monitoring Instrumentation.........................

3/4 3-103 Table 3.3.7.12-1 Offgas Monitoring Instrumentation...............

3/4 3-104 Table 4.3.7.12-1 Offgas Monitoring Instrumentation Surveillance Requirements................

3/4 3-107 3/4.3.8 (Deleted) The information on pages 3/4 3-110 and 3/4 3-111 has been intentionally omitted.

Refer to note on page 3/4 3-110.................

3/4 3-110 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.......................................

3/4 3-112 Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation.........

3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip l

System Actuation Instrumen-tation Setpoints.........................

3/4 3-114 j

Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumenta-tion Surveillance Require-ments.....................................

3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation loops.......................................

3/4 4-1 LIMERICK - UNIT 2 x

-f Section 3/4.3.8 (Deleted) h i

4 b

}

THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATD TO THE UFSAR.

TECHNICAL SPECIFICATIONS PAGE 3/4 3-III HAS BEEN INTENTIONALLY OMITTED.

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i LIMERICK - UNIT 2 3/4 3-110 l

l l

INSTRUMENTATION BASES FIRE DETECTION INSTRUMENTATION (Continued) l The loss of detection capability for fire suppression systems, actuated l

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

cny 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 l

in the affected areas is required to provide detection capability until the l

inoperable instrumentation is restored to OPERABILITY.

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 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 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.

(Deleted) - INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.

3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/ main turbine trip system actuation instrumentation is i

l provided to initiate action of the feedwater system / main turbine trip system in the event of failure of feedwater controller under maximum demand.

l LIMERICK - UNIT 2 B 3/4 3-7 l

l i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE ELECTRICAL POWER SYSTEMS (Continued)

Table 4.8.2.1-1 Battery Surveillance Requirements....................

3/4 8-13 D.C. Sources - Shutdown...............................

3/4 8-14 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS Distribution -

Operating.............................

3/4 8-15 Distribution -

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

3/4 8-18 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES

( D el e t e d )............................................

3/4 8-21 Motor-Operated Valves Thermal Overload Protection....

3/4 8-27 Reactor Protection System Electric Power Monitoring...

3/4 8-28 3/4.9 REFUELING OPERATIONS l

3/4.9.1 REACTOR MODE SWITCH...................................

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

3/4 9-3 3/4.9.3 CONTROL R0D P0SITION..................................

3/4 9-5 3/4.9.4 DECAY TIME............................................

3/4 9-6 j

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

3/4 9-7 3/4.9.6 REFUELING PLATF0RM....................................

3/4 9-8 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L................

3/4 9-10 3/4.9.8 WATER LEVEL - REACTOR VESSEL..........................

3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE P00L.................

3/4 9-12 LIMERICK - UNIT 1 xv

1 Section 3/4.8.4.I (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HAS BEEN RELOCATED TO THE UFSAR. TECHNICAL SPECIFICATIONS PAGES 3/4 8-21 THROUGH 3/4 8-26 0F THE SECTION HAVE BEEN INTENTIONALLY OMITTED.

LIMERICK - UNIT I 3/4 8-21

ELECTRICAL POWER SYSTEMS BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES The bypassing of the motor operated valves thermal overload protection continuously by integral bypass devices ensures that the thermal overload pro-tection will not prevent safety related valves from performing their function.

The Surveillance Requirements for demonstrating the bypassing of the thermal j

overload protection continuously are met by functionally testing the automatic operation of the motor operated valve and ensuring that the motor thermal i

overload protection design does not change and is in accordance with Regulatory Guide 1.106 " Thermal Overload Protection for Electric Motors on Motor Operated Valves", Revision 1, March 1977.

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k LIMERICK - UNIT 1 B 3/4 8-3

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE ELECTRICAL POWER SYSTEMS (Continued)

Table 4.8.2.1-1 Battery Surveillance Requirements........................

3/4 8-13 D.C. Sources - Shutdown.....................................

3/4 8-14 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS Distribution - Operating.....................................

3/4 8-15 Distribution -

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

3/4 8-18 i

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES (Deleted)....................................................

3/4 8-21 Hotor-Operated Valves Thermal Overload Protection.............

3/4 8-27 Reactor Protection System Electric Power Monitoring...........

3/4 8-28 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH...........................................

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

3/4 9-3 l

l 3/4.9.3 CONTROL R0D P0SITION..........................................

3/4 9-5 3/4.9.4 DECAY TIME....................................................

3/4 9-6 3/4.9.5 COMMUNICATIONS................................................

3/4 9-7 i

3/4.9.6 REFUELING PLATF0RM............................................

3/4 9-8 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L........................

3/4 9-10 1

3/4.9.8 WATER LEVEL - REACTOR VESSEL..................................

3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE P00L.........................

3/4 9-12 LIMERICK - UNIT 2 xv

Section 3/4 8.4.1 (Deleted) i l

l THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HAS BEEN RELOCATED TO THE UFSAR. TECHNICAL SPECIFICATIONS PAGES 3/4 8-2I THROUGH 3/4 8-26 0F THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.

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l LIMERICK - UNIT 2 3/4 8-21

ELECTRICAL POWER SYSTEMS BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES The bypassing of the motor operated valves thermal overload protection e

continuously by integral bypass devices ensures that the thermal overload pro--

tection will not prevent safety related valves from performing their function.

t The Surveillance Requirements for demonstrating the bypassing of the thermal overload protection continuously are met by functionally testing the automatic operation of the motor operated valve and ensuring that the motor thermal overload protection design does not change and is in accordance with Regulatory i

Guide 1.106 " Thermal Overload Protection for Electric Motors on Motor Operated Valves", Revision 1, March 1977.

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l LIMERICK - UNIT 2 B 3/4 8-3

t i

TABLE 3.3.9-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLE APPLICABLE CHANNELS PER OPERATIONAL i

l TRIP FUNCTION TRIP SYSTEM CONDITIONS 1.

Reactor Vessel Water Level-High, Level 8 4

1*

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r l

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With Thermal Power greater than or equal to 25% of Rated Thermal Power.

LIMERICK - UNIT 1 3/4 3-113

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TABLE 4.3.9.1-1 FEE 0 WATER / MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS l

OPERATIONAL CONDITIONS i

CHANNEL FOR WHICH CHANNEL FUNCTIONAL CHANNEL SURVEILLANCE TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED 1.

Reactor Vessel Water 0

Q R

1*

l Level-High, Level 8 i

i o With Thermal Power greater than or equal to 25% of Rated Thermal Power.

l LIMERICK - UNIT 1 3/4 3-115 l

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TABLE 3.3.9-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION 1

MINIMUM OPERABLE APPLICABLE CHANNELS PER OPERATIONAL TRIP FUNCTION TRIP SYSTEM CONDITIONS i

1.

Reactor Vessel Water Level-High, level 8 4

1*

i 1

l l

o Mith Thermal Power greater than or equal to 25% of Rated Thermal Power.

LIMERICK - UNIT 2 3/4 3-113 I

TABLE 4.3.9.1-1 l

FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 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 I

l With Thermal Power greater than or equal to 25% of Rated Thermal Power.

0 l

LIMERICK - UNIT 2 3/4 3-115 1

EMERGENCY CORE COOLING SYSTEMS 3/4 5.2 ECCS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:

a.

Core spray system (CSS) subsystems with a subsystem comprised of:

1.

Two OPERABLE CSS pumps, and 2.

An OPERABLE flow path capable of taking suction from at least one of the following water sources and transferring the water through the spray sparger to the reactor vessel:

a)

From the suppression chamber, or b)

When the suppression chamber water level is less than the limit or is drained, from the condensate storage tank contair.ing at least 135,000 available gallons of water, equivalent to a level of 29 feet.

b.

Low pressure coolant injection (LPCI) system subsystems with a subsystem comprised of:

1.

One OPERABLE LPCI pump, and 2.

An OPERABLE flow path capable of taking suction from the

)

suppression chamber and transferring the water to the reactor vessel.**

APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5*.

ACTION:

a.

With one of the above required subsystems inoperable, restore at least two subsystems to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspend all operations with a potential for draining the reactor vessel.

b.

With both of the above required subsystems inoperable, suspend CORE ALTERATIONS and all operations with a potential for draining the reactor vessel. Restore at least one subsystem to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish SECONDARY CONTAINMENT INTEGRITY within 1

the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

j l

The ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, and water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

- 0ne LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

LIMERICK - UNIT 1 3/4 6

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0VIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1.*

4.5.2.2 The core spray system shall be determined OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the condensate storage tank required volume when the condensate storage tank is required to be OPERABLE per Specification 3.5.2a.2.b).

00ne LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

LIMERICK - UNIT 1 3/4 5-7 l

EMERGENCY CORE COOLING SYSTEMS 3/4 5.2 ECCS - SHUTOOWN l

LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:

a.

Core spray system (CSS) subsystems with a subsystem comprised of:

l 1.

Two OPERABLE CSS pumps, and 2.

An OPERABLE flow path capable of taking suction from at least one of the following water sources and transferring the water through the spray sparger to the reactor vessel:

l a)

From the suppression chamber, or b)

When the suppression chamber water level is less than the i

limit or is drained, from the condensate storage tank l

containing at least 135,000 available gallons of water, l

equivalent to a level of 29 feet.

b.

Low pressure coolant injection (LPCI) system subsystems with a subsystem comprised of:

l 1.

One OPERABLE LPCI pump, and i

2.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.**

APPLICABILITY: OPERATIONAL CONCITIONS 4 and 5*.

ACTION:

l a.

With one of the above required subsystems inoperable, restore at l

least two subsystems to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspend l

all operations with a potential for draining the reactor vessel.

l i

b.

With both of the above required subsystems inoperable, suspend CORE l

ALTERATIONS and all operations with a potential for draining the i

reactor vessel.

Restore at least one subsystem to OPERABLE status I

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish SECONDARY CONTAINMENT INTEGRITY within l

the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

l oThe ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the spent fuel pool gates are removed, and water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

000ne LPCI subsystem may be considered OPERABLE during alignment and operatien for decay heat removal if capable of being manually realigned and not otherwise inoperable.

l l

LIMERICK - UNIT 2 3/4 5-6

l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0VIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE per Surveillance Requirement 4.5.1.*

4.5.2.2 The core spray system shall be determined OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the condensate storage tank required volume when the condensate storage tank is required to be OPERABLE per Specification 3.5.2a.2.b).

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l 1

00ne LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

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LIMERICK - UNIT 2 3/4 5-7 I

i DEFINITIONS TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE 1.' POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature i

i

3. HOT SHUTDOWN Shutdown # ***

> 200 F

4. COLD SHUTDOWN Shutdown # ## ***

s 200*F

5. REFUELING

Shutdown or Refuel ** #

NA l

l i

The reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

l

1. The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

l

- See Special Test Exceptions 3.10.1 and 3.10.3.

- - The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled provided that the one-rod-out interlock'is OPERABLE.

LIMERICK - UNIT 1 1-10

REFUELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each hoist has sufficient load capacity for handling fuel assemblies and control rods, (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations, and (4) inadvertent criticality will not occur due to fuel being loaded into a unrodded cell.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the nominal weight of a fuel assembly and associated lifting device over other fuel assemblies in the storage pool ensures that in the event this load is dropped 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.

This minimum water depth is j consistent' with the assumptions of the accident analysis.

l3/4.9.10 CONTROL R0D REMOVAL l

.These specifications ensure that maintenance or repair of control rods or

control rod drives will be performed under conditions that limit the probability

of inadvertent criticality.

The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification

' provides for the core to remain subtritical with only one control rod fully

< withdrawn.

l3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be OPERABLE I or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensures that 1) suf-

' ficient cooling capacity is available to remove decay heat and maintain the

eater in the reactor pressure vessel below 140*F, and 2) sufficient coolant circulation l

l would be available through the reactor core to assure accurate temperature indication

! and to distribute and prevent stratification of the poison in the event it becomes j necessary to actuate the standby liquid control system.

The requirement to have two shutdown cooling mode loops OPERABLE when there is less than 22 feet of water above the reactor vessel flange ensures that a l single failure of the operating loop will not result in a complete lo.ts of resid-ual heat removal capability.

With the reactor vessel head removed ard 22 feet

!of water above the reactor vessel flange, a large heat sink is avai'.able for lcorecooling. Thus, in the event a failure of the operating RHR loop, adequate

! time is provided to initiate alternate methods capable of decay heat removal f or emergency procedures to cool the core.

l LIMERICK - UNIT 1 B 3/4 9-2

-. = -.

I l

DEFINITIONS TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE

1. POWER OPERATION Run Any temperature

2. STARTUP Startup/ Hot Standby Any temperature

3. HOT SHUTDOWN Shutdown # ***

> 200*F

4. COLD SHUTDOWN Shutdown # ## ***

s 200*F

5. REFUELING Shutdown or Refuel ** #

NA Y

The reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

1. The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1.

0 Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

t

" See Special Test Exceptions 3.10.1 and 3.10.3.

f

- The reactor mode switch may be placed in the Refuel pos'ition while a single l

control rod is being recoupled provided that the one-rod-out' interlock is l

OPERABLE.

I LIMERICK - UNIT 2 1-10

REFUELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM 1

[

The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each hoist has sufficient load capacity for handling fuel assemblies l and control rods, (3) the core internals and pressure vessel are protected from i excessive lifting force in the event they are inadvertently engaged during (liftingoperations,and(4)inadvertentcriticalitywillnotoccurduetofuel i being loaded into a unrodded cell.

l3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P0OL The restriction on movement of loads in excess of the nominal weight of a fuel assembly and associated lifting device over other fuel assemblies in the

' storage pool ensures that in the event this load is dropped 1) the activity l release will be limited to that contained in a single fuel assembly, and 2) any

, possible distortion of fuel in the storage racks will not result in a critical array.

This assumption is consistent with the activity release assumed in the

' safety analyses.

3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE POOL l

The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.

This minimum water depth is l consistent with the assumptions of the accident analysis.

l3/4.9.10 CONTROL R0D REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives _will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be OPERABLE or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensures that 1) suf-ficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F and 2) sufficient coolant circulation l

' would be available through the reactor core to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes

,! necessary to actuate the standby liquid control system.

The requirement to have two shutdown cooling mode loops OPERABLE when there is less than 22 feet of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete loss of resid-l ual heat removal capability. With the reactor vessel head removed and 22 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event a failure of the operating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal or emergency procedures to cool the core.

LIMERICK - UNIT 2 B 3/4 9-2

REACTOR COOLANT SYSTEM 3/4.4.9 RESIDUAL HEAT REMOVAL H0T SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.1 Two* shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation ** ***

eith each loop consisting of at least:

a.

One OPERABLE RHR pump, and b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint.

ACTION:

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify the availability l

of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

Be in.at least COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .****

b.

With no RHR shutdown cooling mode loop in operation, immediately initiate corrective action to return at least one loop to operation as soon as possible. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish reactor coolant circu-lation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

SURVEILLANCE RE0VIREMENTS 4.4.9.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

0ne RHR shutdown cooling mode loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop is OPERABLE and in operation.

- The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period provided the other loop is OPERABLE.

- - The RHR shutdown cooling mode loop may be removed from operation during i

hydrostatic testing.

(

o***Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

LIMERICK - UNIT 1 3/4 4-25 n.-

REACTOR COOLANT SYSTEM COLD SHUTDOWN l

LIMITING CONDITION FOR OPERATION 3.4.9.2 Two* shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation ** ***

l with each loop consisting of at least:

l L

a.

One OPERABLE RHR pump, and b.

One OPERABLE RHR heat exchanger.

j APPLICABILITY: OPERATIONAL CONDITION 4.

ACTION:

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, within I hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, l

verify the availability of at least one alternate method capable l

l of decay heat removal for each inoperable RHR shutdown cooling mode j

loop.

l b.

With no RHR shutdown cooling mode loop in operation, within I hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

l SURVEILLANCE RE0VIREMENTS 4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i i

0ne RHR shutdown cooling mode loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop is OPERABLE and in operation.

- The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period provided the other loop is OPERABLE.

o**The shutdown cooling mode loop may be removed from operation during hydrostatic testing.

j l

l l

l LIMERICK - UNIT 1 3/4 4-26 j

i REFUELING OPERATIONS 3/4.9.11 RESIDVAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL

_ LIMITING CONDITION FOR OPERATION 3.9.11.1 At least one shutdown cooling mode loop of the residual heat removal i

(RHR) system shall be OPERABLE and in operation

with at least:

1 a.

One OPERABLE.RHR pump, and b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is greater than or equal to 22 feet above the top of the reactor pressure vessel flange.

ACTION:

a.

With no RHR shutdown cooling mode loop OPERABLE, within I hour and i

at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify the availabilty

- of at least one alternate method capable of decay heat removal.

Otherwise, suspend all operations involving an increase in the reactor decay heat load and establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

With no RHR shutdown cooling mode loop in operation, within I hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

l SURVEILLANCE RE0VIREMENTS 4.9.11.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating l

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

l l

OThe shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

-LIMERICK - UNIT 1 3/4 9-17 l

REFUELING OPERATIONS LOW WATER LEVIL l

LIMITING CONDITION FOR OPERATION l

3.9.11.2 Two shutdown cooling mode loop of the residual heat removal (RHR) l system shall be OPERABLE and at least one loop shall be in operation,* with l

each loop consisting of at least:

a.

One OPERABLE RHR pump, and b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is less than 22 feet above the top of the l

reactor pressure vessel flange.

j ACTION:

a.

With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months tiiereafter, verify the availability of at least one alternate method capable of decay heat removal for each inoperable RHR shut-down cooling mode loop.

i b.

With no RHR shutdown cooling mode loop in operation, within I hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REQUIREMENTS 4.9.11.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l oThe shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

l LIMERICK - UNIT 1 3/4 9-18

REACTOR COOLANT SYSTEM BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be j

OPERABLE. The surveillance requirements are based on the operating history of I

this type valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

l 3/4.4.8 STRUCTURAL INTEGRITY

(

The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

l The inservice inspection program for ASME Code Class 1, 2, and 3 components l

will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(1). Additionally, the Inservice Inspection Program conforms to the NRC staff positions identified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," as approved in NRC Safety Evaluations i

dated March 6, 1990 and October 22, 1990.

3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be verified available by either calculation (which includes a review of component and system availability to verify that an alternate decay heat removal method is available) or by demonstration, and that an alternate method of coolant mixing be operational.

l

[

i l

LIMERICK - UNIT 1 B 3/4 4-6 l

Mf_UELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each hoist has sufficient load capacity for handling fuel assemblies and control rods, (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations, and (4 inadvertent criticality will not occur due to fuel being loaded into a unrodde)d cell.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P0OL The restriction on movement of loads in excess of the nominal weight of a fuel assembly and associated lifting device over other fuel assemblies in the storage pool ensures that in the event this load is dropped 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.

This minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.10 CONTROL R0D REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of core than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be OPERABLE or that an alternate method capable of decay heat removal be verified available by either calculation (which includes a review of component and system availability to verify that an alternate decay heat removal method is available) or by demonstration, and that an alternate method of coolant mixing be operational ensures that 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during REFUELING, and 2) sufficient coolant circulation would be available through the reactor core to i

assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby liquid control system.

The requirement to have two shutdown cooling mode loops OPERABLE when there is less than 22 feet of water above the reactor vessel flange ensures that a l

single failure of the operating loop will not result in a complete loss of resid-l ual heat removal capability.

With the reactor vessel head removed and 22 feet of water above the reactor vessel flange, a large heat sink is available for core cooling.

Thus, in the event a failure of the operating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal or emergency procedures to cool the core.

LIMERICK - UNIT 1 B 3/4 9-2

REACTOR C00LRNT SYSTEM 3/4.4.9 RESIDUAL HEAT REMOVAL HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.1 Two* shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation ** ***

with each loop consisting of at least:

a.

One OPERABLE RHR pump, and b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint.

ACTION:

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify the availability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

Be in at least COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .****

b.

With no RHR shutdown cooling mode loop in operation, immediately initiate corrective action to return at least one loop to operation as soon as possible. Within I hour establish reactor coolant circu-lation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

SURVEILLANCE RE0VIREMENTS 4.4.9.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

0ne RHR shutdown cooling mode loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other loop is OPERABLE and in operation.

- The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period provided the other loop is OPERABLE.

- - The RHR shutdown cooling mode loop may be removed from operation during hydrostatic testing.

- - - Whenever two or more RHR subsystems are inoperable, if. unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

LIMERICK - UNIT 2 3/4 4-25

l l

REACTOR COOLANT SYSTEM l

COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.9.2 Two* shutdown cooling mode loops of the residual heat removal (RHR).

system shall be OPERABLE and, unless at least one recirculation pump is in.

operation, at least one shutdown cooling mode loop shall be in operation ** ***

with each loop consisting of at least:

a.

One OPERABLE RHR pump, and b.

One OPERABLE RHR heat exchanger.

(

APPLICABILITY: OPERATIONAL CONDITION 4.

ACTION:

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, within-1 hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify the availability of at.least one alternate method capable l

of decay heat removal for each inoperable RHR shutdown cooling mode loop.

b.

With no RHR shutdown cooling mode loop in operation, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

SURVEILLANCE REQUIREMENTS 4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

'l l

l

0ne RHR shutdown cooling mode loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for l

surveillance testing provided the other loop is-0PERABLE and in operation.

- The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period provided the other loop is OPERABLE.

o**The shutdown cooling mode loop may be removed from operation during hydrostatic testing.

l l

l LIMERICK - UNIT 2 3/4 4-26

l l

l REFUELING OPERATIONS 3/4.9.11 RESIDUAL HEAT REMOVAL AND C0OLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.1 At least one shutdown cooling mode loop of the residual heat removal (RHR) system shall be OPERABLE and in operation

with at least:

a.

One OPERABLE RHR pump, and b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is greater than or equal to 22 feet above the top of the reactor pressure vessel flange.

ACTION:

a.

With no RHR shutdown cooling mode loop OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify the availabilty l

of at least one alternate method capable of decay heat removal.

Otherwise, suspend all operations involving an increase in the reactor decay heat load and establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

With no RHR shutdown cooling mode loop in operation,- within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE RE0VIREMENTS 4.9.11.1 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l OThe shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

l LIMERICK - UNIT 2 3/4 9-17 i

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11.2 Two shutdown cooling mode loop of the residual heat removal (RHR) system shall be OPERABLE and at least one loop shall be in operation,* with each loop consisting of at least:

a.

One OPERABLE RHR pump, and b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor vessel and the water level is less than 22 feet above the top of the reactor pressure vessel flange.

ACTION:

a.

With less than the above required shutdown cooling mode loops of the RHR system OPERABLE, within I hour.and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify the availability of at least one alternate l

method capable of decay heat removal for each inoperable RHR shut-down cooling mode 1 cop.

b.

With no RHR shutdown cooling mode loop in operation, within I hour establish reactor coolant circulation by an alternate niethod and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE RE0VIREMENTS 4.9.11.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

oThe. shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

LIMERICK - UNIT 2 3/4 9-18

REACTOR COOLANT SYSTEM BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of i

this type valve. The maximum closure time has been selected to contain fission i

products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit in:ervice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

The inservice inspection program for ASME Code Class 1, 2, and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i). Additionally, the Inservice Inspection Program conforms to the NRC staff positions identified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," as approved in NRC Safety Evaluations dated March 6, 1990 and October. 22, 1990.

l l

3/4.4.9 RESIDUAL HEAT REMOVAL l

A single shutdown cooling mode loop provides sufficient heat removal l

capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be l

OPERABLE or that alternate methods capable of decay heat removal be verified available by either calculation (which includes a review of component and system availability to verify that an alternate decay heat removal method is available) or by demonstration, and that an alternate method of coolant mixing be operational.

1 l

l l

i LIMERICK - UNIT 2 B 3/4 4-6

REFUELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM The OPERABILITY requirements ensure that (1) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each hoist has sufficient load capacity for handling fuel assemblies and control rods, (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations, and (4)d cell. inadvertent criticality will not occur due to fuel l

being loaded into a unrodde 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the nominal weight of a i

fuel assembly and associated lifting device over other fuel assemblies in the j

storage pool ensures that in the event this load is dropped 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the l

l safety analyses.

I 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL l

STORAGE POOL i

The restrictions on mirimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released l

l from the rupture of an irradiated fuel assembly. This minimum water depth is l

consistent with the assumptions of the accident analysis.

l 3/4.9.10 CONTROL R00 REMOVAL These specifications ensure that maintenance or repair of control rods or l

control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be OPERABLE or that an alternate method capable of decay heat removal be verified available by either calculation (which includes a review of component and system availabilty to verify that an alternate decay heat removal method is available) or by l

demonstration, and that an alternate method of coolant mixing be operational ensures that 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140*F as required during REFUELING, and 2) sufficient coolant circulation would be available through the reactor core to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby liquid control system.

l The requirement to have two shutdown cooling mode loops OPERABLE when there is less than 22 feet of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete loss of resid--

ual heat removal capability. With the reactor vessel head removed and 22 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event a failure of the operating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal or emergency procedures to cool the core.

LIMERICK - UNIT 2 B 3/4 9-2

. -

ML20072Q958

Text

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