Text

Proposed Tech Specs Supporting Change in Calculated Peak Fuel Cladding Temp
	ML20056G492
Person / Time
Site:	Limerick
Issue date:	08/27/1993
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML19310D669	List: ML20056G486; ML20056G492;
References
	NUDOCS 9309030184
	Download: ML20056G492 (88)
	v • d • e

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ATTACHMENT 2 LIMERICK GENERATING STATION UNITS 1 AND 2 l

DOCKET N05. 50-352 i l

50-353 !

LICENSE NOS. NPF-39 )

NPF-85 ;

PROPOSED TECHNICAL SPECIFICATIONS CHANGES 1

NO. 92-08-0 LIST OF ATTACHED PAGES UNIT 1 UNIT 2 1

i i l 11 11 tii iii vi vi xviii xviii 1-2 B 2-7 1-2 B 2-7 1-3 B 3/4 1-3 1-3 B 3/4 1-3 1-4 1-4 1-4 1-4 1-5 1-5 1-5 1-5 1-6

2-1* 1-6 "

2-1*

1-7 2-2 1-7 "

2-2 2-4 2-4 2-4 "

2-4 3/4 1-18 2-5

3/4 1-18 "

2-5*

1-19 4-1 1-19 4-1 1-20 6-3 "

1-20 "- 6-3 2-1 2-1 2-7 6-18a 2-7 6-18a l

2-8 "

2-8 2-9 2-9 3-8 3-8 3-59 3-59 j "

3-60 3-60 l 3-60a 3-60a ,

3-61 3-61 3-62 3-62 4-1 4-1 I 4-la 4-la

These pages reflect changes due to the use of SAFER /GESTR Loss-of-Coolant Accident (LOCA) methodology.

930903o234 DR 930s p ADOCK O PTP M 52 h l gf

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1 INDEX ;

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DEFINITIONS PAGE SECTION 1-1 1.1 ACT10N..................................................... ,

1-1 l 1.2 AVERAGE PLANAR EXP05URE....................................

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( 1-1 l 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RAT E. . . . . . . . . . . . . . . . . .

1-1 j 1.4 CHANNEL CAllBRATI0N.........................................

1-1 1.5 CHANNEL CHECK..............................................

1-1 1.6 CHANNEL FUNCTIONAL TEST....................................

1-2 1.7 CORE ALTERATION............................................

1-2 1.7A CORE OPERATING LIMITS REP 0RT............................... i 1-2 l 1.8 CRITICAL POWER RATI0................... . . . . . . . . . . . . . . . . . .

l 1-2 l 1.9 DOSE EQUIVALENT I ' 31. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 1-2 1.9a DOWNSCALE TRIP SET POINT (DTSP) ........................... ,

i 1-2 1.10 E-AVERAGE DISINTEGRATION ENERGY............................

1-2 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME.........

1-3 1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME..

1-3 1.13 (DELETED)..................................................

1-3 1.14 (DELETED)..................................................

1-3 1.15 FREQUENCY N0TATION..........................................

(HTSP)........................... 1-3 1.15a HIGH (POWER) TRIP SETPOINT 1-3 1.16 IDENTIFIEC LEAKAGE.........................................

1-3 l 1.16a INTERMEDI ATE (POWER) TRIP SFTP0lNT (ITSP) . . . . . . . . . . . . . . . . .

1-3 1.17 ISOLATION SYSTEM RESPONSE TIME.............................

1.18 LIMITING CONTROL R0D PATTERN...............................

1-3 1-4 1.19 LINEAR HEAT GENERATION RATE................................

1-4 1.20 LOGIC SYSTEM FUNCTIONAL TEST...............................

LIMERICK - UNIT 1 i l l

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-,y- p,yyem . -. -'w q, ,.- - - y p--Mi-:g- y- g9y*r yo- -,

i INDEX DEFINITIONS :

1 SECTION PAGE DEFINITIONS (Continued)

(LTSP)........................ 1-4 l 1 1.20a LOW (POWER) TRIP SET POINT 1-4 1.21 (DELETED)................................................... .

1-4 1.22 MEMBER (S) 0F THE PUBLIC....................................

i 1-4 j 1.22a MAPFAC(F) - (MAPLHGR FLOW FACT 0R).......................... 6 l

1.22b MAPFAC(p) - (POWER DEPENDENT MAPLHGR MULTIPLIER)........... 1-4 MINIMUM CRITICf.L POWER RATIO (MCPR)......................... 1-4 1 1.23 0FFSITE DOSE CALCULATION MANUAL............................

1-4 1.24 ,

1-4 1.25 O P E R AB L E - O P E RAB I L I T Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1.26 OPERATIONAL CONDITION - CONDITION.......................... 1-5 1.27 PHYSICS TESTS............................................... 1-5 i

1.28 PRESSURE B0UNDARY LEAKAGE.................................. 1-5 1.29 PRIMARY CONTAINMENT INTEGRITY.............................. 1-5 1.30 PROCESS CONTROL PR0 GRAM.................................... 1-5 l 1.31 PURGE - PURGING............................................ 1-6 1.32 RATED THERMAL P0WER........................................ 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY.......... 1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME.................... 1-6 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY............ 1-6 1.36 REPORTABLE EVENT............................................ 1-7 1.37 ROD DENSITY................................................ 1-7 1.38 SHUTDOWN MARGIN.......................,.................... 1-7 1.39 SITE B0VNDARY.............................................. 1-7 1-7 1.40 (DELETED)..................................................

1.41 SOURCE CHECK............................................... 1-7 LIMERICK - UNIT 1 ii n

INDEX DEFINITIONS PAGE SECTION OEFINITIONS (Continued) 1-8 1.42 STAGGERED TEST BASIS.......................................

1-8 1.43 THERKAL P0WER..............................................

1-8 1.43A TURBINE BYPASS SYSTEM RESPONSETIME........................

1-8 1.44 UNIDENTIFIED LEAKAGE.......................................

1-8 1.45 UNRESTRICTED AREA..........................................

1-8 1.46 VENTILATION EXHAUST TREATMENT SYSTEM.......................

1-8 1.47 VENTING....................................................

1-9 Table 1.1 Surveill ance Frequency Notation. . . . . . . . . . . . . . . . . . . . . . .

1-10 T abl e 1.2 Operational Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

iii LIMERICK - UNIT 1

INDEX !

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION POWER DISTRIBUTION LIMITS (Continued)

(DELETE 0)............................................. 3/4 2-7 3/4 2.2 MINIMUM CRITICAL POWER RATI0.......................... 3/4 2-8 3/4 2.3 .

Information on pages 3/4 2-10 thru 3/4 2-11 has been INTENTIONALLY OMITTED, refer to Note on page 3/4 2-10. 3/4 2-10 LINEAR HEAT GENERATION RATE........................... 3/4 2-12 3/4.2.4 3/4.3 INSTRUMENTATION REACTOR PROTECTION SYSTEM INSTRUMENTATION............. 3/4 3-1 3/4.3.1 Table 3.3.1-1 Reactor Protection System .

Instrumentation............... 3/4 3-2 l Table 3.3.1-2 Reactor Protection System Instrumentation............... 3/4 3-6 ;

1 Table 4.3.1.1-1 Reactor Protection System i Instrumentation Surveillance Requirements................. 3/4 3-7 l

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LIMERICK UNIT - I vi

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INDEX j BASES i

PAGE SECTION 3/4.0 APPLICABILITY............................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS l 3.4.11 SHUTDOWN MARGIN....................................... B 3/4 1-1 l

REACTIVITY AN0MALIES.................................. B 3/4 1-1 3/4.1.2 3/4.1.3 CONTROL R005............................ ............ B 3/4 1-2 ,

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3.4.1.4 CONTROL R0D PROGRAM CONTR0LS.......................... B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM......................... B 3/4 1-4 3 /4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION ,

B 3/4 2-1 RATE..................................................

LEFT INTENTIONALLY BLANK.................................... ... B 3/4 2-3 3 1

3/4.2.2 (DELETED)............................................. B 3/4 2-2 1

3/4.2.3 MINIMUM CRITICAL POWER RATI0.......................... B 3/4 2-4 l 3/4.2.4 LINEAR HEAT GENERATION RATE........................... B 3/4 2-5 3 /4. 3 INSTRUMENTATION l

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............. B 3/4 3-1 !

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3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION. . . . . . . . . . . . . . . . . . B 3/4 3-2 ;

l 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION i INSTRUMENTATION....................................... B 3/4 3-2 l 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION..... B 3/4 3-3 ;

l 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION....................................... B 3/4 3-4 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION..................... B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.................. B 3/4 3-4 LIMERICK - UNIT 1 xviii

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs, IRMs, TIPS, or special movable detectors is not considered a l CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement os a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that i provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.1.12. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO l.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE E0VIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133,1-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

DOWNSCALE TRIP SET POINT (DTSP) 1.f a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.

E-AVERAGE DIS!NTEGRATION ENERGY 1.10 i shall be the average, weighted in proportion to the concentration of i each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set- ;

point at the channel sensor until the ECCS equipment is capable of cerforming ,

its safety function, i.e., the valves travel to their required positions, :

pump discharge pressures reach their required values, etc. Times shall !

include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMERICK - UNIT 1 1-2 f

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DEFINITIONS -

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i END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME l.12 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric are between the fully open contacts of the recirculation pump circuit breaker from '

initial movement of the associated:

a. Turbine stop valves, and >

b. Turbine control valves.

This total system response time consists of two components, the instrumen- !

tation response time and the breaker arc suppression time. These times may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. P 1.13 (Deleted) 1.14 (Deleted)

FRE0VENCY NOTATION '

1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

HIGH (POWER) TRIP SET POINT (HTSP) '

1.15a The high power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable above 85% reactor thermal power.

IDENTIFIED LEAXAGE 1.16 IDENdFIED LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or Leakage into the containment atmosphere from sources that are both

b. '

specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE. .

INTERMEDIATE (POWER) TRIP SET POINT (ITSP) 1.16a The intermediate power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable between 65% and 85% reactor thermal power.

ISOLATION SYSTEM RESPONSE TIME 1.17 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions.

Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMITING CONTROL R00 PATTERN 1.18 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, OR MCPR.

LINEAR HEAT GENERATION RATE 4 1.19 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LIMERICK - UNIT 1 1-3

DEFINITIONS LOGIC SYSTEM FUNCTIONAL TEST 1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

LOW (POWER) TRIP SET POINT (LTSP) 1.20a The low power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable between 30% and 65% reactor thermal power.

1.21 (Deleted)

MEMBER (S) 0F THE PUBLIC 1.22 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include p 2rsons who use portions of the site for recrea-tional, occupational, or other purposes not associated with the plant.

MAPFAC(F)-(MAPLHGR FLOW FACTOR) 1.22a A core flow dependent multiplication factor used to flow bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit.

MAPFAC(P)-(POWER DEPENDENT MAPLHGR MULTIPLIER) 1.22b A core power dependent multiplication factor used to power bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit.

MINIMUM CRITICAL POWER RATIO (MCPR1 l.23 The MAXIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core (for each class of fuel). Associated with the minimum l

critical power ratio is a core flow dependent (MCPR(F)) and core power dependent (MCPR(P)) minimum critical power ratio.

OFFSITE DOSE CALCULATION MANUAL 1.24 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology l and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluent, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.84 and (2) descriptions of the information that should be included in the Annual radiological Environmental Operating and Semi-annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.

OPERABLE - OPERABILITY 1.25 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their relater support function (s).

I LIMERICK - UNIT 1 I-4 L

gFINITIONS m -

r OPERATIONAL CONDITION - CONDITION l.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant tempera-ture as specified in Table 1.2.

PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14 of the FSAR, (2) authorized under the provisions of 10 CFR 50, 59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel' wall.

PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All primary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or Closed by at least one manual valve, blind flange, or 2.

deactivated automatic valve secured in its closed position, except as provided in Table 3.5.3-1 of Specification 3.6.3.

b. All primary containment equipment hatches are closed and sealed.

c. The primary containment air lock is in compliance with the requirements of Specification 3.5.1.3.

d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.

e. The suppression chamber is in compliance with the requirements of Specification 3.6.1.2.

f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or 0-rings, is OPERABLE.

PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure that the solidification or dewatering and packaging of radioactive wastes results in a waste package with propert.ies that meet the minimum and stability requirements of 10 CFR Part 61 and other requirements for transportation to the disposal site and receipt at the disposal site.

With solidification or dewatering, the PCP shall identify the process parameters influencing solidification or dewatering, based on laboratory scale and full scale testing or experience.

f 1

LIMERICK - UNIT 1 1-5

! l DEFINITIONS PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that i replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor ore heat transfer rate to the reactor coolant of 3293 MWT.

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! REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY i

l 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when: ,

a. All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed ;

position, except as provided in Table 3.6.5.2.1-1 of l Specification 3.6.5.2.1. ,

b. All reactor enclosure secondary containment hatches and blowout panels i are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4. t

e. At least one door in each access to the reactor enclosure secondary containment is closed. i

f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.

g. The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.la. !

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:

I LIMERICK - UNIT 1 1-6 i i

DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)

1. Capable of being closed by an OPERABLE secondary containment '

automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed .

position, except as provided in Table 3.6.5.2.2-1 of Specification 3.6.5.2.2. !

b. All refueling floor secondary containment hatches cnd blowout panels are closed and sealed. <

c. The standby gas treatment system is in compliance with the requirements ,

of specification 3.6.5.3.

d. At least one door in each access to the refueling floor secondary .

containment is closed.

e. The sealing mechanism associated with each refueling floor secondary ,

containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.

I

f. The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.

REPORTABLE EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in Section '

50.73 to 10 CFR Part 50. :

R00 DENSITY j 1.37 ROD DENSITY shall be the number of control rod notches inserted as a fraction j of the total number of control rod notches. All rods fully inserted is I equivalent to 100% R0D DENSITY.

SHUTOOWN MARGIN 1.38 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subtritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which i is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.

SITE BOUNDARY l.39 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3.la.

l 1.40 (Deleted)

SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK - UNIT 1 1-7

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE TRIP SETPOINT VALUES FUNCTIONAL UNIT .

1. Intermediate Range Monitor, Neutron Flux-liigh 5 120/125 divisions s 122/125 divisions-of full scale of full scale

2. Average Power Range Monitor:

a. Neutron Flux-Upscale, Setdown s 15% of RATED TilERMAL POWER $ 20% of RATED THERMAL POWER C

$ b. Neutron Flux-Upscale 5 1) During two recirculation loop operation:

Flow Biased s 0.66 w+ 65%, with s 0.66 W+ 68%, with R a)' a maximum of i

a maximum of High Flow Clamped s 115% of RATED s 117% of RATED e b) THERMAL POWER 5 THERMAL POWER H

2) During single recirculation loop operation:

Flow Biased s 0.66 W+ 59%, s 0.66 W+ 62%,

a) Not Required b) High Flow Clamped Not Required OPERABLE OPERABLE

c. Inoperative N.A. N.A.

d. Downstale 2 4% of RATED 2 3% of RATED THERMAL POWER _ _ THERMAL POWER

3. Reactor Vessel Steam Dome Pressure - High s 1037 psig $ 1057 psig

4. Reactor Vessel Water Level - Low, Level 3 2 12.5 inches above instrument 2 11.0 inches above zero instrument zero

{

5. Main Steam Line Iso 7ation Valve - Closure s 8% closed s 12% closed

6. Main Steam Line Radiation - High s 3.0 x full power background 5 3.6 x full power background

7. Drywell Pressure - High 5 1.68 psig s 1.88 psig

8. Scram Discharge Volume Water level - High

a. Level Transmitter s 260' 9 5/8" elevation ** 5261'55/8" elevation

b. Float Switch 5 260' 9 5/8" elevation ** $261'55/8" elevation

9. Turbine Stop Valve - Closure s 5% closed 5 7% closed ~

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 2 500 psig 2 465 psig

11. Reactor Mode Switch Shutdown Position N.A. N.A.

12. Manual Scram N.A. N.A.

oSee Bases Figure B 3/4.3-1.

90 Equivalent to 25.45 gallons / scram discharge volume.

REACTIVITY CONTROL SYSTEMS ROD BIDCK MONITOR LIMITING CONDITION FOR OPEPATION <

3.1.4.3 Both rod block monitor (RBM) channels shall be OPERABLE.

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APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or i

equal to 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER with '

l MCPR less than 1.70, or THERMAL POWER greater than or equal to 90% of rated with MCPR less than 1.40 .

ACTION:

a. With one REM channel inoperable: l Verify that the reactor is not operating on a LIMITING CONTROL

1. [

ROD PATTERN, and

2. Restore the inoperable REM channel to OPERABLE status within ,

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

Otherwise, place the inoperable rod block monitor channel in the

) tripped condition within the next hour.

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t FURVEILLANCE REOUIREMENTS ,

4.1.4.3 Each of the above required REM channels shall be demonstrated OPERABLE I by performance of a:

a. CHANNEL FUNCTIONING TEST and CHANNEL CALIBRATION at the frequencies and for the OPERATIONAL CONDITIONS specified in Table 4.3,6-1.

b. CRGNEL FUNCTIONAL TEST prior to contrcl rod withdrawal when the j

reactor is operating on a LIMITING CONTROL ROD PATTERN. l LIMERICK - UNIT 1 3/4 1-18 l 1

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REACTIVITY CONTROL SYSTEMS

3/4.1. 5 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system consisting of a minimum of two pumps and corresponding flow paths, shall be OPERABLE.

APPLICABILITY: OPERATION CONDITIONS 1, 2, and 5*

ACTION:

a. In OPERATIONAL CONDITION 1 or 2:

1. With only one pump and corresponding explosive valve OPERABLI, ;

restore one inoperable pump and correspon ling explosive valve 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 .

2. With standby liquid control system otherwise inoperable, i restore the system 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 ,

b. In OPERATION CONDITION 5*:

I

1. With only one pump and corresponding explosive valve OPERABLE, restore one inoperable pump and corresponding explosive valve '

to OPERABLE status within 30 days or insert all insertable control rods within the next hour.

2. With the standby liquid control system otherwise inoperable, insert all insertable controi rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SURVEILLANCE REOUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE: )

i

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that:

1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

l

2. The available volume of sodium pentaborate solution is at least j 3160 gallons. l

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis. l l

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

l LIMERICK - UNIT 1 3/4 1-19 l

REACTIVITY CONTROL SYSTEMS j i

SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days by:

1. Verifying the continuity of the explosive charge. J

2. Determining by chemical analysis and calculation

that the available weight of sodium pentaborate is greater than or equal [

to 3754 lbs; the concentration of sodium pentaborate in l solution is less than or equal to 13.8% and within the limits J

( of Figure 3.1.5-1 and; the following equation is satisfied:

l C x E x 0 21 13% wt. 29 atom % 86gpm where C - Sodium pentaborate solution (% by weight)

Q - Two pump flowrate, as determined per surveillance requirement 4.1.5.c.

! E - Boron 10 enrichment (atom % Boron 10) l t

3. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

A

c. Demonstrating that, shen tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of greater j

than or equal to 1190 psig is met.

At least once per 24 months by:

d.

l. Initiating at least one of the standby liquid control system !

loops, including an explosive valve, and verifying that a flow l path from the pumps to the reactor pressure vessel is available l

by pumping demineralized water into the reactor vessel. The ;

replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch dhich has been certified by having one of the batch success-fully fired. All injection loops shall be tested in 3 operating j cycles.

2. ** Demonstrating that all heat traced piping is unblocked by pumping l

from the storage tank to the test tank and then draining and ,

flushing the piping with demineralized water.

l

3. Demonstrating that the storage tank heaters are OPERABLE by verifying the expected temperature rise of the sodium pentaborate solution in the storage tank after the heaters are energized.

! e. Prior to addition of Boron to storage tank verify sodium pentaborate enrichment to be added is 2 29 atom % Boron 10.

This test shall also be performed anytime water or boron is added to the solu-tien or when the solution temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or boron addition or solution temperature is restored.

- This test shall also be performed whenever suction piping temperature Props below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 3

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution teaperature is restored.

LIMERICK - UNIT 1 3/4-1-20

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of axial location and AVERAGE PLANAR EXPOSURE shall be within limits based on applicable APLHGR limit values which have been ,

determined by approved methodology for the respective fuel and lattice types. 7 When hand calculations are required, the APLHGR for each type of fuel as a l ;

function of AVERAGE PLANAR EXPOSURE shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) as shown in the CORE OPERATING LIMITS REPORT (COLR). During operation, the APLHGR for each fuel type shall not exceed the above values multiplied by the appropriate reduction factors for power and flow as defined in the COLR. ,

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than l l l or equal to 25% of RATED THERMAL POWER.

l l ACTION:

With an AP!HGR exceeding the limiting value, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 ,

hours.

SURVEILLANCE RE0VIREMENTS ,

4.2.1 All APLHGRs shall be verified to be equal to or less than the limiting value. i

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R0D PATTERN for APLHGR.

d. The provisions of Specification 4.0.4 ar not applicable.

l i

LIMERICK - UNIT 1 3/4 2-1

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Section 3/4.2.2 (DELETED)

P t

l INFORMATION CONTAINED ON t

THIS PAGE HAS BEEN l r DELETED r t

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l LIMERICK - UNIT I 3/4 2-7

g POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATION (MCPR) shall be equal to or greater than the rated MCPR limit adjusted by the MCPR(P) and MCPR(F) factors as shown in the l CORE OPERATIONS LIMITS REPORT, provided that the end-of-cycle recirculation pump trip (E0C-RPT) system is OPERABLE per Specification 3.3.4.2 and the main turbine bypass system is OPERABLE per Specification 3.7.8, with:

r - (' ave - 'B)

'A 'B where:

'A = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3,

'B = 0.672 + 1.65 ( "I )2/'(0.016) ,

n N, I

i=1

n I

i ' ave = i=1 "i'i ,

j n N, l E i=1 n = number of surveillance tests performed to date in cycle, N' = number of active control rods measured in the i" surveillance test, r, - average scram time to notch 39 of all rods measured in the i" surveillance test, and N, - total number of active rods measured in Specification 4.1.3.2.a.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

l l

LIMERICK - UNIT 1 3/4 2-8 l

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

ACTION

a. With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue provided that, within I :

hour, MCPR is determined to be greater than or equal to the rated MCPR ]

limit as a function of the average scram time (shown in the CORE OPERA-TING LIMITS REPORT) EOC-RPT inoperable curve, adjusted by the MCPR(P) :

and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT.

b. With MCPR less than the applicable MCPR limit adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. With the main turbine bypass system inoperable per Specification 3.7.8, operation may continue provided that, within I hour, MCPR is determined !

to be greater than or equal to the rated MCPR limit as a function of the l average scram time (shown in the CORE OPERATING LIMITS REPORT) inain turbine bypass valve inoperable curve, adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT.

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

a. t = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or l b. r as defined in Specification 3.2.3 used to determine the limit '

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test f required by Specification 4.1.3.2, shall be determined to bc equal to or greater than the applicable MCPR limit including application of the MCPR(P) and MCPR(F) factors as determined from the CORE OPERATING LIMITS REPORT.

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least l 15% of RATED THERMAL POWER, and j

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating j with a LIMITING CONTROL R0D PATTERN for MCPR.

d. The provisions of Specification 4.0.4 are not applicable.

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

. TABLE 4.3.1.1-1 (Continued) ,

y -

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS m

5 CllANNEL OPERATIONAL E CilANNEL FUNCTIONAL CilANNEL CONDITIONS FOR WHICH i FUNCTIONAL (WIT CilECK TEST CALIBRATION"8 SURVEILLANCE RE0VIRED

9. Turbine Stop valve - Closure N.A. Q R 1

~ 10. Turbine Control Valve Fast

Closure, Trip Oil Pressure - Low N.A. Q R 1 l 11. Reactor Mode Switch Shutdown Position N.A. R N.A. 1,2,3,4,S 2 12. Manual Scram N.A. W N.A. 1,2,3,4,5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

! (b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after

entering OPERATION CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

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

" (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by i

a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL-POWER. Adjust the APRM '

! Y channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

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

, signal.

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

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

j l (h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 1 3.10.1.

l (i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

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

l (k) Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 4 3.10.3.

i

l TABLE 3.3.6-1 (Continued)

CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION t I

i ACTION STATEMENTS '

l Declare the REM inoperable and take the ACTION required by l ACTION 60 -

Specification 3.1.4.3.

With the number of OPERABLE channels one or more less than ACTION 61 -

required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the j

tripped condition within one hour, i

ACTION 62 - With the number of OPERABLE channels less than required by the f Minimum OPERABLE Channels per Trip Function requirements, place l the inoperable channel in the tripped condition within one hour.

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

a rod block.

NOTES j

For OPERATIONAL CONDITION of specification 3.1.4.3. !

i

- With more than one control rod withdrawn. Not applicable to control rods ,

removed per specification 3.9.10.1 or 3.9.10.2.

- - These channels are not required when sixteen or fewer fuel assemblies, f adj acent to the SRMs, are in the core.

(a) The REM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of l j RATED THERMAL FOWER. i l

l (b) This function shall be automatically bypassed if detector count rate is i

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

l (c) This function is automatically bypassed when the associated IRM channels i are on range B or higher. l (d) This function is automatically bypassed when the IRM channels are on j range 3 or higher. :

?

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

(f) Required to be OPERABLE only prior to and during shutdown margin ;

demonstrations as performed per Specification 3.10.3. 1 l

J l

LIMERICK - UNIT 1 3/4 3-59 I

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!B

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWADLE VALUE ,

1. R00 BLOCK MONITOR

a. Upscale

1) Low Trip Setpoint (LTSP)

2) Intermediate Trip Setpoint (ITSP)

g 3) liigh Trip Setpoint (HTSP) m >

. 5 b. Inoperative N/A N/A O c. Downscale (DTSP) N i d. Power Range Setpoint

. c 1) Low Power Setpoint (LPSP) 23% RATED THERMAL POWER 26% RATED THERMAL POWER 5 2) Intermediate Power Setpoint (IPSP) 58% RATED THERMAL POWER 61% RATED THERMAL POWER

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

2. ftPRM

a. Flow Biased Neutron Flux - Upscale

1) During two recirculation loop 5 0.66 W + 51%* s 0.66 W + 55%* l

, s operation

2) During single recirculation loop s 0.66 W + 45% $ 0.66 W + 49%*

y operation

. m b. Inoperative N.A. N.A.

I

c. Downscale 2 4% of RATED THERMAL POWER 2 3% of RATED THERMAL POWER

d. Neutron Flux - Upscale, Startup $ 12% of RATED THERMAL POWER $ 14% of RATED THERMAL POWER

3. SOURCE RANGE MONITORS

a. Detector not full in N.A. N.A.

b. Upscale s 1 x 10' cps s 1.6 x 10' cps

c. Inoperative N.A. N.A.

j d. Downscale 2 3 cps ** 2 1.8 cps **

4 INTERMEDIATE RANGE MONITORS

a. Detector not full in N.A. N.A.

i b. Upscale s 108/125 divisions of 5 110/125 divisions of i full scale full scale 4 c. Inoperative N.A. N.A. /

. d. Downscale 2 5/125 divisions of full 2 3/125 divisions of' full scale scale

! 5. SCRAM DISCHARGE VOLUME 3

a. Water Levei-Higd s 257' 5 9/16" elevation *** $ 257" 7 9/16" elevation

_ - - . = . . __. .

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

6. REACTOR COOLANT SYSTEM RECIRCULATION M l C l M a. Upscale s 111% of rated flow $ 114% of rated flow 5 b. Inoperative N.A. N.A.

R c. Comparator s 10% flow deviation s 11% flow deviation i

e 7. REACTOR MODE SWITCH SHUTDOWN 5 POSITION N.A. N.A. i s

.i Y

8

Refer to the COLR for these setpoints.

- May be reduced provided the Source Range Monitor has an observed count rate and signal-to-noise ratio on or above the curve shown in Figure 3.3.6-1.

- - Equivalent to 13 gallons / scram discharge volume.

(a) There are three upscale trip levels. Each is aplicable only over its specified operating core thermal power range. All RBM trips are automatically bypassed below the low power setpoint (LPSP). The upscale LTSP is applied between the low power setpoint (LPSP) and the intermediate power setpoint (IPSP). The Wscale ITSP is applied between the intermediate power setpoint and the high power setpoint ,

(HPSP). The HTSP is applied above the high power setpoint.

(b) Power range setpoints control enforcement of appropriate upscale trips over the proper core thermal power ranges. The power signal to the RBM is provided by the APRM.

4 4

TABLE 4.3.6-1 IS01.ATION ACTUATION INSTRUMENTATION SURVElllANCE REQUIREMENTS CilANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION") SURVEILLANCE RE0VIRE

1. R00 BLOCK MONITOR

a. Upscale N.A. S/U""*), ") R 1*

b. Inoperative N.A. S/U""", ")

N.A. 1*

c. Downscale N.A. S/ U *)") , ")

R 1* l

2. APRM

a. Flow Biased Neutron Flux-4 Upscale N.A. S/US ),0 SA 1 R b. Inoperative N.A. S/U"),Q N.A. 1, 2, 5***

i c. Downscale N.A. S/U"),Q SA 1 e d. Neutron Flux - Upscale, Startup N.A. S/U"),Q SA 2, 5***

}3.SOURCERANGEMONITORS

~

a. Detector not full in N.A. S/U*),W N.A. 2, 5

b. Upscale N.A. S/U") W SA 2, 5

c. Inoperative N.A. S/U"),W N.A. 2, 5

d. Downscale N.A. S/U"),W, SA 2, 5

4. INTERMEDIATE RANGE MONITORS

a. De'.ector not full in N.A. S/US),W N.A. 2, 5 w b. Upscale N.A. S/US),W SA 2, 5

& c. Inoperative N.A. S/U"),W N.A. 2, 5

~

d. Downscale N.A. S/U"),W SA 2, 5

5. SCRAM DISCHARGE VOLUME

a. Water Level - High N.A. Q R 1, 2, 5**

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

! a. ' Upscale N.A. S/US),Q SA 1

b. Inoperative N.A. S/U"),Q N.A. I

c. Comparator N.A. S/US),Q SA 1

7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. R N.A. 3, 4

_ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _m._____.____._. m__ _ _ _ . - _ _ _ - - --, . - - - - ~ - _ - , ,,---_ .. - - - _ , , _ . - _ . . . , , _ . .

TABLE 4.3.6-1 (Continued)

?

CONTROL ROD BIDCK INSTRUMENTATION SURVEILLANCE REOUIRFEENTS i

TABLE NOTATIONS k

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

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

1 4

(c) Includes reactor manual control multiplexing system input.

For OPERATIOKAL CONDITION of Specification 3.1.4.3.

1 j

i ** Vith more than one control rod withdrawn. Not applicable to control ;

l rods removed per Specification 3.9.10.1 or 3.9.10.2.

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

demonstrations as performed per Specification 3.10.3. ,

i' 1

e I

r 1

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I

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LIMERICK - UNIT 1 3/4 3-62 l

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{ l 1

i 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM 1

RECIRCULATION LOOPS I

LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:

Total core flow greater than or equal to 45% of rated core flow, or a.

b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1.

f APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

i ACTION:

a. k'ith one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a. Place the recirculation flow control system in the local :

l Manual mode, and I

b. Reduce THERMAL POWER to s 70% of RATED THERMAL POWER, and ,

t l

l

c. Limit the speed of the operating recirculation pump to less l than or equal to 90% of rated pump speed, and j

d. Verify that the differential temperature requirements of l ;

Surveillance Requirement 4.4.2.1.5 are met if THERMAL POWER is s 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is s 50% of rated loop flow, or suspend the THERMAL POWER or recirculation loop ;

flow increase. I i

See Special Test Exception 3.10.4.

l I

LIMERICK - UNIT I 3/4 4-1 1

I I

REACTOR COOLANT SYSTEM f LIMITING CONDITION FOR OPERATION (Continued) i ACTION: (Continued)

2. Within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s:

Reduce the Average Power Range Monitor (APRM) Scram and Rod l Block Trip Setpoints and Allowable Values, to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6, or declare the associated channel (s) inoperable and take the actions required by the referenced specifications, and, ,

3. The provisions of Specification 3.0.4 are not applicable.

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

b. With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER such that it is not within the restricted zone of Figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . .

c. With one or two reactor coolant system recirculation loops in .

operation and total core flow less than 45% but greater than 39% of rated core flow and THERMAL POWER within the restricted zone of Figure 3.4.1.1-1: ;

l. Determine the APRM and LPRM** noise levels (Surveillance 4.4.1.1.3):

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

b. Within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.

i

2. With the APRM or LPRM** neutron flux noise levels greater than three times their established baseline noise levels, within 15 minutes initiate corrective action to restore the noise levels within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow or by reducing THERMAL POWER.

d. With one or two reactor coolant system recirculation loops in operation and total core flow less than or equal to 39% and THERMAL POWER within the restricted zone of Figure 3.4.1.1-1, within 15 minutes initiate corrective action to reduce THERMAL POWER to within '

the unrestricted zone of Figure 3.4.1.1-1 or increase core flow to graater than 39% within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ,

- Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

LIMERICK - UNIT 1 3/4 4-la

t j LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Averare Power Ranne Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because ceveral rods must be moved to change power l by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform v rod withdrawal approach to the trip level, the rate of power rise is not rnore than 5% of RATED THERMAL POWER per minute and the APRM system would be more !

than adequate to assure shutdown before the power could exceed the Safety Limit.

The 15% neutron flux trip remains active until the mode switch is placed in ;

the Run position.

i The APRM trip system is calibrated using heat balance data taken during ,

j steady state conditions. Fission chambers provide the basic input to the (

system and therefore the monitors respond directly and quickly to changes due i to transient operation for the case of the Neutron Flux-Upscale flow bias l setpoint- i.e., for a power increase, the THERMAL power of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat i

transfer associated with the fuel. l i

The APRM setpoints were selected to provide adequate margin for the Safety j Limits and yet allow operating margin that reduces the possibility of unneces- I sary shutdown.

2

3. Reactor Vessel Steam Dome Pressure-Hirh High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by co= pressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design 1 pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power /fic. conditions when the turbine stop valve and control fast closure trips are bypassed. For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

d LIMERICK - UNIT 1 B 2-7

.. __4

i t

N ACTIVITY CONTROL SYSTEMS i f

BASES CONTROL RODS (Continued)

Control rod coupling integrity is required to ensure compliance with the analysis of the red drop accident in the FSAR. The e,vutravel position feature ;

provides the only positive means of determining that a rod is properly coupled i and therefore this check must be performed prior to achieving criticality after l completing CORE ALTERATIONS that could have affected the control rod coupling !

integrity. The subsequent check is performed as a backup to the initial demon-stration. ,

In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position ,

indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of ;

rod reactivity which could be added by 1his small amount of rod withdrawal is ,

less than a normal withdrawal increment ad will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3 /4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm .in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL-POWER is greater than 10% of RATED THERMAL POWER, there is no possible rod worth which, of dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RWM to be Operable when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate control.

The RWM provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis ar presented in a topical report, l Reference 1, and two supplements, References 2 and 3. Additional. pertinent analysis is also contained in Amendment U iu the Reference 4 topical report.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density over the range of power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods. RBM OPERA-BILITY is required when the limiting condition described in Specification 3.1.4.3 exists.

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

REACTIVITY CONTROL SVSTEMS BASES _

3/4.1.5 STANDBY LIOUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel. To allow for potential leakage and improper mixing, this concentration is increased by 25%. The required concentration is achieved by having available a minimum quantity of 3,160 gallons of sodium pentaborate solution containng a minimum of 3,754 lbs of sodium pentaborate having the requisite B 10 atom

% enrichment of 29% as determined from Reference 5. This quantity of solution is a net amount which is above the pump suction shutoff level setpoint thus allowing for the ;

i portion which cannot be injected. The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cool s , and xenon decay. The temperature requirement ensures that the sodium ;

j pentaborate always remains in solution.

j With redundant pumps and explosive injection valves and with a highly '

reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

The SLCS system consists of three separate and independent pumps and i explosive valves. Two of the separate and independent pumps and explosive i valves are required to meet the minimum requirements of this technical I specification and, where applicable, satisfy the single failure criterion.

The SLCS must have an equivalent control capacity of 86 gpm of 13% weight sodium pentaborate in order to satisfy 10 CFR 50.62 (Requirements for ;

reduction of risk from anticipated transients without scram (ATWS) events for <

light-water-cooled nuclear power plants. As part of the ARTS /MELLL program the ATWS ,

analysis was updated to reflect the new rod line. As a result of this it was i determined that the Boron 10 enrichment was required to be increased to 29% to prevent exceeding a suppression pool temperature of 190*F. This equivalency requirement is fulfilled by having a system which satisfies the equation given in 4.1.5.b.2.

The upper limit concentration of 13.8% has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss of tank heating, which allow the solution to cool.

I 1 i LIMERICK - UNIT 1 B 3/4 1-4

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! REACTIVITY CONTROL SYSTEMS r

f BASES i

STANDBY LIOUID CONTROL SYSTEM (Continued) .

Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on I the temperature and volume once each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

.)

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i

1. C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis !

for Large BWR's," G. E. Topical Report NED0-10527, March 1972. 'l i

2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement I to NED0-10527, July 1972.

3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, " Exposed Cores."

Supplement 2 to NED0-10527, January 1973. ;

4. Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A,

" General Electric Standard Application for Reactor Fuel". '

5. " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," NEDC-32193P, July 1993.

LIMERICK - UNIT 1 B 3/4 1-5

3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10 CFR 50.46 and that the fuel design analysis limits specified in NEDE-240ll-P-A (Reference ?) will not be exceeded.

Mechanical Design Analysis: NRC approved methods (specified in Refer-ence 2) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 2. '

No single fuel rod follows, or is capable of following, this bounding power I history. This bounding power history is used as the basis for the fuel design analysis MAPLHGR limit.

LOCA Analysis: A LOCA analysis is performed in accordance with 10 CAR 50 Appendix K to demonstrate that the permissible planar power (MAPLHGR) limits i comply with the ECCS limits specified in 10 CAR 50.46. The analysis is pernemed for the most limiting break size, break location, and single failure combination '

for the plant, using the evaluation model described in Reference 9. l The MAPLHGR limit as showm in the CORE OPERATING LIMITS REPORT is the most limiting composite of the fuel mechanical design anaylsis MAPLHGR and the ECCS MAPLHGR !

limit. i Only the most limiting MAPLHGR values are shown in the CORE OPERATING :

LIMITS REPORT for multiple lattice fuel. Compliance with the specific lattice !

MAPLHGR operating limits, which are available in Reference 3, is ensured by use of the process computer.

l As a result of no longer utilizing an APRM trip setdown requirement, additional constraints are placed on the MAPLHGR limits to assure adherence to the fuel-mechanical i design bases. These constraints are introduced through the MAPFAC(P) and MAPFAC(F) l factors as defined in the COLR. !

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t POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRS at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The evaluation of a given transient begins with the system initial para-meters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior l transient computer program. The codes used to evaluate transients are discussed in Reference 2.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPR,, and MCPR,, respectively) to ensure adherence to fuel design limits during the worst transient that occurs l with moderate frequency (Ref. 6). Flow dependent MCPR limits are (MCPR,) are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 7) to analyze slow flow runout transients. The operating limit is dependent on the maximum : ore flow limiter setting in the Recirculation Flow Control System.

Power dependent MCPR limits (MPCR,) are determined mainly by the one dimensional transient code (Ref. 8). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPR , operating limits are provided for operating between 25% RTP and the previous y mentioned bypass power level.

The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR ie determined by the larger of the MCPR,, and MCPR, limits.

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f P_0HER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, cperating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow l when there have not been significant power or control rod changes. The requirement for l calculating MCPR when a limiting control rod pattern is approached l

ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

i 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

Reference:

1

1. Deleted. l

2. " General Electric Standard Application for Reactor Fuel,"

NEDE-240ll-P-A (latest approved revision).

3. " Basis of MAPLHGR Technical Specifications for Limerick Unit 1,"

NED0-31401, February 1987 (as amended).

4. Deleted

5. Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 1 Cycle 1, NEDC-31323, October 1986 including Errata and Addenda Sheet No.1 dated November 6,1986.

6. NEDC-32193P, " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," July 1993.

7. NED0-30130-A, " Steady State Nuclear Methods," May 1985.

8. NED0-24154, " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors, " October 1978.

9. HEDC-32170P, " Limerick Generating Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," June 1993.

LIMERICK - UNIT 1 B 3/4 2-5

3/4.4. REACTOR COOLANT SYSTEM BASES i

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3/4.4.1 RECIRCULATION SYSTEti The impact of single recirculation loop operation upon plant safety is '

s:sessed and shows that sing 1t-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and f 3.3.6-2, respectively.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive internals vibration. '

The surveillance on differential temperatures below 30% RATED THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on versel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode. ,

An inoperable jet pu:rp is not, in itself. a sufficient reason to declare ,

a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, '

the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation, f Recirculation pump speed mismatch limits are in compliance with the ECCS l LOCA analysis design criteria for two recirculation loop operation. The limits i will ensure an adequate core flow coastdown from either recirculation loop ;

following a LOCA. In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recir- {

culation loop mode.

i In order to prevent undue stress on the vessel nozzles and bottom head region, -

the recirculation loop temperature shall be within 50'F of each other prior to i startup of an idle loop. The loop temperature must also be within 50*F of the j reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation no:tles. Sudden equalization of a temperature !

difference > 145'F between the reactor vessel by increasing core flow rate would cause !

undue stress in the reactor vessel bottom head.

The objective of GE BWR plant and fuel design is to provide stable opera- !

tion with margin over the normal operating domain. However, at the high power / low I flow corner of the operating domain, a small probability of limit cycle neutron

flux oscillations exists depending on combinations of operating conditions l (e.g. , rod pattern, power shape). To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region Stability tests at operating BWRs were reviewed to determine a. generic region of the power / flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant to j plant variability of decay ratio with core and fuel designs. This generic region j has been determined to correspond to a core flow of less than or equal to 455 of ;

rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.

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, BASES 3/4.6.2 DEPRESSURIZATION SYSTF_MS The specifications of this section ensure that the primary containment pressure will not e.xceed the design pressure of 55 psig during primary system

,' blowdown from full operating pressure.

The suppression chamber water provides the heat sink for the reactor i coolant system energy release following a postulated rupture of the system.

The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from rated conditions. Since all of the gases in the drywell are purged into the suppression chamber air space during a loss-of-coolant accident, the pressure of the suppression chamber air space must not exceed 55 psig. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber..

Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design bpsis accident is below the design pressure. Maximum water volume of 134,600 ft3 results in a downcomer submergence of 12'3" and the minimum volume of 122,120 ft results in a submergence approximately

[ 2'3" less. The majority of the Bodega tests were run with a submerged length of 4 j' feet and with complete condensation. Thus. with respect to the downcomer submergence, l this specification is adequate. The maximum temperature at the end of the Llowdown tested during the Humboldt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperature above 170*F.

,. l l Should it be necessary to make the suppression chamber inoperable, this '

l shall only be done as specified in Specification 3.5.3.

l Under full power operating conditions, blowdown through safety / relief valves assuming an initial suppression chamber water temperature of 95*F results in a bulk water temperature of approximately 136*F immediately following blowdown which is below the 190*F bulk temperature limit used for complete condensation via T-quencher devices. At this temperature and atmospheric pressure, the avail-able NRSH exceeds that required by both the RHR and core spray pumps, thus there l is no dependency on containment overpressure during the accident injection phase.

If both RRR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.

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ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b. MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1. ;

c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3, I

d. The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,

e. LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4, .

f. The power biased Rod Block Monitor setpoints and the Rod Block Monitor MCPR OPERABILITY limits of Specification 3.3.6.

6.9.1.10 The analytical methods used to datermine the core operating limits :

shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

1

a. NEDE-240ll-P-A " General Electric Standard Application for Reactor fuel" (Latest approved revision).

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the ;

NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

LIMERICK - UNIT 1 6-18a l

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l INDEX DEFINITIONS l

PAGE l SECTION 1-1

! 1.1 ACTI0N.....................................................

1-1 1.2 AVERAGE PLANAR EXP05URE....................................

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................. r 1-1 1.4 CHANNEL CALIBRATION.........................................

1-1 i 1.5 CHANNEL CHECK..............................................

CHANNEL FUNCTIONAL TEST....................................

1-1 1.6 1.7 CORE ALTERATION............................................

1-2 1.7A CORE OPERATING LIMITS REP 0RT............................... 1-2 1.8 CRITICAL POWER RATI0....................................... 1-2 ,

1.9 DOSE EQUIVALENT I-131......................................

1-2 l

1.9a DOWNSC ALE TRI P SET POINT (DTSP) . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 l l 1.10 E-AVERAGE DISINTEGRATION ENERGY............................ 1-2 l

1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME......... 1-2 l

l 1.12 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.. 1-3 l

1.13 (DELETED).................................................. 1-3 1.14 (DELETED).................................................. 1-3 1.15 FREQUENCY N0TATION.......................................... 1-3 1.15a HIGH (POWER) TRIP SETPOINT (HTSP)........................... 1-3 l

1.16 IDENTIFIED LEAKAGE......................................... 1-3 1.16a INTERMEDIATE (POWER) TRIP SETPOINT (ITSP) ................. 1-3 l 1.17 ISOLATION SYSTEM RESPONSE TIME............................. 1-3 1.18 LIMITING CONTROL R0D PATTERN............................... 1-3 1.19 LINEAR HEAT GENERATION RATE................................ 1-4 1.20 LOGIC SYSTEM FUNCTIONAL TEST............................... 1-4 LIMERICK - UNIT 2 i l

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i'm7 - -- = _W -us- -W' m m -9 y v v v,mc w g' '

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SECTION PAGE DEFINITIONS (Continued) 1-4 1.20a LOW (POWER) TRIP SET POINT (LTSP)........................

1-4 1.21 (DELETED)...................................................

l 1-4 l

1.22 MEMBER (S) 0F THE PUBLIC....................................

l 1-4 l 1.22a MAPFAC(F) - (MAPLHGR FLOW FACT 0R)..........................

1.22b MAPFAC(p) - (POWER DEPENDENT MAPLHGR MULTIPLIER)........... 1-4 MINIMUM CRITICAL POWER RATIO (MCPR) . . . . . . . . . . . . . . . . . . . . . . . . .

1-4 ,

1.23 0FFSITE DOSE CALCULATION MANUAL............................

1-4 1.24 l

1-4 l 1.25 OPERABLE - OPERABILITY.....................................

l.26 OPERATIONAL CONDITION - CONDITION.......................... 1-5 ,

(

1-5 1.27 PHYSICS TESTS...............................................

1.28 PRESSURE BOUNDARY LEAKAGE..................................

1-5 ,

i 1.29 PRIMARY CONTAINMENT INTEGRITY..............................

1-5 1.30 PROCESS CONTROL PR0 GRAM.................................... 1-5 1.31 PURGE - PURGING............................................ 1-6 1.32 RATED THERMAL P0WER........................................ 1-6 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY.......... 1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME.................... 1-6 l 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY............ 1-6 1.36 REPORTABLE EVENT........................................... 1-7 1 -7 1.37 ROD DENSITY.................................................

1.38 SHUTDOWN MARGIN............................................ 1-7 1.39 SITE B0VNDARY.............................................. 1-7 1-7 1.40 (DELETED)..................................................

1.41 SOURCE CHECK............................................... 1-7 LIMERICK - UNIT 2 ii

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i INDEX DEFINITIONS PAGE SECTION DEFINITIONS (Continued) 1-8 1.42 STAGGERED T EST BAS I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-8 1.43 THERMAL P0WER..............................................

1-8

.l.43A TURBINE BYPASS SYSTEM RESPONSE TIME........................

1-8 1.44 UNIDENTIFIED LEAKAGE.......................................

1-8 1.45 UNRESTRICTED AREA..........................................

1-8 1.46 VENTILATION EXHAUST TREATMENT SYSTEM.......................

1.47 VENTING.................................................... 1-8 Tabl e 1.1 Surveill ance Frequency Notation. . . . . . . . . . . . . . . . . . . . . . . 1-9 ;

Tabl e 1. 2 Operati onal Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-10 i

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! LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS h

PAGE j SECTION i

i POWER DISTRIBUTION LIMITS (Continued) ,

I j 3/4 2.? (DELETED)............................................. 3/4 2-7 i

MINIMUM CRITICAL POWER RATI0.......................... 3/4 2-8 ,

3/4 2.3 Information on pages 3/4 2-10 thru 3/4 2-11 has been :

INTENTIONALLY OMITTED, refer to Note on page 3/4 2-10. 3/4 2-10 3/4.2.4 LINEAR HEAT GENERATION RATE........................... 3/4 2-12 t 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............. 3/4 3-1

[ Table 3.3.1-1 Reactor Protection System .

Instrumentation............... 3/4 3-2 I i

! Table 3.3.1-2 Reactor Protection Systea Instrumentation............... 3/4 3-6 l Table 4.3.1.1-1 Reactor Protection System ll Instrumentation Surveillance Requirements................. 3/4 3-7 !

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INDEX BASES !

SECTION PAGE ,

3!4.0 APPLICABILITY............................................. B 3/4 0-1 t

3/4.1 REACTIVITY CONTROL SYSTEMS 3.4.11 SHUTDOWN MARGIN....................................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.................................. B 3/4 1-1 i 3/4.1.3 CONTROL R0DS.......................................... B 3/4 1-2 3.4.1.4 CONTROL R0D PROGRAM CONTR0LS.......................... B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM......................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS i 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.................................................. B 3/4 2-1 LEFT INTENTIONALLY BLANK........................................ B 3/4 2-3 3/4.2.2 (DELETED)............................................. B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RAT 10.......................... B 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE........................... B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............. B 3/4 3-1 ;

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................. B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION....................................... B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.. . .. B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION l INSTRUMENTATION....................................... B 3/4 3-4 3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION..................... B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.................. B 3/4 3-4 LIMERICT - UNIT 2 xviii 1

r DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of i fuel, sources, or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs, IRMs, TIPS, or special movable detectors is not considered a .

Suspension of CORE ALTERATIONS shall not preclude ;

CORE ALTERATION.

I completion of the movement of a component to a safe conservative position.

! CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload 4

cycl e. These cycle-specific core operating limits shall be determined :

for each reload cycle in accordance with Specifications 6.9.1.9 thru ,

6.9.12. Plant operation within these limits is addressed in individual specifications. l r

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the ;

assembly which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. ,

DOSE EOUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per '

l gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. t The thyroid dose conversion factors used for this calculation shall be i those listed in Table III of TID-14844, " Calculation of Distance Factors i for Power and Test Reactor Sites." '

i DOWNSCALE TRIP SET POINT (DTSP1 !

1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) !

rod block trip setting.

i E-AVERAGE DISINTEGRATION ENEPCd 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, i for isotopes, with half lives greater than 15 minutes, making up at least l 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCSI RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time !

interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

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DEFINITTONS -

END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME I 1.12 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be

! that time interval to complete suppression of the electric are between the fully open contacts of the recirculation pump circuit breaker from I initial movement of the associated:

Turbine stop valves, and j I a.

b. Turbine control valves. l This total system response time consists of two components, the instrumen- [

tation response time and the breaker arc suppression time. These times may be measured by any series of sequential, overlapping or total steps 1 I

l such that the entire response time is measured.

1.12 (Deleted) !

p 1.13 (Deleted) l FRE00ENCY NOTATI0ff 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance ,

Requirements shall correspond to the intervals defined in Table 1.1.

HIGH (POWER) TRIP SET POINT (HTSP) 1.15a The high power trip setpoint associated with the Rod Block Monitor (RBM) l t

rod block trip setting applicable above 85% reactor thermal power.

I IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAVAGE shall be: !

!. a. Leakage into collection systems, such as pump seal or valve packing i

> leaks, that is captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

INTERMEDIATE (POWER) TRIP SET POINT (ITSP) 1.16a The intermediate power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable between 65% and 85% reactor thermal power. i ISOLATION SYSTEM RESPONSE TIME 1.17 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when '

the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. i Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. <

LIMITING CONTROL R0D PATTERN 1.18 A LIMITING CONTROL R0D PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, OR MCPR.

LINEAR HEAT GENERATION RATE 1.19 LINEAR HEAT GENERAT'ON RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LIMERICK - UNIT 2 1-3

DEFINITIONS i i LOGIC SYSTEM FUNCTIONAL TEST 1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated ,

device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be :

performed by any series of sequential, ovarlapping or total system steps such that the entire logic system is tested. l LOW (POWER) TRIP SET POINT (LTSP) 1.20a The low power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable between 30% and 65% reactor thermal power.

1.21 (Deleted) l MEMBER (5) 0F THE PUBLIC 1.22 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of i the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. j This category does include persons who use portions of the site for recrea- j tional, occupational, or other purposes not associated with the plant. t MAPFAC(F)-(MAPLHGR FLOW FACTOR) l 1.22a A core flow dependent multiplication factor used to flow bias the standard l Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit. >

MAPFAC(P)-(POWER DEPENDENT MAPLHGR MULTIPLIER) 1.22b A core power dependent multiplication factor used to power bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit. j

( MINIMUM CRITICAL POWER RATIO (MCPR) ;

1.23 The MAXIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core (for each class of fuel). Associated with the minimum critical power ratio is a core flow dependent (MCPR(F)) and core power dependent (MCPR(P)) minimum critical power ratio.

OFFSITE DOSE CALCULATION MANUAL 1.24 The 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology i and parameters used in the calculation of offsite doses resulting from

radioactive gaseous cnd liquid effluent, in the calculation of gaseous

! and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall .

also contain (1) the Radioactive Effluent Controls and Radiological '

Environmental Monitoring Programs required by Section 6.8.4 and (2)

descriptions of the information that should be included in the Annual l radiological Environmental Operating and Semi-annual Radioactive Effluent

Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.

! o OPERABLE - OPERABILITY l 1.25 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s) and I when all necessary attendant instrumentation, controls, electrical power, l cooling or seal water, lubrication or other auxiliary equipment that are l required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

LIMERICK - UNIT 2 1-4 I

l DEFINITIONS OPERATIONAL CONDITION - CONDITION l.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive ;

combination of mode switch position and average reactor coolant tempera-i ture as specified in Table 1.2.

^

PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental l nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14 of the FSAR, (2) authorized under the ,

provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. t PRESSURE BOUNDARY LEAKAGE

! 1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel wall.

l PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All primary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or l

[

! 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, .

except as provided in Table 3.6.3-1 of Specification 3.6.3.

1

b. All primary containment equipment hatches are closed and sealed. ,

c. The primary containment air lock is in compliance with the requirements of Specification 3.6.1.3. ,

d. The primary containment leakage rates are within the limits of Specification 3.6.1.2. i

e. The suppression chamber is in compliance with the requirements !

! of Specification 3.6.2.1.

with each primary containment i

f. The sealing mechanism associated penetration; e.g., welds, bellows, or 0-rings, is OPERABLE.

PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure that the solidification or dewatering and packaging of radioactive wastes ,

results in a waste package with properties that meet the minimum and stability requirements of 10 CFR Part 61 and other requirements for trans- '

portation to the disposal site and receipt at the disposal site.

With SOLIDIFICATION or dewatering, the PCP shall identify the process parameters influencing solidification or dewatering, based on laboratory scale and full scale testing or experience.

l LIMERICK - UNIT 2 1-5 l

DEFINITIONS PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER i

1.32 RATED THERMAL POWER shall be a total reactor ore heat transfer rate to the reactor coolant of 3293 MWT. l REACTOR ENCLOSURE SECONDARY CONTAIUMENT INTEGRITY 1.33 REACTOR ' ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when: ,

a. All reactor enclosure secondary containment penetrations required to [

be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment !

l automatic isolation system, or l

! 2. Closed by at least one manual valve, blind flange, slide gate !

l damper, or deactivated automatic valve secured in its closed '

position, except as provided in Table 3.6.5.2.1-1 of !

Specification 3.6.5.2.1. ;

! b. All reactor enclosure secondary containment hatches and blowout panels !

are closed and sealed. ,

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c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

d. The reactor enclosure recirculation system is in compliance with the (

requirements of Specification 3.6.5.4. I

e. At least one door in each access to the reactor enclosure secondary j containment is closed. l

f. The sealing mechanism associated with each reactor enclosure secondary !

containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE. !

t

g. The pressure within the reactor enclosure secondary containment is :

less than or equal to the value required by Specification 4.6.5.1.la. l REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from '

when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. !

)

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY l.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrati.ons required to ;

be closed during accident conditions are either: !

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4 DEFINITIONS REFUEllNG FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)

1. Capable of being closed by an OPERABLE secondary containment :

automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.5.2.2-1 of Specification 3.6.5.2.2.

All refueling floor secondary containment hatches and blowout panels are b.

closed and sealed.

c. The standby gas' treatment system is in compliance with the requirements ,

of specification 3.6.5.3. {

At least one door in each access to the refueling floor secondary d.

containment is closed. ;

1 l e. The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.

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f. The pressure within the refueling floor secondary containment is less l

than or equal to the value required by Specification 4.6.5.1.2a.

REPORTABLE EVENT l 1.36 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. ,

ROD DENSITY ;

1.37 R0D DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% R0D DENSITY.

1 SHUTDOWN MARGIN 1.38 SHUTDOWN MARGIN shall be the amount of reactivity- by which the reactor is i subcritical or would be subcritical assuming all control rods are fully l

inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.

l SITE BOUNDARY l

1.39 The SITE. B0UNDARY shall be that line as defined in Figure 5.1.3.la.

1.40 (Deleted)

SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

LIMERICK - UNIT 2 1-7 I

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TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE TRIP SETPOINT VALUES FUNCTIONAL UNIT r 1. Intermediate Range Monitor, Neutron Flux-High s 120/125 divisions s 122/125 divisions c of full scale of full scale

$ 2. Average Power Range Monitor:

s 15% of RATED THERMAL POWER 5 20% of RATED M a. Neutron flux-Upscale, Setdown

" THERMAL POWER

' b. Neutron Flux-Upscale E!

M 1) During two recirculation loop operattun:

u a) Flow Biased s 0.66 W+ 65%, with 5 0.66 W+ 68%, with l a maximum of a maximum of b) High Flow Clamped s 115% of RATED s 117% of RATED l THERMAL POWER THERMAL POWEP

2) During single recirculation loop operation:

Flow Biased 5 0.66 W+ 59%, s 0.66 W+ 62%, l a) Not Required b) High Flow Clamped Not Required OPERABLE OPERABLE

c. Inoperative N.A. N.A.

o A d. Downscale 2 4% of RATED 2 3% of RATED THERMAL POWER THERMAL POWER

3. Reactor Vessel Steam Dome Pressure - High s 1037 psig s 1057 psig

4. Reactor Vessel Water Level - Low, Level 3 2 12.5 inches above instrument 2 11.0 inches above zero* instrement zero

5. Main Steam Line Isolation Valve - Closure s 8% closed s 12% closed

! 6. Main Steam line Radiation - High s 3.0 x full power background s 3.6 x full power background

7. Drywell Pressure - High 5 1.68 psig s 1.88 psig

8. Scram Discharge Volume Water Level - High

a. Level Transmitter s 261' 1 1/4" elevation ** s 261' 9 1/4" elevation

b. Float Switch s 261' 1 1/4" elevation ** s 261' 9 1/4" elevation o See 545es Figure B 3/4.3-1.

oo Ect ,:st to 25.58 gallons / scram discharge volume.

- _.m_._ _ _ _ . _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ - - .-c.- ~ . _ _ _ _ . . - ~ + _ . _ m __ _.._ - .-_ m__ _ _ _ _ _ _-_ .m

t REACTIVITY CONTROL SYSTEMS t

I ROD BLOCK MONITOR l j l J.IMITING CONDITION FOR OPERATION l

l 3.1.4.3 Both rod block monitor (RBM) channels shall be OPERABLE.

APPLICABILITY: OPERATIONA', CONDITION 1, when THERMAL POWER is greater than or !

equal to 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER with MCPR less than 1.70, or THERMAL POWER greater than or equal to 90% of rated with

! MCPR less than 1.40.

I I

! ACTION: )

I '

! a. With or.e RBM channel inoperable:

1

1. Verify that the reactor is not operating on a I.IMITING COETROL ROD PATTERN, and L ,

f; 2. Restore the inoperable RBM channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l 1

l; Otherwise, place the inoperable . rod block monitor channel in the l tripped condition within the next hour, j b. With both RBM channel inoperable, place at leas one inoperable rod block monitor channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . I

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I l; pURVEILLANCE REOUIREMENTS ,

)

1' 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE l by performance of a:

a. CHANNEL FUNCTIONAL D i CHANNEL CALIBRATION at the frequencies and for the OPERATIOT ' ITIONS specified in Table 4.3.6-1.

b. CHANNEL FUNCTIONAL TEST prior to control rod withdrawal when the reactor is operating on a LIMITING CORTROL ROD PATTERN.

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REACTIVITY CONTROL SYSTEMS '

3/4.1.5 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system consisting of a minimum of two pumps and corresponding flow paths, shall be OPERABLE. l APPLICABILITY: OPERATION CONDITIONS 1, 2, and 5*

ACTION:

a. In OPERATIONAL CONDITION 1 or 2:

1. With only ona pump and corresponding explosive valve OPERABLE, restore one inoperable pump and corresponding explosive valve 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 .

2. With standby liquid control system otherwise inoperable, restore the system 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 .

b. In OPERATION CONDITION 5*:

1. With only one pump and corresponding explosive valve OPERABLE, restore one inoperable pump and corresponding explosive valve r to OPERABLE status within 30 days or insert all insertable control rods within the next hour. <

2. With the standby liquid control system otherwise inoperable, ;

insert all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . !

SURVEILLANCE REOUIREMENTS ,

I 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

l l a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that:

1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2. The available volume of sodium pentaborate solution is at least 3160 gal'<.ns. l

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.

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

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REACTIVITY CONTROL SYSTEMS -

U EURVEILIANCE REOUIREMENTS (Continued)

b. At least once per 31 days by: ,

1. Verifying the continuity of the explosive charge.

2. Determining by chemical analysis and calculation

that the available weight of sodium pentaborate is greater than or equal ,

to 3754 lbs; the concentration of sodium pentaborate in solution l is less than or equal to 13.8% and within the limits of Figure 3.1.5-1 and; the following equation is satisfied: ,

C x E X 0 21 13% wt. 29 atom % 86 gpm where C - Sodium pentaborate solution (t by weight) '

Q - Two pump flowrate, as determined per surveillance requirement 4.1.5.c.

E - Boron 10 enrichment (atom % Boron 10) l ,

3. Verifying that each valve (manual, power-operated, or automatic) f in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1190 psig is met.

d. At least once per 18 months during shutdown by:

1. Initiating at least one of the standby liquid control system j loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch I which has been certified by having one of the batch success-fully fired. All injection loops shall be tested in 3 operating cycles.

2. Verify all heat-treated piping between -torage tank and pump suction is unblocked.**

e. Prior to addition of Boron to storage tan'k verify sodium pentaborate l enrichment to be added is 2 29 atom % Boron 13.

1

This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below the limits of Figure 3.1.5-1 for ,

the most recent concentration analysis, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or boron j addition or solution temperature is restored.

- This test shall also be performed whenever suction piping temperature drops below the limits of Figure 3.1.5-1 for the most reent concentration analysis, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored.

LIMERICK - UNIT 2 3/4 1-20 1 r----- , ,w , g& c------.-wg, v.-_-- -.y -- y<.-9 ,-9...-. yn,* y9 >+% .. --,-sm- sp 9 q --s

3/4.2 POMER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of axial location and AVERAGE PLANAR EXPOSURE shall be within limits based on applicable APLHGR limit values which have been determined by approved methodology for the respective fuel and lattice types. When hand calculations are required, the APLHGR for each type of fuel as a function of AVERAGE l l PLANAR EXPOSURE shall not exceed the limiting value for the most limiting lattice 1 (excluding natural uranium) as shown in the CORE OPERATING LIMITS REPORT (COLR).

During operation, the APLHGR for each fuel type shall not exceed the above values multiplied by the appropriate reduction factors for power and flow as defined in the COLR.

APPLICABILITY: OPERATIONAL CONDITION I, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER >

ACTION:

With an APLHGR exceeding the limiting value, initiate corrective action within l 15 minutes and restore APLHGR to within tbt *equired limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or I

reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REOUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limiting value.

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R0D PATTERN for APLHGR.

d. The provisions of Specification 4.0.4 are not applicable.

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Section 3/4.2.2 (DELETE 01 ,

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INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETED I

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POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO ,

LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATION (MCPR) shall be equal to or greater than :

the rated MCPR limit adjusted by the MCPR(P) and 97.PR(F) factors as shown in the i CORE OPERATIONS LIMITS REPORT, provided that the end-of-cycle recirculation pump trip (E0C-RPT) system is OPERABLE per Specification 3.3.4.2 and the main turbine bypass system is OPERABLE per Specification 3.7.8, with:

7- Pave *B)

'A 'B where:

'A = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3,

'B = 0.672 + 1.65 ( "I )'/'(0.016),

n N, l E i i=I n

I

' ave = i =1 "i'i ,

e n N, I

i-I n = number of surveillance tests performed to date in cycle, !

N' = number of active control rods measured in the i" surveillance test, r,= average scram time to notch 39 of all rods measured in the i" surveillance test, and N3 = total number of active rods measured in Specification 4.1.3.2.a.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

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r POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) -l ACTION

a. With the end-of-cycle recirculation pump trip system inoperable per !

Specification 3.3.4.2, operation may continue provided that, within I hour, MCPR is determined to be greater than or equal to the rated MCPR l limit as a function of the average scram time shown in the appropriate figure in the CORE OPERATING LIMITS REPORT, for E0C-RPT inoperable curve, adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT.

b. With MCPR less than the applicable MCPR limit adjusted by the MD'l (P) and )

MCPR(F) factors as. shown in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes ~ and restore MCPR to within the t

i required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. With the main turbine bypass system inoperable per. Specification 3.7.8, operation may continue provided that, within I hour, MCPR is determined ,

to be greater than or equal to the ated MCPR limit as a function of the l average scram time (shown in the CORE OPERATING LIMITS ' REPORT) main ,

turbine bypass valve inoperable curve, adjusted by the MCPR(P) and MCPR(F) ~

i' factors as shown in the CORE OPERATING LIMITS REPORT.

SURVEILLANCE RE0VIREMENTS :

4.2.3 NCPR, with:

a. r = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or

b. r as defined in Specification 3.2.3 used to determine the limit within~ 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, 3 shall be determined to be equal to or greater than the applicable MCPR limit including application of the MCPR(P) and MCPR(F) factors as determined from the CORE OPERATING LIMITS REPORT.

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and l c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

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d. The provisions of Specification 4.0.4 are not applicable.

LIMERICK - UNIT 2 3/4 2-9 l

I i__._. _ __. . _ _ .. _ . - . . . _ , . _ . _ _ . . _ _ . . , - , _ _ , . . . , . . . _ .

TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEllLANCE REQUIREMENTS C

CHANNEL OPERATIONAL l5 5 CilANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH R FUNCTIONAL UNIT CHECK TEST CALIBRATION M f.'IRVEILLANCE RE0VIRED

9. Turbine Stop Valve - Closure N.A. Q R 1

10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N.A. Q R 1

11. Reactor Mode Switch Shutdown Position N.A. R N.A. 1,2,3,4,5

12. Manual Scram N.A. W -N.A. 1,2,3,4,5 j (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

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

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

This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by Y

(d) a heat balance during 0PERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

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

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

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

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

(i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. -

(j) If the RPS shorting links are required to be removed per specification 3.9.2, they may be reinstalled for up 'to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall

.be moved from its existing position.

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

I

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

TABLE 3.3.6-1 (Continued) i CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ACTION 60 - Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.

ACTION 61 - With the number of OPERABLE channels one or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.

ACTION 62 - With the number of OPERABLE channels less than required by the '

Minimum OPERABLE Channels per Trip Function requirements, place the inoperable channel in the tripped condition within one hour.

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

NOTES

For OPERATIONAL CONDITION of specification 3.1.4.3.

- With more than one control rod withdrawn. Not applicable to control ' rods ,

removed per specification 3.9.10.1 or 3.9.10.2.

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

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

(b) This function shall be automatically bypassed if detector count rate is

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

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

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

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

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

l l

LIMERICK - UNIT 2 3/4 3-59 i

i

INR L 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP SETPolNT ALLOWABLE VALUE TRIP FUNCTION

1. ROD BLOCK MONITOR

a. Upicale h 1) Low Trip Setpoint (LTSP) *

g 2) Intermediate Trip Setpoint (ITSP) *

y 3) liigh Trip Setpoint (HTSP)

' b. I V.erative N/A N/A

@ c. Downscale (DTSP) y d. Power Range Setpoint

26% RATED THERMAL POWER Low Power Setpoint (LPSP) 23% RATED THERMAL POWER y 1) 61% RATED TilERMAL POWER

2) Intermediate Power Setpoint (IPSP) 58% RATED THERMAL POWER 78% RATED THERMAL POWER 81% RATED THERMAL POWER

3) High Power Setpoint (HPSP)

2. APRM

a. Flow Biased Neutron Flux - Upscale 5 0.66 W + 51%* 5 0.66 W + 55%* l 5 1) ' During two recirculation loop operation During single recirculation s 0.66 W + 45% $ 0.66 W + 49%* l Y 2) 8 loop operation Inoperative N.A. N.A.

b.

c. Downscale 2 4% of RATED THERMAL POWER 2 3% of RATED THERMAL POWER

d. Neutron Flux - Upscale, Startup s 12% of RATED TiiERMAL POWER $ 14% of RATED THERMAL POWER

3. SOURCE RANGE MONITORS N.A. N.A.

a. Detector not full in s 1 x 10' cps 5 1.6 x 10' cps

b. Upscale N.A. N.A.

c. Inoperative 2 1.8 cps **

d. Downscal e 2 3 cps **

CONTROL R00 BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 4 INTERMEDIATE RANGE MON 110RS e-. a. Detector not full in N.A. N.A.

E b. Upscale s 108/125 divisions of s 110/125 divisions of la full scale full scale M c. Inoperative N.A. N.A.

" d. Downscale 2 5/125 divisions of full 2 3/125 divisions of full

' scale scale E5. SCRAM DISCHARGE VOLUME G a. Water level-High $ 257' 7 3/8" elevation *** s 257" 9 3/8" elevation w a. Float Switch

6. REACTOR COOLANT SYSTEM RECIRCULATION f.LOH

a. Upscale s 111% of rated flow s 114% of rated flow

b. Inoperative N.A. N.A.

, y c. Comparator s 10% flow deviation s 11% flow deviation

7. REACTOR MODE SWITCH SHUTDOWN i w POSITION N.A. N.A.

4

Refer to the COLR for these setpoints. l

- May be reduced provided the source range monitor has an observed count rate and signal-to-noise ratio on or above the curve shown in Figure 3.3.6-1.

]

- - Equivalent to 13.56 gallons / scram discharge volume.

(a) There are three upscale trip levels. Each is applicable only over its specified operating core thermal power range. All RBM tr!ps are automatically bypassed below the low power setpoint (LPSP). The upscale LTSP is applied between the low power setpoint (LPSP) and the intermediate power cetpoint (IPSP). The upscale ITSP . is applied between the intermediate power setpoint and the high power setpoint (DSP). The HTSP is applied above the high power setpoint.

i (b) Power range setpoints control enforcement of appropriate upscale trips over the proper core thermal power ranges. The power signal to the RBM is provided by the APRM.

. . _ . . _ _ _ _ _ . _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _ _ . _ _ . _ _ _ . _ _ _ _ _ _ - _ _ _ ___.____ ____________m m2 . - . - - . - . - ._ -, .. . , . . . , w_. ..r., - .w -

1ABLE 4.3.6-1 i

l CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CllANNEL OPERATIONAL

[ CilANNEL FUNCTIONAL CilANNEL CONDITIONS FOR WHICH 4

TRIP FUNCTION CHECK TEST CAllDRATION"' SURVEILLANCE RE0VIRE ,

j 1. R00 BLOCK MONITOR

a. Upscale ,, N.A. S/U*"",Q"'

t R 1*

Inoperative N.A. S /U (*"" , Q"' N.A. 1*

b. l '

-c. . Downscale N.A. S/Ut *"", Q"' R 1*

i

, .2. APRM

-a. Flow Biased Neutron Flux-Upscale N.A. S/U(*) , Q SA 1 1

b. - Inoperative N.A. S/U("),Q N.A. 1, 2, 5***

c. Downscale N.A. S/U t *' , Q SA 1 l E d. Neutron Flux - Upscale, Startup N.A. t S/U*),Q SA ' 2, 5***

3. SOURCE-RANGE MONITORS

. r S/U(*),W N.A. t k

5 a.-

b.

' Detector not full in Upscale N.A.

N.A. S/U(*),W SA 2, 5 2, 5 Inoperative N.A. S/U(*) N.A. 2, : 5 i -p c. S/U(*),W

d. Downscale. 'N.A. , W SA 2, 5 l4. INTERMEDIATE RANGE MONITORS

h H a.

b.

- Detector not full in

. Upscale

.N.A.

N. A. -

S/U(*),W S/U(*),W N.A.-

SA-2, 5 2, 5

-' c. Inoperative N.A. S/U(*),W N.A. 2, . 5

" d. Downscale N.A. S/U(*),W SA - 2, 5 .

<5. SCRAM DISCHARGE VOLUME I

a. Water Level - High N.A.- .Q R 1, 2, 5**

l 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

\.w

! 3 a. Upscale N.A. S/U(*), Q .SA- -I

b. Inoperative N. A. - S/Ut*) Q N.A. 1

c. Comparator N.A. S/U(*),Q SA 1

w ,

{

7. REACTOR MODE SWITCH' SHUTDOWN POSITION N.A. R N.A. 3, 4 -

i i e-..---.em.--,*s - _ _ ...,.4 s e.-,---+---.-s.-...m~..m.,

- a = - -e, ,- --- <- -. ww- ,--.-w>>..-v...----..m.- ...--=,--...--...-.,,-..--...-.._-.-r-m.-.-.+ -- m-m_-.--__.-_-m_m__-

TABLE 4.3.6-1 (Continued) - t j

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REOUIREMENTS i

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

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

(c) Includes reactor manual control multiplexing system input.

For OPERATIONAL CONDITION of Specification 3.1.4.3. !

With more than one control rod withdrawn. Not applicable to control j rods removed per Specification 3.9.10.1 or 3.9.10.2. ,

i i

j

- - Required to be OPERABLE only prior to and during shutdown margin demonstrations.as performed per Specification 3.10.3. l 6

I F

i LIMERICK - UNIT 2 3/4 3-62 l

3/4.4 REACTOR C001 ANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION iDOPS LIMITING CONDITION FOR OPERATION _

Two reactor coolant system recirculation loops shall be in operation 3.4.1.1

.. i th :

a. Total core flow greater than or equal to 45% of rated core flow, or

b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. Vith one reactor coolant syttem recirculation loop not in operation:

1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ;

a. Place the recirculation flow control system in the local Manual mode, and

b. Reduce THERMAL POWER to 5 70% of RATED THERMAL POWER, and

c. Limit the speed of the operating recirculation pump to l 1ess than or equal to 90% of rated pump speed, and

d. Verify that the differential temperature requirements of l Surveillance Requirement 4.4.1.1.5 are met if THERMAL POWER is 5 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is 5 50% of rated loop flow, or suspend the THERMAL POWER or recirculation loop flow increase.

See Special Test Exception 3.10.4.

LIMERICK - UNIT 2 3/4 4-1

REACTOR COOLANT SYSTEM l LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

2. Within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s:

Reduce the Average Power Range Monitor (APRM) Scram and Rod ,

i Block Trip Setpoints and Allowable Values, to those applicable for

single recirculation loop operation per Specifications 2.2.1 and 3.3.6, or declare the associated channel (s) inoperable and take the actions required by the referenced specifications, and,

3. The provisions of Specification 3.0.4 are not applicable.

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

4 b. With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER such that it is i not within the restricted zone of Figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With one or two reactor coolant system recirculation loops in operation and total core flow less than 45% but greater than 39% of rated core flow and THERMAL POWER within the restricted zone of Figure 3.4.1.1-1:

1. Determine the APRM and LPRM** noise levels (Surveillance 4.4.1.1.3):

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

b. Within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.

2. With the APRM or LPRM** neutron flux noise levels greater than three times their established baseline noise levels, within 15 minutes initiate corrective action to restore the noise levels within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow or by reducing THERMAL POWER.

d. With one or two reactor coolant system recirculation loops in .

operation and total core flow less than or equal to 39% and THERMAL l POWER within the restricted zone of Figure 3.4.1.1-1, within 15 minutes initiate corrective action to reduce THERMAL POWER to within the unrestricted zone of Figure 3.4.1.1-1 or increase core flow to i greater than 39% within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . I

- Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

LIMERICK - UNIT 1 3/4 4-la 1

- - _ . ~ ____, ._ , .I

i LIMITING SAFETV 5 JM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continuedl Averaae Power Ranae Monitor (Continued)

< Because the flux distribution associated with uniform rod withdrawals does not I involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more l

l than adequate to assure shutdown before the power could exceed the Safety Limit. .

! The 15% neutron flux trip remains active until the mode switch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the l

system and therefore the monitors respond directly and quickly to changes due i to transient operation for the case of the Neutron Flux-Upscale flow bias setpoint; i.e., for a power increase, the THERMAL power cf the fuel will be less than that indicated by the neutron flux due to the time constants of the heat l

transfer associated with the fuel.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.

3. Reactor Vessel Steam Dome Pressure-Hiah High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine stop valve and control fast closure trips are bypassed. For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

LIMERICK - UNIT 2 B 2-7 )

1

)

REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) l Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demon-stration.

In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3/4.1.4 CONTROL R0D PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control !

rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RWM to be Operable when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate control.

The RWM provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3. Additional pertinent analysis is also contained in Amendment 17 to the Reference 4 topical report.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density over the range of l power operation. Two channels are provided. Tripping one of the channels will i

block erroneous rod withdrawal to prevent fuel damage. This system backs up the l written sequence used by the operator for withdrawal of control rods. RBM OPERA-BILITY is required when the limiting condition described in Specification 3.1.4.3 exists.

LIMERICK - UNIT 2 B 3/4 1-3 l -- _. - - . .. -.

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIOUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concen- ;

i tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel. To allow for potential leakage and improper mixing, this con- l centration is increased by 25%. The required concentration is achieved by having !

i available a minimum quantity of 3,160 gallons of sodium pentaborate solution l containing a minimum of 3,754 lbs of sodium pentaborate having the requisite B 10 atom

% enrichment of 29% as determined from Reference 5. This quantity of solution is a net amount which is above the pump suction shutoff level setpoint thus allowing for the !

portion which cannot be injected. The pumping rate of 41.2 gpm provides a negative I reactivity insertion rate over the permissible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, l

increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator ,

cools, and xenon decay. The temperature requirement ensures that the sodium pentaborate always remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

The SLCS system consists of three separate and independent pumps and explosive valves. Two of the separate and independent pumps and explosive valves are required to meet the minimum requirements of this technical specification l and, where applicable, satisfy the single failure criterion.

The SLCS must have an equivalent control capacity of 8G gpm of 13% weight sodium pentaborate in order to satisfy 10 CFR 50.62 (Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants. As part of the ARTS /MELLL program the ATWS analysis was updated to reflect the new rod line. As a result of this it was determined that the Baron 10 enrichment was required to be increased to 29% to prevent exceeding a l suppression pool temperature of 190 *F. This equivalency requirement is fulfilled by having a system which satisfies the equation given in 4.1.5.b.2.

The upper limit concentration of 13.8'i, has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss of tank heating, which allow the solution to cool.

t i

\ l

)

l l

l LIMERICK - UNIT 2 8 3/4 1-4

REACTIVITY CONTROL SYSTEMS BASES

~

3/4.1.5 STANDBY LIOUID CONTROL SYSTEM (Contined)

Surveillance requirements are established on a frequency that assures a high r reliability of the system. Once the solution is established, boron concentration will not vary unless more bcron or water is added, thus a check on the temperature ;

and volume once each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.

i t Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges. ;

i i

T l

1. C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NED0-10527, March 1972.

2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement I to NED0-10527, July ,

1972.

3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, " Exposed Cores." ,

Supplement 2 to NED0-10527, January 1973. l Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, l 4.

" General Electric Standard Application for Reactor Fuel".

" Maximum Extended Load Limit and ARTS Improvement Program Analyses for !

5.

Limerick Generating Station Units 1 and 2," NEDC-32193P, July 1993. j LIMERICK - UNIT 2 B 3/4 1-5 l

l

e i

THIS PAGE INTENTIONALLY LEFT BLANK r

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3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10 CFR 50.46 and that the fuel design analysis limits specified in NEDE-240ll-P-A (Reference 2) will not be exceeded. -

Mechanical Design Analysis: HRC approved methods (specified in :

Reference 2) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 2. No single fuel rod follows, or is capable of following, this l bounding power history. This bounding power history is used as the basis for the fuel design analysis MAPLHGR limit.

LOCA Analysis: A LOCA analysis is performed in accordance with 10CFR50 Appendix K to demonstrate that the permis','Sle planar power (MAPLHGR) limits comply with the ECCS limits specified in 10t'D50.46. The analysis is performed for the most limiting break size, break le ation, and single failure combination for the plant, using the evaluation model duscribed in Reference 9. l The MAPLHGR limit as shown in the CORE OPERATING LIMITS REPORT is the ;

most limiting composite of the fuel mechanical design analysis MAPLHGR and the ECCS MAPLHGR limit. ,

Only the most limiting MAPLHGR values are shown in the CORE OPERATING LIMITS REPORT for multiple lattice fuel. Compliance with the specific lattice MAPLHGR operating limits, which are available in Reference 3, is ensured by use of the process computer.

As a result of no longer utilizing an APRM trip setdown requirement, additional constraints are placed on the MAPLHGR limits to assure adherence to the fuel-mechanical l design bases. These constraints are introduced through the MAPFAC(P) and MAPFAC(F) factors as defined in the COLR.

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

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POMER DISTRICUTION LIMITS BASES k

3/4.2.2 (DELETED) f INFORMATION CONTAINED ON I

THIS PAGE HAS BEEN DELETED l

)

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LIMERICK - UNIT 2 B 3/4 2-2 I

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i .

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRS at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The evaluation of a given transient begins with the system initial pap-meters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate transients are discussed in Reference 2.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPR,, and MCPR,, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Ref. 6). . Flow dependent MCPR limits are (MCPR,) are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 7) to i analyze slow flow runout transients. The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

Power dependent MCPR limits (MPCR,) are determined mainly by the one dimensional transient code (Ref. 8). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPR , operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPR,, and MCPR, limits.

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

l E. -- - -- . _ _ _

l

POWER DISTRIBUTION l!MITS BASES MINIMUM CRITICAL POWER RATIO (Continued) l t

i At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be cperating at minimum recirculation pump speed ano the moderator !

void content will be very small. For all designated control rod patterns which may !

be employed at this point, operating plant experience indicates that the resulting l MCPR value is in excess of requirements by a considerable margin. Durir.g initial :

start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL j POBER level with minimum recirculation pump speed. The MCPR margin will thus be ;

demonstrated such that future MCPR evaluation below this power level will be shown to i be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is ;

greater than or equal to 25% of RATED THERMAL POWER is sufficient since power 4 distribution shifts are very slow when there have not been significant power or [

control rod changes. The requirement for calculating MCPR when a limiting control i rod pattern is approached ensures that MCPR will be known following a change in l THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3_f4.2.4 llNEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) {

in any rod is less than the design linear heat generation even if fuel pellet l densification is postulated. !

Reference:

1. Deleted. !

2. " General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A (latest approved revision). !

i

3. " Basis of MAPLHGR Technical Specifications for Limerick Unit 1," ;

NED0-31401, February 1987 (as amended). !

4. Deleted

5. Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 1 Cycle 1, NEDC-31323, October 1986 :

including Errata and Addenda Sheet No. I dated November 6,1986. l

6. NEDC-32193P, " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," July 1993.

7. NED0-30130-A, " Steady State Nuclear Methods," May 1985.

8. NEDD-24154, " Qualification of the One-Dimensional Core Transient Mo61 for Boiling Water Reactors, " October 1978.

9. NEDC-32170P, " Limerick Generating Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Caolant Accident Analysis," June 1993. l l

LIMERICK - UNIT 2 B 3/4 2-5 l

3/4.4. REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM lhe impact of single recirculation loop operation upon plant safety is ,

assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and !

3.3.5-2, respectively.

Asitionally, surv:illance on the pump speed of the operating )

recirculation loop is imposed to exclude the possibility of excessive internals vibration. The surveillance on differential temperatures below 30%

RATED THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; '

thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure '.an be detected by monitoring jet pump performance on a pre-scribed schedule for significant degradation.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In the case where the mismatch limits cannot be maintained i during two loop operation, continued operation is permitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperature shall be within 50 F of each other prior to startup of an idle loop. The loop temperature must also be within 50'F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Sudden equalization of a temperature difference > 145 F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

The objective of GF BWR plant and fuel design is to provide stable opera-tion with margin over the normal operating domain. However, at the high power / low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g., rod pattern, power supe). To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region. 1 Stability tests at operating BWRs were reviewed to determine a generic region of the power / flow cap in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0.6 :ss chosen as the bases for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This generic region has been determined to correspond to a core finw of less than or equal to 45% of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.

LIMERICK - UNIT 2 B 3/4 4-1

CONTAINMENT SYSTEMS BASEC 3/4.6.2 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 55 psig during primary system blowdown from full operating pressure.

The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.

The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from rated conditions. Since all of the gases in the drywell are purged into the l suppression chamber air space during a loss-of-t.colant accident, the pressure of the suppression chamber air space must rat exceed 55 psig. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water volumes given in this specification, l suppression pool pressure during the design basis accident is below the design pressure. Maximum water volume of 134,600 ft' results in a downcomer submergence of 12'3" and the minimum volume of 122,120 ft' results in a submergence approximately 2'3" less. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to the downcomer submer-gence, this specification is adequate. The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperature above 170*F.

Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.

Under full power operating conditions, blowdown through safety / relief valves assuming an initial suppression chamber water temperature of 95 F results in a bulk water temperature of approximately 136*F immediately following blowdown which is below the 190 F bulk temperature limit used for complete condensation via T-quencher devices. At this temperature and atmospheric pressure, the avail-able NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase. l If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations. ;

LIMERICK - UNIT 2 B 3/4 6-3

ADMINISTRATIVE CONTRES CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b. MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1.

c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,

d. The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,

e. LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

f. The power biased Rod Block Monitor setpoints and the Rod Block '

Monitor MCPR OPERABILITY limits of Specification 3.3.6.

6.9.1.10 The analytical methods used to determine the core operating limits l shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

l

a. NEDE-240ll-P-A " General Electric Standard Application for Reactor !

Fuel (Latest approved revision).

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core I thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplement. shall be provided upon issuance for each reload cycle to the I

NRC Document Ccatrol Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special repori.s shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

LIMERICK - UNIT 2 6-18a

1 l

i ATTACHMENT 3 LIMERICK GENERATING STATION Units 1 and 2 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 NEDC-32193P, " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2," Revision 1, Class III, dated July 1993 Affidavit Supporting General Electric Company's request to withhold NEDC-32193P from public disclosure.

_______a

GENERAL ELECTRIC COMPANY AFFIDAVIT I, DAVID J. ROBARE, being duly sworn, depose and state as follows:

(1) I am Project Manager, Plant Licensing, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph 2.which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the GE proprietary report NEDC-32193P, " Maximum' Extended Load Line Limit and ARTS Improvement Program for Limerick Generating Station Units 1 and 2", Revision 1, Class III, dated July 1993. This information is delineated by bars marked in the margin adjacent to the specific material.

(3) In making this application for withholding of proprietary information of which'it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552 (b) (4) , _ and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17 (a) (4 ) , 2. 790 (a) (4 ) , and 2.790(d) (1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from-disclosure is here sought is all " confidential commercial information",

and some portions also qualify _under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enercy Proiect v. Nuclear Reculatory Commission. 975F2d871 (DC Cir. 1992), and Public Citizen Health Research GrouD v. FDA, 704F2d1280 (DC Cir.

1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would l

reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, pres er.c , ,

or future General Electric customer-funded j

^

development plans and programs, of potential commercial value to General Electric;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.

(5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in con.4dence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures

-3 -

outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2) is classified as proprietary because it contains detailed results of analytical models, methods and processes, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of plant improvements to increase operational flexibility and-efficiency for the BWR.

The development and approval of the transient, accident, ,

heat transfer and structural computer codes used in this analysis was achieved at a significant cost, on the order of several million dollars, to GE.

The develcpment of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of prafit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

l

The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

STATE OF CALIFORNIA )

) SS:

COUNTY OF SANTA CLARA )

David J. Robare, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at San Jose, California, this I day of U U C[ ,193 hdbh David J.'Robare General Electric Company Subscribed and sworn before me this ~ day of tb , 19 h l O O f ,

OFFICIAL SEAL Club WUlk k Notary Public, State 6of California Q PAULA F. HUSSEY fJOvky PerLC - CAUFCC*"a

,Yp SASTA CLARA COUNTY nm- My comm. expires APR 5.1994 ;

_me-me . -

6/29/93RTH

, ATTACHMENT'4-LIMERICK GENERATING STATION Units 1 and 2 Docket Nos.-50-352-50-353 License Nos. NPF-39 NPF-85 NEDC-32170P, " Limerick Generating' Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis,"

Revision 1, Class III, dated June 1993 Affidavit supporting General Electric Company's request to withhold NEDC-32170P from public disclosure.

9 l

GENERAL ELECTRIC COMPANY AFFIDAVIT I, DAVID J. ROBARE, being duly sworn, depose and state as follows:

(1) I am Project Manager, Plant Licensing, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph 2 which is sought to be withheld, and have been authorized to apply l for its withholding. l (2) The information sought to be withheld is contained in the GE proprietary report NEDC-32170P, " Limerick Generating Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis", Revision 1, Class III, dated June 1993.

This information is delineated by bars marked in the margin adjacent to the specific material.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552 (b) (4 ) , and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17 (a) (4) , 2.790 (a) (4) , and 2.790(d) (1) for " trade secrets and commerci:1 or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information",

and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Enerav Proiect v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FCA, 704F2d1280 (DC Cir.

1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would

reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers; l

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential j

commercial value to General Electric;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent j protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs ,

i (4 ) a. and (4)b., above. ;

i (5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so l

held. Its initial designation as proprietary information, i and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. The information sought to be withheld has, to the best of my ;

knowledge and belief, consistently been held in confidence l by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been 1.ade, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis. 1 (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent ,

authority, by the manager of the cognizant marketing {

function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures

i

. i i

outside GE are limited to regulatory bodies, customers, and ;

potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2) is classified as proprietary because it contains detailed results of analytical models, methods and processes, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of the loss-of-coolant accident for the BWR.

The development and approval of the loss-of-coolant >

accident computer codes used in this analysis was achieved at a significant cost, on the order of several million dollars, to GE.

The development of the evaluation process along with the ;

interpretation and application of the analytical results is j derived from the extensive experience database that !

constitutes a major GE asset.

i (9) Public disclosure of the information sought to be withheld j is likely to cause substantial harm to GE's competitive ;

position and foreclose or reduce the availability of '

profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its '

commercial value extends beyond the original development ;

cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and '

apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing ,

analyses done with NRC-approved methods.

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation 1 process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

l l GE's competitive advantage will be lost if its competitors l are able to use the results of the GE experience to normalize or verify their own process or if they are able ,

to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the l

l l

e a s f .

' A (K , , , 1 I y). '

f.

information were disclosed to the public. Making such ;

information available to competitors without their having been required to undertake a similar expenditure of i resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its '

competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools, r

l ;

l STATE OF CALIFORNIA ) ;

) SS:

COUNTY OF SANTA CLARA )

David J. Robare, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated ;

therein are true and correct to the best of his knowledge, l information, and belief.

Executed at San Jose, California, this O day of SUNE ,1923 David J. Robare General Electric Company .

Subscribed and sworn before me this day of bjli,19M dt G b415_ f t1/

, , Notary Public, State ff California i OFFICIAL SEAL PAULA F. HUSSEY l

1( Te, TJCTARY PUBli - CAUFCRMI- !

] .x j MNTA Ct/S.A COUNTY J. 3, My comm. expres APR 5.1904 Qn,. mn.c wnn - -

~

l 6/17/93RTH l l

l i

ML20056G492

Text

Reference:

Reference:

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