ML20206A559

Proposed Tech Specs,Identifying Operating Limits for All Fuel Types for Cycle 2 Operation & Incorporating Change in Slope of Flow Biased APRM Scram & Rod Block Setpoints

ML20206A559
Person / Time
Site:	Limerick
Issue date:	04/03/1987
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML19292H074	List: ML20206A541 ML20206A550 ML20206A559 ML20206A569 ML20206A576
References
NUDOCS 8704080073
Download: ML20206A559 (23)

Text

i I

l INDEX )

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ,

l SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow ................ 2-1 THERMAL POWER, High Pressure and High Flow ............. 2-1 Reactor Coolant System Pressure ........................ 2-1 Reactor Vessel Water Level ............................. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints .... 2-3 Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints .......... 2-4 SASES 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow ................ B 2-1 THERMAL POWER, High Pressure and High Flow ............. B 2-2 Left Intentionally Blank ............................... B 2-3 Left Intentionally Blank ............................... B 2-4 Reactor Coolant System Pressure ......................... B 2-5 Reactor Vessel Water Level .............................. B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints ..... B 2-6 LIMERICK - UNIT 1 iv B70'40B0073 B7C403 i PDR ADOCK 05000352 i P , PDR;  ;

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

Figure 3.2.1-2 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial  :

Core Fuel Types P8CIB248 .............. 3/4 2-3 Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB163............... 3/4 2-4 Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB094............... 3/4 2-5 Figure 3.2.1-5 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB071............... 3/4 2-6 Figure 3.2.1-6 Maximum Averaqe Planar Linear Heat Generation.a 2 (MAPLHGR) Versus Average Planar Exposure For Fuel Type BC320A (GE8X8EB)........ 3/4 2-6a 3/4 2.2 A P RM S ET PO I N T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 2-7 3/4 2.3 MINIMUM CRITICAL POWER RATIO............................. 3/4 2-8 Table 3.2.3-1 Deleted Figure 3.2.3-la Minimum Critical Power Ratio (MCPR)

Versus (P8X8R/BP8X8R Fuel),,,,,,,,,,,3/4 2-10 Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR)

Versus (GE8X8EB Fuel)...............,3/4 2-10a Figure 3.2.3-2 Kf Fa c t o r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 2 - 11 3/4.2.4 LINEAR HEAT GENERATION RATE .............................3/4 2-12  !

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ............. 3/4 3-1 Table 3.3.1-1 Reactor Protection System Ins t rumenta tion . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times......................... 3/4 3-6 Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance l Requirements ........................... 3/4 3-7 i vi

-, .. ~ - - , ..

e INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY .................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN .................................. B 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES ............................. B 3/4 1-1 3/4.1.3 CONTROL RODS ..................................... B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTROLS ..................... 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 RATE ....... B 3/4 2-1 LEFT INTENTIONALLY BLANK .................................. B 3/4 2-3 3/4.2.2 APRM SETPOINTS ................................... B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO ..................... 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 i 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION .. B 3/4 3-3 1

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION l INSTRUMENTATION .................................. B 3/4 3-4 l

l 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION ................ B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation ............. B 3/4 3-4 l

LIMERICK - UNIT 1 xviii

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

i With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated f3ow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.07 and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE ,

2.1.3 The reactor coolant system pressure, as measured in the reactor l vessel steam dome, shall not exceed 1325 psig. l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

l l

LIMERICK - UNIT 1 2-1

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

TABLE 2.2.1-1 REACTOR PPOTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT TABLES 1 Intermediate Range Monitor, Neutron Flus-High 1 120/125 divisions 1 122/125 divisions of full scale of full scale l'

2. Average Power Range Monitors

a. Neutron Flum-Upscale. Setdown i 15% of RATED THERMAL POWER 1 20% of RATED j THERMAL POWER I

b. Neutron Flux-Upscale

1) Flow Biased 1 0.58W +59%. =1th 1 0.58W + 62%. with a maximum of a maximum of

2) High Flow Clamped 1 116.5% of RATED 1 118.5% of RATED THERMAL POWER THERMAL POWEP ,

c. Inoperative N.A. N.A.

c. Downscale 1 4% of RATED 1 3% of RATED THERMAL POWER THERMAL POWER

3. Reactor Vessel Stam Dome Pressure - High 1 1037 psig 1 1057 Asig 4 Reactor Vessel Water Level - Low. Level 3 1 12.5 inches above instrument 1 11.0 inches above aero* instrument zero 4

5. Main Steam Line Isolation Valve - Closure 1 8% Closed i 12% Closed

6. Main Steam Line Radiation - High 1 3.0 X full power background 1 3.6 X full power background

7. Drywell Pressure - High 1 1.68 psig i 1.88 psig

8. Scram Discharge Volume water Level - High

a. Level Transmitter 1 260' 9 5/8" elevation ** 1 261' 5 5/8" elevation

b. Float Switch 1 260' 9 5/8" elevation ** 1 261* 5 5/8 elevation

9. Turbine Stop Valve - Closure 1 5% Closed 1 7% Closed

10. Turbine Control Valve Fast Closure.

Trip 011 Pressure - Low 1 500 psig 1 465 psig

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

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

• See Bases Figure B 3/4.3-1.

• • Equivalent to 25.45 ga11ons/ scram discharge volume.

LIMERICK - UNIT 1 2-4

)

1

)

2.1 SAFETY LIMITS A

BASES j

2.0 INTRODUCTION

l l

l The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive i materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and  ;

anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07. MCPR greater than 1.07 represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

i The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur frca reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

1 2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the (GEXL) correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety )

Limit is established by other means. This is done by establishing a l

limiting condition on core THERMAL POWER with the following basis.

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will I always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 X 10 lb/h, bundle pressure drop is nearly independent of bundle

power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 X 103 lb/h. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL

, POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER i

limit of 25% of' RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

l LIMERICK - UNIT 1 B 2-1

2.1 SAFETY LIMITS BASES 2.1.2 THERMAL POWER, High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity Safety Limit is def1ned as the CPR in the limiting fuel assembly for which more than 99.91 of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. Calculation of the l

Safety Limit MCPR is described in Reference 1.

Reference:

1. " General Electric Standard Application for Reactor Fuel", NEDE-240ll-P-A (latest approved revision).

LIMERICK - UNIT 1 B 2-2

1 LEFT INTENTIONALLY BLANK i

l l

LIMERICK - UNIT 1 B 2-3

m O O LEFT INTENTIONALLY BLANK LIMERICK - UNIT 1 B 2-4

I 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 approved for the respective fuel and lattice types. When hand calculations are required, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) as shown in the applicable figures for BP/P8X8R and GE8X8EB fuel types.

APPLICABILITY: OPERATIONAL CONDITION 1, 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 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of

, RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

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

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

d. The provisions of Specification 4.0.4 are not applicable. .

l LIMERICK - UNIT 1 3/4 2-1

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6.5 i i i i 0 10000 20000 30000 40000 50000 Average Planar Exposure (mwd /ST)

MAXIMUM AVEPAGE PLANAR UNEAPs HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FOR FUEL TYPE EC320A (GE8X8EB)

RGURE 3.2.1-6

W' 3/4.2 POWER DISTRIBUTION LIMITS 3 /4 . 2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased neutron flux-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (g3 ) shall **

be established according to the following relationships:

TRIP SETPOINT ALLOMABLE VALUE S 5 (0.58W + 59%)T S 5 (0.58W < 62%)T SRB $ (0.58W + 50%)T S RB where: S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr.

T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY. T is applied only if less than or equal to 1.0.

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

ACTION:

With the APRM flow biased neutron flux-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or SRB , as above determined, initiate corrective action within 15 minutes and adjust S and/or SRB to be consistent with the Trip Setpoint values

• within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP 2nd'the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased neutron flux-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

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

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when tha reactor is operating with MFLPD greater than or equal to FRTP.

d. The provisions of Specification 4.0.4 are not applicable.

• With MFLPD greater than the FRTP during power ascension up to 90% of Rated Thermal Power rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times MPLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

LIMERICK - UNIT 1 3/4 2-7 r

POWER DISTRIT!UTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit shown in Figure 3.2.3-la (BP/P8X8R fuel) and Figure 3.2.3-lb (GE8X8EB fuel), times the Kg shown in Figure 3.2.3-2, provided that the end-of-cycle recirculation pump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2, with:

1[ = (7'a ve - 7B)

TA - TB where:

! TA = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, 4

GI B = 0.672 + 1.65 ( } (0.016) n

%N i=1 g

n E

Tave = i=1 N171 n

I

=l n = number of surveillance tests performed to date in cycle.

Ni = number of active control rods measured in the ith surveillance test, i = average scram time to notch 39 of all rods measured in the 1(th) surveillance test, and N1 = 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.

LIMERICK - UNIT 1 3/4 2-8

POWER DISTRIBUTION LIMITS LI'MITING CONDITION FOR OPERATION ACTION:

a. With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, MCPR is determined to be greater than or equal to the MCPR limit as a function of the average scram time shown in Figure 3.2.3-la (BP/P8X8R fuel) and Figure 3.2.3-lb (GE8X8EB fuel), EOC-RPT inoperable curve, times the K g shown in Figure 3.2.3-2.

b. With MCPR less than the applicable MCPR limit shown in Figures 3.2.3-la, 3.2.3-lb and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

a. 1r = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or

b. "b as defined in Specification 3.2.3 used to determine the limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit determined from Figures 3.2.3-la, 3.2.3-lb and 3.2.3-2:

i

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

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

d. The provisions of Specification 4.0.4 are not applicable.

LIMERICK - UNIT 1 3/4 2-9

1.44 . . . . . . . . . 1.44 1.42-  ;-  ;-  ; ---;-

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DEFINITIO NS-LGE - INCREASED CORE FLOW (UP TO 1057. RATED)

FHOOS - FEEDWATER HEATING OUT OF SERVICE THROUGHOUT CYCLE (UP TO 60 DEG.F TEMP. REDUCTION: ACHIEVED BY REMOVAL OF FEEDWATER HEATER (S))

FFWTR - FINAL FEEDWATER TEMPERATURE REDUCTION AT END-OF-CYCLE UP TO 60 DEG.F TEMP REDUCTION: ACHIEVED BY REMOVAL OF ALL 6TH STAGE HEA ERS)

MNMUM CRmCAL POWER RATIO (MCPR)

VERSUS T (P8X8R/BP8X8R FUEL)

FlGURE 3.2.3-1a Limerick - Unit 1 3/4 2 -10 _

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WITH EOC-gpT' 1007.

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i* INCLUDES FHOOS i. i i

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:  :  :  :  : l 1.20- i i i [

[ i i i i 1.20 0

0.10.20.30.40.50.60.70.80.9 1 T

D EFINITIO N S-LGE - INCREASED CORE FLOW (UP TO 105% RATED)

FHOOS - FEEDWATER HEATING OUT OF SERVICE THROUGHOUT CYCLE (UP TO TEMP. REDUCTION: ACHIEVED BY REMOVAL OF FEEDWATER HEATER FFWTR - FINAL FEEDWATER TEMPERATURE REDUCTION AT END-OF-CYCLE TEMP REDUCTION: ACHIEVED BY REMOVAL OF ALL 6TH STAGE HEA ER MINIMUM CRmCAL POWER RATIO CPR) l VERSUS T (GE8X8EB FU FIGURE 3.2.3-1b Limerick - Unit 1 3/4 2-10a

POWER DISTRIBUTION LIMITS

'$/4.2.4 LINEAR HEAT GENERNTION RATE LIMITING CONDITION FOR OPERATION ,

3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.4 l kW/ft for BP/P8X8R fuel and 14.4 kW/ft for GE8X8EB fuel. I

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

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS

(

4.2.4 LHGRs shall be determined to be equal to or less than the limit:

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

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

I

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is -

f operating on a LIMITING CONTROL ROD PATTERN for LHGR.

d. The provisions of Specification 4.0.4 are not applicable.

LIMERICK - UNIT 1 3/4 2-12 1 1

i i

TABLE 3.3.6-2

• CONTROL ROD BLOCK INSTRUMENTATION SETPOTNTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

1. ROD BLOCK MONITOR

a. Upscale

1. flom biased 1 0.66 w + 41%. with a manimum of. 1 0.66 W + 44%. with a maximum of,

11. hign flow clamped 1 107% 1 110%

b. Inoperative N.A. N.A.

C. Down5Cale 1 5% of RATED THERMAL POWER 1 3% of RATED THERMAL POWER 2, APRM

a. Flow Biased Neutron Flum - Upscale 1 0.58 w + 50%* 1 0.58 W + 53%*

l-

b. Inoperatike N.A. N.A.

c. Downscale 1 4% of RATED THERMAL POWER 1 3% of RATED THERMAL POWER

c. Neutron Flum - upscale. Startup < 12% of RATED THERMAL POWER 1 14% of RATED THERMAL POWER 3 SOURCE RANGE MONITORS

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

b. Upscale 11 X10 5 cps 1 1.6 xiO 5c ,,

c. Inoperative N.A. N.A.

-d. Downscale 1 3 cos** 1 1.8 cos**

4 INTERMEDIATE RANGE MONITORS

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

b. Upscale 1 108/125 divisions of 1 110/125 divisions of full scale full scale

c. Inoperative N.A. N,A.

c. Downscale 1 5/125 divisions of full scale 1 3/125 divisions of full scale

5. SCRAM DISCHARGE VOLUME

a. water Level-Mign 1 257' 5 9/16" elevation *** 1 257' 7 9/16" elevation

a. Float Switcn

6. REACTOR COOLANT SYSTEM RECTRCULATION FLOW

a. Upscale 1 111% of rated flow i 114% of rated flow

b. Inoperative N.A. N.A.

c. Comparator 1 10% flow ceviation i 11% flow ceviation.

7. REACTOR MODE Sw1TCH SHUTDOWN POSITION N.A. N.A, eThe Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of tnis function must be maintaineo in accordance with Specification 3.2.2.

• • may be reducea to 0.T cos proviced the signal-to-noise ratio is 1 2.

eseEquivalent to 13 gallons / scram cisenarge volume.

LIMERICK - UNIT 1 3/4 3-60

- - ~ p-

l e

I 1

3/4.2 POWER DISTRIBUTION LIMITS

, 1 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 10CFR50.46 and that the fuel design analysis limits specified in NEDE-240ll-P-A (Reference 2) will not be exceeded.

Mechanical Design Analysis: NRC 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 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 permissible planar power (MAPLHGR) limits comply with the ECCS limits specified in 10CFR50.46.

The analysis is performed for the most limiting break size, break location, and single failure combination for the plant.

The Technical Specification MAPLHGR limit is the most limiting composite of the fuel mechanical design analysis MAPLHGR and the ECCS

~

MAPLHGR limit.

Only the most and least limiting MAPLHGR values are shown in the Technical Specifications 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 1

J 1

j i

e LIMERICK - UNIT 1 B 3/4 2-1

. ,o POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER. The flow biased neutron flux-upscale scram trip setpoint.and flow biased neutron flux-upscale control rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than the Safety Limit MCPR specified in Reference 2 or that > 1% plastic strain does not occur in the degraded situation. The scram and rod block setpoints are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

LIMERICK - UNIT 1 B 3/4 2-2

}

I i

LEFT INTENTIONALLY BLANK LIMERICK - UNIT 1 B 3/4 2-3

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 specified in Reference 2, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial conditions 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 transients 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 limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.3-1.

The evaluation of a given transient begins with the system initial parameters 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 purpose of the Kg factor of Figure 3.2.3-2 is to define operating limits at other than rated core flow conditions. At less than 100% of rated flow the required MCPR is the product of the MCPR and the K g factor. The K f factors assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor-generator speed control failure. The K g factors may be applied to both manual and automatic flow control modes.

The Kg factors values shown in Figure 3.2.3-2 were developed generically and are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The K g factors were derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow.

For the manual flow control mode, the Kr factors were calculated such that for the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kr.

LIMERICK - UNIT 1 B 3/4 2-4

o s ,;PpWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated thermal flow.

The K f factors shown in Figure 3.2.3-2 are conservative for the General Electric Boiling Water Reactor plant operation because the operating limit MCPRs of Specification 3.2.3 are greater than the original 1.20 operating limit MCPR used for the generic derivation of Kf.

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, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial startup 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 when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached 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.

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.

References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

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", NEDO-31401, February 1987.

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.

LIMERICK - UNIT 1 B 3/4 2-5