ML20246F382

Amend 30 to License NPF-39,revising Tech Specs to Permit Operation of Reactor W/One of Two Reactor Recirculation Loops in Svc Under Certain Specified Conditions

ML20246F382
Person / Time
Site:	Limerick
Issue date:	06/30/1989
From:	Butler W Office of Nuclear Reactor Regulation
To:
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ML20246F384	List: ML20246F382 ML20246F398
References
NUDOCS 8907130236
Download: ML20246F382 (39)

Text

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NUCLEAR REGULATORY COMMISSION r,

j WASHINGTON, D. C. 20656

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l PHILADELPHIA ELECTRIC COMPANY l

DOCKET NO. 50-352

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LIMERICK GENERATING STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE

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Amendment No. 30 License No. NPF-39 1.

The Nuclear Regulatory Commission (the Comission) has found that A.

The application for amendment by Philadelphia Electric Company (the licensee) dated November 4, 1988 as supplemented March 29, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

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

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-39 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and-the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 30, are hereby incorporated into this license. Philadelphia Electric Company shall operate the facility I

in accordance with the Technical Specifications and the Environmental Protection Plan.

8907130236 890630 PDR ADOCK 0500

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This license amendment is effective within 30 days of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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Walter R. Butler, Director Project Directorate I-2 l

Division of Reactor Projects I/II l

l At.tachment:

Changes to the Technical Specifications Date of Issuance:

June 30, 1989

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1 ATTACHMENT Ti) LICENSE AMENDMENT NO. 30 FACILITY OPERATING LICFNSE NO. NPF-39 DOCKET NO. 50-352 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pagos are provided to maintain document completeness.*

Remove Insert 2 2-1 i

2-2 2-2*

2-3 2-3*

2-4 2-4 B 2-1 B 2-1 B 2-2 B 2-2*

3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2*

3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8*

3/4 2-9 3/4 2-9*

3/4 2-10 3/4 2-10 3/4 2-10a 3/4 2-10a 3/4 2-10b 3/4 2-10b 3/4 2-10c 3/4 2-10c 3/4 3-59 3/4 3-59*

3/4 3-60 3/4 3-60 3/4 3-60a 3/4 3-60a 3/4 4-1 3/4 4-1 3/4 4-2 3/4 4-la 3/4 4-2 3/4 4-3 3/4 4-3 l

3/4 4-4 3/4 4-4 3/4 4-4a r

- - - _ - _ - _ _ - - _ _ - _ _ _ _. = - _. - _. - - - - -..

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4 -

, ATTACHMENT TO LICENSE AMENDMENT NO. 30 FACILITY OPERATING LICENSE NO. NPF-39 l

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DOCKET NO. 50-352 Remove Insert 3/4 4-5 3/4 4-5 3/4 4-6 3/4 4-6*

8 3/4 1-1 8 3/4 1-1*

8 3/4 1-2 8 3/4 1-2 8 3/4 2 B 3/4 2-1 8 3/4 2-2 B 3/4 2-2 8 3/4 2-3 8 3/4 2-3*

B 3/4 2-4 8 3/4 2-4 8 2/4 4-1 B 3/4 4-1 B 3/4 4-2 8 3/4 4-2

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• V 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:

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 for two recirculation loop operation and shall not be less than 1.08 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

l-APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

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ACTION:

With MCPR less than 1.07 for two recirculation loop operation or less than 1.08 for single recirculation loop operation 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.

l REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

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 J

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.

i LIMERICK - UNIT 1 2-1 Amendment No.M 30 J

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4

2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow l

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:

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 for two recirculation loop operation and shall not be less than 1.08 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.07 for two recirculation loop operation or less than 1.08 for single recirculation loop operation 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 vessel steam dome, shall not exceed 1325 psig.

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.

LIMERICK - UNIT 1 2-1 Amendment No.M 30 L__.___-.__m_

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued)

REACTORVESSELWAfERLEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

1 APPLICABILITY: OPERATIONAL CONDITIONS 3, 4, and 5 ACTION:

With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required. Comply with the requirements of Specification 6.7.1.

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LIMERICK - UNI' 1 2-2

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

APPLICABILITY As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

LIMERICK - UNIT 1 2-3

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2.0 INTRODUCTION

l The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive 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 for two recirculation loop operation and 1.08 for single recirculation loop operation. MCPR greater than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain (bal cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

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 incre-mentally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation sipificantly above design conditions and the Limiting Safety System Settings.

Whlie 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 signi-ficant departure from the condition intended by design for planned operation.

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 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 alwa s 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 criti-cal 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 limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

LIMERICK - UNIT 1 B 2-1 Amendment No.,T,' 30 3

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SAFETY LIMITS l

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 calculcted to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, i

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 and in the procedures used to calculate the critical power result in an uncertainty l

in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are* expected to avoid boiling transition consider'ing the power distribution within the core and all uncertainties.

TheSafetyLimitMCPRisdeterminedusingastatisticalmodelthatcombiriesl l

all of the uncertainties in operating parameters and the procedures tised to calculate critical power. Calculation of the Safety Limit MCPR is described in Reference 1.

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" General Electric Standard Applicatic-for Reactor Fuel," NEDE-24011-P-A

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(latest approved revision).

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LDERICK - LMIT 1 8 2-2 Amensment No. 7 AUG 14 '--

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3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR 1.INEAR 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 for two recirculation loop operation.

1 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.

The limits shall be reduced to a value of C.89 times the two recirculation loop operation limit when in single recirculation loop operation.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than

• or equal to 25% of RATED THE,RMAL 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 hourt.

SURVEILLANCE RE0VIREMENTS 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 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 R00 PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

LIMERICK - UNIT 1 3/4 2-1 Amendment No.,X 30

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POWER DISTRIBUTED' JMITS 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 (SRB) shall be established according to the following relationships:

TRIP SETPOINT ALLOWABLE VALUE During two recirculation 5 < (0.58W + 59%)T 5 < (0.58W + 62%)T loop operation Sf5(0.58W+50%)T Sj5(0.58W+53%)T R

R During single recirculation S < (0.58W + 54%)T S < (0.58W + 57%)T loop operation Sf5(0.58W+45%)T Sj5(0.58W+48%)T p

R where:

S and S are in percent of RATED THERMAL POWER, RB 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 l

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 S

as above determined, initiate corrective action within 15 minutes abb,adjustSand/or5 to be consistent with.the Trip Setpoint values

• within6hoursorred0$eTHERMALPOWERtolessthan25%ofRATEDTHERMAL 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 and the MFLPD shall be determined, the value of T calculated, and the mcst 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 the 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 MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THFRMAL POWER and a notice of adjustment is posted on the reactor control panel.

LIMERICK - UNIT 1 3/4 2-7 Amendment No.f, 30

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t POWER DISTRIBUTION LIMITS j

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), Figure 3.2.3-lb (BP/P8X8R fuel), Figura 3.2.3-Ic (GE 8X8E8 fuel) and Figure 3.2.3-1d j

(GE 8X8EB fuel) times the K, shown in Figure 3.2.3-2, provided that the l

end of-cycle recirculation bump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2, with:

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T

~I A

B where:

A = 0.86 seconds, control rod average scram insertion I

T time limit to notch 39 per Specification 3.1.3.3, l

B = 0.672 + 1.65[

.]b(0.016),

1 T

n N

I i

i=1 a

I t,y,,

$_y N t, g

n

\\

I NI i=1 I

i n = number of surveillance tests performed to date in cycle, Ng = number of active control rods measured in the ith surveillance test, tg = average scram time to notch 39 of all rods measured th in the i surveillance test, and Ny = 4.1.3.2.a. total number of active rods measured in Specification 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 AmenpPE:

nt No.

19 4 1989

l POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) i ACTION With the end-of-cycle recirculation pump trip system inoperable per a.

Specification 3.3.4.2, operation may continue provided that, within I hour, MCPR is determined to be greater than or equal to the MCPR t

l limit as a function of the average scram time shown in Figure l

3.2.3-la (BP/P8X8R fuel), Figure 3.2.3-lb (8P/P8X8R fuel),

Figure 3.2.3-Ic (GE8X8EB fuel) and Figure 3.2.3-1d (GE8X8EB fuel),

EOC-RPT inoperable curve, times the k7 shown.in Figure 3.2.3-2.

t b.

With MCPR less than the appitcable MCPR limit shown in Figures 1.2.3-la, 3.2.3-1b and 3.2.3-2, initiate corrective action within 15 minutes and 3

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 THERNAL POWER to less than 25% of RATED THERMAL POWER within th 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:

t = 1.0 prior to performance of the initial scram time measurements a.

for the cycle in accordance with Specification 4.1.3.2, or b.

I 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.

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 Amendment No.

19 I

APR 2 4 1989

4 1.44 c

1,44 1.42 -

1.42 1.4 0. -

1,40 1.38 -

1,33 OUT.E')C.RP_T., WM

1.4-

[ 1.3 2 -

t o2 1.32 1.30 -

i 1.30 1.28 -

EOC-RPT, ICF#

~ 1.28 wt 1.26 1.26 e

1.24-1.24

• lNCLUDES FHOOS j

1.22 -

1.22 1.20 i

i i 1.20 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

l T

i 1

Note:

These limits apply. to Both Two Recirculation Loop and Single Recirculation Loop Operation.

DEFINITIONS:

ICF - INCREASED CORE FLOW (UP TO 105% RATED)

FH005 - FEEDWATER HEATING OUT OF SERVICE THROUGHOUT CYCLE (UP TO 60'F TEMP. REDUCTION; ACHIEVED BY REMOVAL OF FEEDWATER HEATER (s))

FFWTR - FINAL FEE 0 WATER TEMPERATURE REDUCTION AT END OF CYCLE (UP TO 60 F TEMP. REDUCTION: ACH7.EVED BY REMOVAL OF ALL 6TH STAGE HEATERS)

MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS t (P8x8R/BP8x8R FUEL) i (B0C TO E0C - 2000 MWD /ST)

FIGURE 3.2.3-la LIMERICK - UNIT 1 3/4 2-10 Amendment No.Z9, 30

s

~

j 1.44

,44 1.42 -

1.42 ICF+ff g o C E P T g T g iC F--

1.40-g,4 g 1.38 - -WN 1.38 1.36-1.36 WITH EOC-RPT ' ICF+M '

+

1.34 -

1,34 e

$ 1.32 -

1.32 FLOW 1.30 -

p 3,3 o 1.28 1.28 1.26-1.26 1.24 -

~ 1.2 4

~

• lNCLUDES FHOOS 1.22-1.22 1.20 i i 1.20 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

T Note:

These limits apply to Both Two Recirculation Loop and Single Recirculation loop Operation.

DEFINITIONS:

ICF - INCREASED CORE FLOW (UP TO 105% RATED)

FH005 - FEEDWATER HEATING OUT OF SERVICE THROUGHOUT CYCLE (UP TO 60 F TEMP. REDUCTION; ACHIEVED BY REMOVAL OF FEEDWATER HEATER (S))

FFWTR - FINAL FEEDWATER TEMPERATURE REDUCTION AT END-0F-CYCLE (UP TO 60 F TEMP. REDUCTION: ACHIEVED BY REMOVAL OF ALL 6TH STAGE HEATERS)

MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS I (P8x8R/BP8x8R FUEL)

(E0C - 2000 MWD /ST TO E0C)

FIGURE 3.2.3-lb LIMERICK - UNIT 1 3/4 2-10a Amendment No. D, 30 l

1 44 1.44 1.42-1.42 1.40' 1.40 1.3 8 ~

1,33 1.3 8 ~

1.33 M cp. CF+

1.34 '

wiWOUT W 1.34 g

1.3 2 ~

1.32 1.30-1,3 o E'_gv W

• 1.28 -

pgpT--

1.28 l

1.25 1.26 1.24" 1.24

• lNCL.UDES PHOOS

{

1.22 -

1.22 1.20,

i i

i i

i i

i i 1.20 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.h 1

7 Note:

These Limits Apply To Both Two Recirculation Loop and Single Recirculation Loop Operation.

DEFINITIONS:

ICF - INCREASED CORE FLOW (UP TO 105% RATED)

FH005 - FEEDWATER HEATING OUT OF SERVICE THROUGHOUT CYCLE (UP TO 60"E TEMP.

REDUCTION; ACHIEVED BY REMOVAL OF FEEDWATER HEATER (S))

FFWTR - FINAL FEEDWATER TEMPERATURE REDUCTION AT END-OF-CYCLE (UP TO 60 F TEMP. REDUCTION; ACHIEVED BY REMOVAL OF ALL 6th STAGE. HEATERS)

MINIMUM CRITICAL POWER RATIO (MPCR) VERSUS t (GE8x8EB FUEL)

(BOC TO E0C - 2000 MWD /ST)

FIGURE 3.2.3-1c LIMERICK - UNIT 1 3/4 2-10b Amendment No.19, 30 l

- o.*'

1.44 l g,44 1.42 -

1 42 1.40

___ _.cp c gun 3,4 o

-po@ EOC 1.3 s -

_1.3 s t

I 1 38 i

1.36 4

1 34 -

g p pT.KF M

" 1.34 b

1.3 2 ~

1.32

-w!TH EOCAM #

py,100%-Ftoo..,

- 3,3 o 1.30 1.28 1.28 1.26-1.26 i

i t

1.24 -

1.24

• lNCLUDES FH005 1.22-1.22 1.20 4

e i

i i

i 1.20 0

C.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

7 i

i Note:

These Limits Apply To Both Two Recirculation Loop and Single Recirculation i

Loop Operation.

DEFINITIONS:

ICF - INCREASED CORE FLOW (UP TO 105% RATED)

{

FH005 - FEEDWATER HEATING OUT OF SERVICE THROUGHOUT CYCLE (UP TO 60 F TEMP. REDUCTION; ACHIEVED BY REMOVAL OF FEEDWATER HEATER (S))

FPWTR - FINAL FF9 WATER TEMPERATURE REDUCTION AT END-0F-CYCLE (UP TO 60 F TEMP. C Efl0N; %.HIEVED BY REMOVAL OF ALL 6th STAGE HEATERS)

MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS T (GE8x8EB FUEL)

(EOC - 2000 MWD /ST TO E0C)

FIGURE 3.2.3-1d LIMERICK - UNIT 1 3/4 2-10c Amendment No.19, 30 l

-m__._________

2 TABLE 3.3.6-1 (Continued)

CONTROL 800 WITHDRAWAL BLOCK INSTALMEWTATION ACTION STATEMENTS l

ACT10N 60 Declare the RSM inoperable and take the ACTION required by Specification 3.1.4.3.

ACTION 61 -

With the neber of.0PERABLE channels one er more less than required by the Minisium OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.

ACTICN 62 With the number of OPERABLE channels less than required by the Minious OPERA 8LE Channels per Trip Function requimment, place the inoperable channel in the tripped condition within one hour.

ACTION 63 With the number of OPERA 8LE channels'less than requimd by the Minimum OPERA 8LE Channels per Trip Function requirement, initiate a rod block.

E With THERMAL POWER 133 of RATEp THERMAL POWER.

With more than one control red withdrawn. Not applicable to control rods removed per specification 3.9.10.1 or 3.9.10.2.

Thesa channels are not reeufred when sixteen or fewer fuel assemblies,

.sejacent to the $RMs, are in the core.

(a) The R8M shall be autenatically bypassed when a peripheral control rod is selected or the reference APD channel indicates less than 3S of RATED TMERMAL POWER.

(b) This function shall be autenstically byp: dred if detector count rate is

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

(c) This function is automatically bypassed when the associated IM channels are en rence 4 of higher.

(d) This function is autenatically bypassed when tne IM channels are en range 3 er higher.

(e) This funetten is autenatically bypassed when the IM channels are on rente 1.

LIMERICK - UNIT 1 3/4 3*59 Asendment No. 4 mat i i *3A7

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