Text

Amend 20 to License NPF-58,increasing MCPR from 1.06 to 1.07,adding Limiting Lattice most-limiting MAPLHGR Curves to Tech Specs to Account for New Fuel Types Used During Current Cycle & Deleting MAPLHGR Curve for Natural U
	ML20245J135
Person / Time
Site:	Perry
Issue date:	04/26/1989
From:	Hannon J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20245J137	List: ML20245J135; ML20245J139;
References
	NUDOCS 8905040009
	Download: ML20245J135 (39)
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UNITED STATES y

g NUCLEAR REGULATORY COMMISSION 7.

!j WASHINGTON, D. C. 20555

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i THE CLEVELAND ELECTRIC ILLUMINATING COMPANY, ET AL.

DOCKET NO. 50-440 PERRY NUCLEAR POWER PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. ?O License No. NFF-58 1.

The Nuclear Regulatory Commission (the Comission) has found that:

y A.

The application for amendment by The Cleveland Electric Illuminating Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania Power Company, and Toledo Edison Company (the licensees) dated November 20, 1988, as amended December 29, 1988 complies with the standards cnd requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the

provisions of the Act, and the rules and regulations of the Commission; C.

Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amendment ~can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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-58 is hereby amended to read as follows:

8905040009 890426 ADOCK0500g4O DR C

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C UNITED STATES g

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NUCLEAR REGULATORY COMMISSION 5

ij WASHING TON, D. C. 20555 s*...*/

THE CLEVELAND ELECTRIC ILLUMINATING COMPANY, ET AL.

@CKEYN0.50-440 PERRY NUCLEAR POWER PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. ?O License No. NE -58 1.

The Nuclear Regulatory Commission (the Convaission) has' found that:

3

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A.

The application for amendment by The Cleveland Electric Illuminating Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania Power Company, and Toledo Edison Company (the licensees) dated l

November 20, 1988, as amended December 29, 1988 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the I

provisions of the Act, and the rules and regulations of the Commission; C.

Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amendment can be conducted without endangering the herith and safety of the public, and (ii) that such activities will be ;onducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the comran 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 licer.se amendment, and paragraph 2.C.(2) i of Facility Operating License No. NPF-58 is hereby amended to read as follows:

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.Mr. Alvin Keplan Perry Nuclear Power Plant

.The Cleveland Electric-Unit 1

, Illuminating Company 1

cc: Jay E. Silberg, Esq.-

Mr. James W. Harris, Director Shaw, Pittman, Potts & Trowbridge Division of Power Generation 2300 N Street, N.W.

Ohio Department of Industrial l

Washington, D.C.

20037

-Relations P. 0. Box 825 David E. Burke Columbus, Ohio 43216 The. Cleveland Electric Illuminating Company The Honorable Lawrence Logan P.O. Box 5000 Mayor, Village of Perry Cleveland, Ohio 44101 4203 Har>er Street l

Perry. 0110 44081 Resident Inspector's Off:ce U.S. Nuclear Regulatory Ionnission The Honorable Robert V. Orosz-Paraly at Center Road >

Mayor, Village of North Perry Perry, Ohio 44081 North Perry Village Hall 4778 Lockwood Road Regional Administrator, Region III North Perry Village, Ohio 44081 U.S. Nuclear Regulatory Commission 799 Roosevelt Road Attorney. General Glen Ellyn, Illinois 60137 Department of Attorney General 30 East Broad Street Frank P. Weiss, Esq.

Columbus, Ohio 43216 Assistant Prosecuting Attorney 105 Main Street Radiological Health Program Lake County Administration Center Ohio Department of Health Painesville, Ohio 44077 1224 Kinnear Road Columbus, Ohio 43212 Ms. Sue Hiatt OCRE Interim Representative Ohio Environmental Protection 8275 Munson Agency Mentor, Ohio 44060 361 East Broad Street Colunbus, Ohio 43266-0558 Terry J. Lodge, Esq.

618 N. Michigan Street Mr. Phillip S. Haskell, Chairman Suite 105 Perry Township Board of Trustees Toledo, Ohio 43624 Box 65 4171 Main Street John G. Cardinal, Esq.

Perry, Ohio 44081 Prosecuting Attorney Ashtabula County Courthouse State of Ohio Jefferson, Ohio 44047 Public Utilities Commission 180 East Broad Street Robert A. Newkirk Columbus, Ohio 43266-0573 Cleveland Electric Illuminating Company Michael D. Lyster Perry Nuclear Power Plant Cleveland Electric P. O. lox 97 E-210 Illuminating Company Perry, Ohio 44081 Perry Nuclear Power Plant P. O. Box 97 58306 Perry, Ohio 44081

s.

q Mr. Alvin Kaplan Perry Nuclear Power Plant The Cleveland Electric Unit 1 Illuminating Company.

cc: Jay E. Silberg, Esq.

Mr. Jarnes W. Harris, Director Shaw, Pittman, Potts & Trowbridge Division of Power Generation 2300 N Street, N.W.

Ohio Department of Industrial Washington, D.C.

20037 Relations P. O. Box 825 David E. Burke Columbus, Ohio 43216 The. Cleveland Electric Illuminating. Company The Honorable Lawrence Logan P.O.. Box 5000 Mayor,. Village of Perry Cleveland, Ohio 44101 4203 Harper Street Perry, Ohio 44081 Resident Inspector's Ofice U.S. Nuclear Regulatory Comission The Honorable Robert V. Orosz Parnly at Center Road Mayor, Village of North Perry Perry 0hio 44081 North Perry Village Hall 4778 Lockwood Road Regional Administrator, Region 111 North Perry Village, Ohio 44081 U.S. Nuclear Regulatory Commission 799 Roosevelt Road Attorney General Glen Ellyn, Illinois 00137 Department of Attorney General 30 East Broad Street i

Frank P. Weiss, Esq.

Columbus, Ohio 43216 l

Assistant. Prosecuting Attorney 105 Main Street Radiological Health Program Lake County Administration Center Ohio Department of Health Painesville, Ohio 44077 1224 Kinnear Road Columbus, Ohio 43212 Ms. Sue Hiatt OCRE Interim Repasentative Ohio Environmental Protection 8275 Munson Agency Mentor, Ohio 44060 361 East Broad Street Columbus, Ohio 43266-0558 Terry J. Lodge, Esq.

618 N. Michigan Street Mr. Phillip S. Haskell, Chairman Suite 105 Perry Township Board of Trustees Toledo, Ohio 43624 Box 65 4171 Main Street John G. Cardinal, Esq.

Perry, Ohio 44081 Prosecuting Attorney Ashtabula County Courthouse State of Ohio Jefferson, Ohio 44047 Public Utilities Comission 180 East Broad Street Robert A. Newkirk Columbus, Ohio 43266-0573 Cleveland Electric Illuminating Company Michael D. Lyster Perry Nuclear Power Plant Cleveland Electric P. O. Box 97 E-210 Illuminating Company Perry, Ohio 44081 Perry Nuclear Power Plant P. O. Box 97 58306 Perry, Ohio 44M,

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' l (2) Technical' Specifications The Technical Specifications contained in Appendix A and the Environ-mental Protection Plan contained in Appendix B, as revised through Amendment No. 20 are hereby incorporated into this license. The Cleveland Electric Illuminating Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLtAR REGULATORY COMMISSION

~..

John N. Hannon, Director Project Directorate III-3 Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 26, 1989 1

7-ATTACHMENT TO LICENSE AMENDMENT N0.20 FACILITY OPERATING LICENSE N0.'NPF-58 I

DOCKET NO. 50-440 I

i Replace the following sages of the Appendix "A" Technical Specifications with the attached pages. T1e revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.

Remove Insert iv iv v

v vi vi xviii <

xviii j

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-e-i DEFINITIONS SECTION DEFINITIONS (Continued)

PAGE 1.47 TURBINE BYPASS SYSTEM RESPONSE TIME..........................

1-9 1.48 UNIDENTIFIED LEAKAGE.........................................

1-9 1.49 UNRESTRICTED AREA............................................

1-9 1.50 VENTILATION EXHAUST TREATMENT SYSTEMS........................

1-9 1.51 VENTING.......................................................

1-9 Table 1.1, Surveillance Frequency Notation................

1-10 1-Table 1.2, Operational Conditions.................................

1-11 5

PERRY - UNIT 1 iii

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION

~PAGE 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 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 Rea cto r Ves s el Wa ter L eve 1...................................

2-2 r

2.2 LIMITING SAFETY SYSTEM SETTINGS i

Reactor Protection System Instrumentation Setpoints..........

2-3 Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints................

2-4 BASES 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow......................B 2-1 THERMAL POWER, High Pressure and High Flow...................B 2-2 l

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Reactor Coolant System Pressure..............................B 2-5 t

Rea ctor Ves sel Water L eve 1................................... B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints..........B 2-6 4

PERRY - UNIT 1 iv Amendment No.20

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. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

. -. h SECTION PAGE 3/4.0 APPLICABILITY.............,..............................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN........................................

3/4 1-1 l-3/4.1.2 RE ACT I V IT Y AN0 MAL I ES...................................

3/4 1 - 2 1

l 3/4.1.3 CONTROL RODS Control Rod Operability................................

3/4 1-3 Control Rod Maximum Scram Insertion Times..............

3/4 1-6

_g Control Rod Scram Accumulators.........................

3/4 1-8 Control Rod Drive Coupling.............................

3/4 1-10 Control Rod Position Indication........................

3/4 1-12 Control Rod Drive Housing Support......................

3/4 1-14 3/4.1.4 CONTROL R0D PROGRAM CONTROLS Control Rod Withdrawa1.................................

3/4 1-15 Rod Pattern Control System.............................

3/4 1-16 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM..........................

3/4 1-18 Figure 3.1.5-1 Sodium Pentaborate Solution Concentration / Volume Require-ments.............................

3/4 1-20 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............

3/4 2-1 Figure 3.2.1-1 MAPFAC...........................

3/4 2-2 f

Figure 3.2.1-2 MAPFAC...........................

3/4 2-3 p

Figure 3.2.1-3 Maximum Average Planar Linear Heat GenerationRate(MAPLHGR)Versus Average Planar Exposure Initial Core Fuel Types BP8 SRB 219.......

3/4 2-4 PERRY - UNIT 1 v

Amendment No. 20

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

Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types BP8 SRB 176....... 3/4 2-5 Fig.are 3. 2.1-5 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Reload Core Fuel Types BS301E............

3/4 2-6 Figure 3.2.1-6 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Reload Core Fuel Types BS301F............

3/4 2-6a 3/4.2.2 MINIMUM CRITICAL POWER RATI0...........................

3/4 2-7 Figure 3.2.2-1 MCPR.............................

3/4 2-8 f

Figure 3.2.2-2 MCPR.............................

3/4 2-9 p

3/4.2.3 LINEAR HEAT GENERATION RATE............................

3/4 2-10 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 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 Requirements......................

3/4 3-7 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION....................

3/4 3-9 Table 3.3.2-1 Isolation Actuation Ins trumentation...................

3/4 3-11 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints.........

3/4 3-17 Table 3.3.2-3 Isolation System Instrumentation Response Time.....................

3/4 3-21 Table 4.3.2.1-1 Isolation Actuation Instruments-tion Surveillance Requirements....

3/4 3-23 PERRY - UNIT 1 vi Amendment No. 20

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE RADI0 ACTIVE EFFLUENTS (Continued)

)

3/4.11.3 SOLID RADWASTE TREATMENT................................

3/4 11-18 3/4.11.4 TOTAL D0SE..............................................

3/4 11-20 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING I

3/4.12.1 MONITORING PR0 GRAM......................................

3/4 12-1

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Table 3.12.1-1 Radiological Environm4ntal Monitoring Program.................

3/4 12-3 Table 3.12.1-2 Reporting Levels For Radio-a activity Concentrations In Environmental Samples..............

3/4 12-9 Table 4.12.1-1 Detection Capabilities For Environmental Sample Analysis Lower Limit of Detection...........

3/4 12-10 3/4.12.2 LAND USE CENSUS.........................................

3/4 12-13 3/4.12.3 INTER LABORATORY COMPARISON PR0G RAM......................

3/4 12-14 PERRY - UNIT 1 xvii i

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s BASES SECTION PAGE 3/4.0 APPLICABILITY............................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SH UT DOWN MARG I N......................................... B 3/ 4 1 - 1 3/4.1. 2 REACTIVITY AN0MALIES.................................... B 3/4 1-1 3/4.1.3 CONTROL R0DS............................................

B 3/4 1-2 3/4.1.4 CONTROL ROD 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 RATE.............. B 3/4 2-1 l

3/4.2.2 MINIMUM CRITICAL POWER RATI0............................ B 3/4 2-4 3/4.2.3 LINEAR HEAT GENERATION RATE.............:............... B 3/4 2-5 Bases Figure B 3/4 2.2-1 Power to Flow Operating Map... B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............... B 3/4 3-1 b

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 INSTRUMENTATION......................................... B 3/4 3-4 3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION....................... B 3/4 3-4 l

PERRY - UNIT 1 xviii Amendment No. 20 L

1 1.0 ' DEFINITIONS i

The following terms are defined so that uniform interpretation of these j

specifications may be achieved.

The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.

ACTION

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1.1 ACTION shall be that part of a Specification' which prescribes remedial measures required under designated conditions.

AVERAGE PLANAR EXPOSURE k

1. 2 The AVERAGE PLANAR EXPOSURE shall be applicabla to a specific planar 1

height and is equal to the sum of the exposure of all the fuel rods in the j

specified bundle at the specified height divided by the number of fuel rods in 1

the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and.is equal to the sum of the LINEAR HEAT GENERATION RATES for.all the fuel' rods in the specified bundle at the l

specified height divided by the number of fuel rods in the fuel bundle.

.j CHANNEL CALIBRATION

1. 4 A CHANNEL CALIBRATION shall be the. adjustment, as necessary, of the channel. output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL cal.IBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The l

CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

i CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

Thir, determination shall include, where pos-sible, comparison of the channel indication and/or status with other j

indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.G A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

PERRY - UNIT 1 1-1

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_]Vl DEFINITIONS

-CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incere instruments or reactivity contrnis within the reactor-pressure vessel with *,he vessel head removed and fuel in the vessel. Normal.

movement of the SRMs, '.RMs,- LPRMs, TIPS, or special ovable detectors is not considered a CORE ALTERATION.

Suspension of CORE Ai.TERATIONS shall not preclude-completion of the movement of a component to a safe conservative position.

CORE MAXIMUM FRACTION OF LIMITING POWER DENSIYY 1.8 The CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD) shall be the highest value of the FLPD which exists in t?.3. core.

CRITICAL' POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of-that power in the assembly wfiich is calculated by application of a General Electric critical power correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.10 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-I-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 heactor Sites."

DRYWELL INTEGRITY 1.11 DRYWELL INTEGRITY shall exist when:

a.

All drywell penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE automatic isolation system, or 2.

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

b.

The drywell equipment hatch is closed and sealed.

c.

The drywell head is installed and sealed.

d.

The drywell air lock is in compliance with the requirements of Specification 3.6.2.3.

e.

The drywell luage rates are within the limits of Specification 3.6.2.2.

PERRY - UNIT 1 1-2 Amendment No. 20

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS W

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 f %TED THERMAL POWER and the reactor vessel steam dome pressure less than 185 psig or core flow less than 10%

of rated flow, be in at least HOT SHUT 00WN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUA CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 with the reactor vessel steam dome pressure greater than 785 psig l

and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CUNDITIONS 1 and 2.

l ACTION:

With MCPR less than 1.07 and the reactor vessel steam dome pressure l

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements l

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 I

reactor coolant system pressure less than or equal to 1325 psig within l

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

j PERRY - UNIT 1 2-1 Amendment No. 20

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

REACTOR VESSEL WATER LEVEL I

2.1.4 The reactor vessel water level shall be above the top of the

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active irradiated fuel.

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APPLICABILITY: OPERATIONAL CONDITIONS 3, 4 and 5 a

ACTION:

With the reactor vessel water level at or below the top of the active ir-radiated fuel, manually initiate the ECCS to restore the water level.

Depressurize the reactor vessel, as necessary for ECCS operation.

Comply with the requirements of Specification 6.7.1.

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PERRY - UNIT 1 22

l 2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

i The fuel cladding, reactor pressure vessel and primary ~ system pipirg 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.

MCPR greater than 1.07 represents a conservative margin relative to the conditions required to maintain fuel l

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 incrementally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from 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.

2.1.1 THERMAL POWER, Low Pressure or Low Flow d

The use of the General Electric critical power correlations (Reference 1) are 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 estab-lishing a limiting condition on core THERMAL POWER with the following basis.

Since the pressure drop in the bypass reg:on is essentially all elevation heau, I

the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, 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 lbs/hr.

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 RATFD THERMAL POWER.

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

1 PERRY - UNIT 1 B 2-1 Amendment No. 20 o_ _ _

SAFETY LIMITS BASES l

2.1.2 THERMAL POWER, High Pressure and High Flow I

The fuel cladding integrity Safety Limit is set such that no fuel damage i

is calculated to occur if the limit is not violated.

Since the parameters which result in fuel damage are not directly observable during reactor j

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 and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the l

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 considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB (Reference 1), which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power.

The probability of the occurrence of boiling transition is detennined using a GE critical power correlation. This correla-tion is valid over the range of conditions used in the tests of the data used to develop the correlation.

Details of the fuel cladding integrity safety limit calculation are given in Reference 2.

Reference 2 provides the uncertainties used in the determination of the Safety Limit MCPR and of the nominal values of the parameters used in the Safety Limit MCPR statistical analysis.

1.

" General Electric BWR Thermal Analyu s Bases (GETAB) Data Correlation I

and Design Application," NED0-10958-A.

2.

" General Electric Standard Application for Reactor Fuel, GESTAR-II,"

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

PERRY - UNIT 1 B 2-2 Amendment No. 20

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This page intentionally left blank PERRY - UNIT 1 B 2-3 Amendment No. 20 1

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i PERRY - UNIT I B 2-4 Amendment No. 20

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POWER DISTRIBUTION LIMITS 3/4.2

.ER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE FLANAR LINEAR HEAT GENERATION RATE Q1IJINGCONDITIONFOROPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall not exceed the result obtained from multiplying the applicable MAPLHGR values

by the smaller of either the flow dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-1 f

or the power dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-2.

4 p

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

ACTION:

If at any time during operation it is determined.that an APLHGR is exceeding i

1 the result of the above multiplication, initiate corrective action within 15 minutes and restore APLHGR to within the required limits wit. in 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 .

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the above limits:

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 in one hour, andi 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 rentor is operating with a LIMITING CONTROL R0D PATTERN for APLHGR.

~

d.

The provisions of Specification 4.0.4 are not applicable.

These applicable MAPLHGR values are:

1) Those that have been approved for the respective fuel and lattice type as a function of the average planar exposure (as determined by the NRC approved methodology described in GESTAR-II) or

2) When hand calculations are required, the MAPLHGR as a function of the average planar exposum for the most limiting lattice (excluding natural uranium) shown in the Fi ares 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.1.6 for 9

the applicable type of fuel.

PERRY - UNIT 1 3/4 2-1 Amendment No. 20 l

q 1.1 1.0 =

y

.r 1

A 2r r

J' i

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MAPFACg = MIN (1.0,.0.4574 + 0.006758F) r 2

l 0.5 = -

i 0.4 -

i i

i 0

20 40 60 80 100 120 i

CORE FLOW (% RATED), F l

FLOW DEPENDENT MAPLHGR FACTOR (MAPFAC )

g FIGURE 3.2.1-1 PERRY - UNIT 1 3/42-2 Amendment No. 20 1

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CORE THERMAL POWER (% RATED), P l

POWER DEPENDENT MAPLHGR FACTOR (MAPFAC, )

FIGURE 3.2.1-2 i

i PERRY - UNIT 1 3/42-3 Amendment No.20

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POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than both MCPR and MCPR, limits at indicated core flow, THERMAL POWER, AT* and f

core average exposure compared to End of Cycle Exposure (EOCE)** as shown in Figures 3.2.2-1 and 3.2.2-2.

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

ACTION:

.With MCPR less than the applicable MCPR limit shown in Figures 3.2.2-1 and 3.2.2-2, 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 .

SURVEILLANCE REQUIREMENTS 4.2.2 MCPR shall be determined to be equal to or greater than the MCPR limit determined from Figures 3.2.2-1 and 3.2.2-2:

l 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 in one hour, and 1

i 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.

l d.

The provisions of Specification 4.0.4 are not applicable.

l 4

This AT refers to the planned reduction of rated feedwater temperature from nominal rated feedwater temperature (420'F), such as prolonged removal of l

feedwater heater (s) from service.

~

- End of Cycle Exposure (EOCE) is deftw.cd as 1) the core average exposures at which there is no longer sufficient reactivity to achieve RATED THERMAL POWER with rated core flow, all control rods withdrawn, all feedwater heaters in service and equilibrium Xenon, or 2) as specified by the fuel vendor.

i PERRY - UNIT 1 3/4 2-7 Amendment No. 20

+

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t l I il l I il l 1 l l l l l 1 l l 1 1 I I l 1 l l l l l l l l l l l l l l [ g MCPR

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CORE FLOW (% RATED), F FLOW DEPENDENT MCPR FACTOR (MCPR )

f FIGURE 3.2.2-1 I

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PERRY - UNIT 1 3/42-8 Amendment No. 20

~ w..

..l' M-

)

4 2.4 THERMAL POWER 255 f P f 405--

2.3 = ----

4

' CORE-FLOW'> 505 a

K 2.2 =

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m THERMAL POWER 255 f P S 405 2.0 --

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i.e.-

a lL 1.

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j (

h

1. 8 =

a.

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g THERMAL POWER 405 (, P f 705 1.6 =

/

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f 1.5 _

' = m

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THERMAL POWER P > 705 1.4 =

i,

BEFORE END OF CYCLE:

f All core average exposures w-1.3 "

and O f AT f 100*F and

'w Core flow f 1055 AFTER END OF CYCLE:

M 1.2 - ::

Core average exposure > EOCE 4 6 and 0 $ AT f 170*F and Core flow f 1055 1.1 -

I 1.0 '

I I

l' I

I l

0 20 40 60 80 100 120 CORE THERMAL POWER (% RATED), P POWER DEPENDENT MCPR FACTOR (MCPR )

p FIGURE 3.2.2-2 PERRY - UNIT 1 3/4 2-9 Amendment No. 20

POWER DISTRIBUTION LIMIT 3/4.2.3 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.3 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed:

a.

13.4 kw/ft for BP8x8R fuel.

b.

14.4 kw/ft for GE8x8EB fuel.

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 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 .

SURVEILLANCE REQUIREMENTS 4.2.3 LHGR's shall be determined to be equal to or less than the limit:

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 in one hour, 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 on a LIMITING CONTROL R0D PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

i l

PERRY - UNIT 1 3/4 2-10 Amendment No. 20

9 :.

4 3/4.1 REACTIVITY CCNTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made sub-critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are' controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be

.subcritical by at least R + 0.38% delta k/k or R + 0.28% delta k/k, as appropriate. The value of R in units of % delta k/k is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R sust be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different' methods of demonstration of the SHUTDOWN MARGIN.

The highest worth rod may be determined analyticeily or by test. The SHUTDOWN MARGIN is demonstrated by an insequence contral rod withdrawal at the beginning of life fuel cycle conditions, and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure; Observation of subcriticality in this condi-tion assur'es subcriticality with the most reactive control rod fully withdrawn.

This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.

3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns.

Since the comparisons are easily done, frequent checks are not an imposition on normal operations.

A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluateo. A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.

I o

PERRY - UNIT 1 8 3/4 1-1 i

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the safety analyses, and (3) limit the potential effects of the rod drop accident.

The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation.

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum.

The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanim could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specificatun, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than 1.07 during the' limiting power transient analyzed in Chapter 15 of the USAR.

This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than 1.07.

The occurrence of l

scram times longer than those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inopt.rable and Specification 3.1.3.1 then applies.

This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure.

Operability of the accumulator ensures that there is a means available to insert the control rods

)

even under the most unfavorable depressurization of the reactor.

PERRY - UNIT 1 B 3/4 1-2 Amendment No. 20

,c I

i 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200 F limit specified in 10 CFR 50.46.

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 GESTAR-II (Reference 1) will not be exceeded.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure depen-dent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor.

The MAPLHGR limits of Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3 are multiplied by the smaller of either the flow dependent MAPLHGR factor (MAPFAC ) or the power f

dependent MAPLHGR factor (MAPFAC ) corresponding to existing core flow and p

power state to assure the adherence to fuel mechanica1' design bases during the most limiting transient.

MAPFAC 's are determined using the three-f dimensional BWR simulator code to analyze slow flow runout transients.

MAPFAC 's are generated using the same data base as the MCPR to protect the p

p core from plant transients other than core flow increases.

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

Fuel Mechanical Design Analysis:

NRCapprovedmethods(specifiedin Reference 1) 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 1.

This bounding power history is used as the basis for the fuel design analysis MAPLHGR value.

LOCA Analysis: A LOCA analysis is performed in accor6ance with 10 CFR Part 50 Appendix K to demonstrate that the MAPLHGR values comply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant.

PERRY - UNIT 1 B 3/4 2-1 Amendment No. 20

e POWER DISTRIBUTION LIMITS BASES L

AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

Only the most limiting MAPLHGR values are shown in the Technical Specifi-cation figures for multiple lattice fuel. When hand calculations are required, j

these Technical Specification MAPLHGR figure values for that fuel type are used j

for all lattices in that bundle.

{

For some GE fuel bundle designs MAPLHGR depends only on bundle type, and burnup.

Other GE fuel bundles have MAPLHGRs that vary axially depending upon the specific combination of enriched uranium and gadolinia' that comprises 'a fuel bundle cross section at a particular axial node.

Each particular combination of enriched uranium and gadolinia, for these fuel bundle types, is called a lattice type by GE.

These particular fuel bundle types have MAPLHGRs that vary by lattice type (axially) as well as with fuel burnup.

Approved MAPLHGR values (limiting values of APLHGR) as a function of fuel and lattice types, and as a function of the average planar exposure are provided in Technical Specification Figures 3.2.1-3 and 3.2.1-6.

PERRY - UNIT 1 B 3/4 2-2 Amendment No. 20

p.

l I

i f

f This page intentionally left blank PERRY - UNIT 1 B 3/4 ?-3 Amendment No. 20

)

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.2 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07 and an analysis of the limiting I

operational transients.

For any abnormal operating transient analysis evalua-tion 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 transients have been analyzed to determine which result in the largest reduc tion in CRITICAL POWER RATIO (CPR).

The type of transients evaluated are documented in'the USAR and Reference 1.

The limiting transient yields the largest delta CPR.

When added to the Safety Limit MCPR, the required operating limit MCPR of Specification 3.2.2 is obtained. The power-flow map of Figure B 3/4 2.2-1 defines the analytical basis for generation of the MCPR operating limits.

The evaluation of a given transient begins with the system initial parameters shown in USAR Chapter 15 and/or Reference 1, and Cleveland Electric's November 28 and December 29, 1988 submittals that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate these events are described in Reference 1.

The purpose of the i1CPR and :1CPR is to define operating limits at other f

p than rated core flow and power conditions. At less than 100% of rated flow and power the required MCPR is the larger value of the MCPR and MCPR at the f

p existing core flow and power state.

The MCPR s are established to protect the f

core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.

Figure 3.2.2-2 also reflects the required MCPR values resulting from the analysis perfonned to justify operation with the feedwater temperature ranging from 420 F to 320*F at 100% RATED THERMAL POWER steady state conditions, and also beyond the end of cycle with the feedwater temperature ranging from 420*F and 250 F.

The MCPR s were calculated such that for the maximum core flow rate and f

the corresponding THERMAL POWER along a conservative steep generic power flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit.

Using this relative bundle power, the MCPRs were calculated at different points along this conservative steep power flow control line corresponding to different core flows.

The calculated MCPR at a given point of core flow is defined as MCPR.f

.s...

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

The MCPR s are established to protect the core from plant transients other p

than core flow increases, including the localized event such as rod withdrawal The MCPR s were calculated based upon the most limiting transient at the error.

p given core power level.

For core power less than or equal to 40% of RATED THERMAL POWER, where the EOC-RPT and the reactor scrams on turbine stop valve t

closure and turbine control valve fast closure are bypassed, separate sets of MRPR limits are provided for high and low core flows to account for the sig-p nificant sensitivity to initial core flows.

For core power above 40% of RATED THERMAL POWER, bounding power dependent MCPR limits were developed.

. 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 1

moderator void content will.be very small.

For all designated control rod patterns which may be employed at this point, operating 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 i

RATED THERMAL POWER is. sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The require-ment 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.

I 3/4.2.3 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.

GESTAR II, General Electric Standar.d Application for Reactor Fuel, NEDE-24011-P-A, (latest approved revision).

l PERRY - UNIT 1 8 3/4 2-5 Amendment No. 20

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ML20245J135

Contents

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2.0 INTRODUCTION

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2.0 INTRODUCTION

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