ML17146B002

Amend 72 to License NPF-14,changing Tech Specs in Support of Fuel Reload for Cycle 4 Operation

ML17146B002
Person / Time
Site:	Susquehanna
Issue date:	10/09/1987
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML17146B003	List: ML17146B002 ML17146B004
References
NUDOCS 8710230132
Download: ML17146B002 (49)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA P014ER

& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 72 License No.

NPF-14 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by the Pennsylvania Power Light Company, dated June 19,

1987, complies with the standards and requirements of the Atomic Er.ergy Act of 1954, as amended (the Act),

and the Comission's 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 regulations of the Commission; C.

There is reasonable assurance:

(i) 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:-set forth in 10 CFR Chapter I; D.

E.

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 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 Specifica-tions as indicated ir. the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No.

NPF-14 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.72 and the Environmental Protection Plan con-tained in Appendix B, are hereby incorporated in the license.

PP&L shall operate the facility in accordance with the Technical Specifica-tions and the Environmental Protection Plan.

87i0230132, 871009

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This license amendment is effective prior to startup for Cycle 4 operation.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 9i l987

/s/

Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II

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This license amendment is effective prior to startup for Cycle 4

operation.

FOR THE NUCLEAR REGULATORY COYiVISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 9, 1987 Walter R. Butler, Director Project Directorate 1-2 Division of Reactor Projects I/II

ATTACHMENT TO LICENSE AMENDMENT NO. 72 FACILITY OPERATING LICENSE NO.

NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

The revised pages a'e identified by Amendment number and contain vertical lines indicating the area of change.

The overleaf pages are provided to maintain document completeness.*

REMOVE i

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iii*

1V XX1 XX11 XXV l-l 1-2*

1-3 1-4*

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

3/4 2-3*

3/4 2-4 3/4 2-5 3/4 2-6*

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

INSERT 1

1 1*

111*

1V XX1 XX11 XXV l-l 2*

1-3 4*

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

3/4 2-3*

3/4 2-4 3/4 2-4a 3/4 2-5 3/4 2-5a 3/4 2-6*

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

3/4 2-9 3/4 2-9a 3/4 2-10a 3/4 2-lob 3/4 4-lb*

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

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

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

5 5*

5-6 3/4 2-9 3/4 2-9a 3/4 2-10a 3/4 2-lob 3/4 2-10c 3/4 4-lb*

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

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

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

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DEFINITIONS INDEX SECTION

1. 0 DEFINITIONS

l. 1 ACTION.

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1. 2 AVERAGE EXPOSURE.

1. 3 AVERAGE PLANAR LIHEAR HEAT GENERATION RATE 1.4 CHANNEL CALIBRATION...

1.5 CHANNEL CHECK.

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PAGE 1-1 1.6 CHANNEL FUNCTIONAL TEST...'..

. l-l 1.7 CORE ALTERATION.

1-2 1.8 CRITICAL POWER RATIO...............,....

1-2

l. 9 DOSE EQUIVALENT 1-131..

1-2

l. 10 K-AVERAGE DISINTEGRATION ENERGY....................;

1-2 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE TIME.........

1-2 1.12 EHD-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPOHSE TIME..

1-2 1.13 FRACTION OF LIMITIHG POWER DEHSITY................

1-3

1. 14 FRACTION OF RATED THERMAL POWER..

1-3 1.15 FREQUENCY NOTATION..

1-3 1.16 GASEOUS RADWASTE TREATMENT SYSTEM..

1-3

l. 17 IDENTIFIED LEAKAGE..........................................

1-3 1.18 ISOLATION SYSTEM RESPONSE TIME.........................

1-3 1.19 LIMITING CONTROL ROD PATTERN................................

1-3 1.20 LINEAR HEAT GEHERATIOH RATE..........................

1-3 1.21 LOGIC SYSTEM FUNCTIONAL TEST.....................:.........

1.22 MAXIMUM FRACTION OF LIMITIHG POWER DENSITY.................

1.23 MEMBER(S) OF TNE PUBLIC.....

1.24 MINIMUM CRITICAL POWER RATIO.............................,.

1.25 OFFSITE DOSE CALCULATIDHMANUAL.........~

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1-4 1-4 1-4 1-4 1-4 SUS)UEHANNA - UNIT 1 Amendment No. 72

DEFINITIONS 5ECT ION DEFINITIONS (Continued)

PAGE 1.26 OPERABLE - OPERABILITY.....................................

1-4 1.27 OPERATIONAL CONDITION - CONDITION........; ~...".............

1-4 1.28 PHYSICS TESTS..........................................

1.29 PRESSURE BOUNDARY LEAKAGE..............................

1-5 1-5

1. 30 PRIMARY CONTAIHMEHT INTEGRITY.............................

1-5 1.31 PROCESS CONTROL PROGRAM........;...........................

1-5

1. 32 PURGE-PURGING.......................

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. 1.33 RATED THERMAL POWER........................................

1-6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME......

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1 6 1.35 REPORTABLE EVENT...........................................

1-6

1. 36

1. 37 SECONDARY CONTAINHEHT INTEGRITY

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1 6 1.38 SHUTDOWN MARGIN...........:................................

1-7 1.39 SITE BOUNDARY..........................,..

1-7 1.40 SOLIDIFICATION.............................................

1-7.

1.41 SOURCE CHECK..................................

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1 7 1.42 STAGGERED TEST BASIS.....................

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1 7 l.43 THERMAL POWERo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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1.44 TURBINE BYPASS SYSTEM RESPONSE TIME........................

1-7 1.45 UNIDENTIFIED LEAKAGE.......................................

1-7 1.46 UNRESTRICTED AREA...........................

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1-8 1.47 VENTILATION EXHAUST TREATMENT SYSTEM...

1-8 l.48 V 1

A EHTING...................,...........................,....

a n

SUSQUEHANNA - UNIT 1 Amendment No. 29

INDEX SAFeeY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE DEFINITIONS (Continued)

2. 1 SAFETY LIMITS THERMAL POMER, Low Pressure or Low Flow...'.......

THERMAL POMER, High Pressure and High Flow.......

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Reactor Coolant System Pressure...........................

2-1 Reactor Vessel Mater Level................................

2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumtentation Setpoints.......

2-3 BASES

2. 1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow...................

B 2-1 THERMAL POKR, High Pressure and High Flow................

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

B 2-5 Reacto~ Vessel Mater Level................................

B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints........

B 2-6 SUS/UEHANNA - UNIT 1

LIMITING CONDITIONS FOR OPERATIOH AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 0 APPLICABILITY 3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4.l. 1 SHUTDOWN MARGIN..

3/4.1. 2 REACTIVITY ANOMALIES.....................

3/4.1. 3 CONTROL RODS Control Rod Operability

. Control Rod Maximum Scram Insertion Times..............

Control Rod Average Scram Insertion Times..............

Four Control Rod Group Scram Insertion Times.........

Control Rod Scram Accumulators.

Control Rod Drive Coupling.....,.....,.

Control Rod Position Indication.

Control Rod Drive Housing Support 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer.

Rod Sequence Control System.

Rod Block Monitor......

3/4.1.5 STANDBY LIQUID COHTROL SYSTEM..

3/4. 2 POWER DISTRIBUTION LIMITS PAGE 3/4 0-1 3/4 1-1 3/4 1-2 3/4 1-3 3/4 1-6 3/4 1-7 3/4 1-8 3/4 1-9 3/4 1-11 3/4 1-13 3/4 1-15 3/4 1-16 3/4 1-17 3/4 1-18 3/4 1-19 3/4.2.1 AVERAGE PLANAR LINEAR NEAT GENERATION RATE.............

3/4 2-1 3/4 2.2 APRM SETPOINTS 3/4.2. 3 MINIMUMCRITICAL POWER RATIO.

3/4.2.4 LINEAR HEAT GENERATION RATE E FUELo ~

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G ANF FUEL..............................

3/4 2-5 3/4 2-6 3/4 2-10 3/4 2-10a SUS(UEHANNA - UNIT 1 iv Amendment No.

72

LIST OF FIGURES IHDEX FIGURE 3.1. 5-1

3. 1. 5-2 3.2. 1-1 3.2. 1-2

3. 2. 1-3 3.2. 2-1

3. 2. 3"1 3.2.3-2 3.2.4.2-1

3. 2.4. 2-2 3.4.1.1-1 3.4.6.1-1 B 3/4 3"1 B 3/4.4.6-1 5.1. 1-1 5.1. 2-1 5.1. 3-1a 5.l. 3-1b 6.2. 1-1

6. 2. 2-1 SODIUM PEHTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS.................

SODIUM PEHTABORATE SOLUTION CONCENTRATION.....

THIS PAGE INTENTIONALLY LEFT BLANK.........~.....

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, GE FUEL TYPE BCR233 (2.33% ENRICHED)................

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE BUNDLE EXPOSURE, ANF Bx8 FUEL.

MAXIMUMAVERAGE PLAHAR LINEAR HEAT GEHERATION RATE (MAPLHGR) VS.

AVERAGE BUNDLE EXPOSURE, ANF 9x9 FUEL.

LINEAR HEAT GEHERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, AHF FUEL............

FLOW DEPEHDEHT MCPR OPERATING LIMIT.................

REDUCED POWER MCPR OPERATING LIMIT.

LINEAR HEAT GEHERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF Bx8 FUEL................

LIHEAR HEAT GEHERATIOH RATE (LHGR) LIMIT VERSUS AVERAGE PLAHAR EXPOSURE,, ANF 9x9 FUEL...............

THERMAL POWER LIMITATIOHS....

MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE........

REACTOR VESSEL 'MATER LEVEL........................

FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE.

EXCLUSION AREA.........

LOW POPULATION ZONE...............................

MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUEHTS...............".......

MAP DEFINING UHRESTRICTED AREAS FOR RADIOACTIVE'ASEOUS AHD LIQUID EFFLUEHTS..................,...

OFFSITE ORGANIZATION............,.................

UNIT ORGANIZATION..............

PAGE 3/4 1-21 3/4 1-22 3/4 2-2 3/4 2-3 I

3/4 2-4 3/4 2-4a 3/4 2-7 3/4 2-9 3/4 2-9a 3/4 2-10b 3/4 2-10c 3/4 4-1b 3/4 4-18 8 3/4 3-8 8 3/4 4-7 5"2 5a3 5a4 5-5 6-3 6-4 SUSQUEHANNA - UNIT 1 Amendment Ho. 72

LIST OF TABLES INDEX TABLE PAGE SURVEILLANCE FREQUENCY NOTATION 1"9 1.2 OPERATIONAL CONDITIONS..............;... ~.........

1-10

2. 2. 1-1 3.3. 1-1

3. 3. 1-2 4.3.1. 1-1.

3.3. 2-1 3.3. 2-2

3. 3. 2-3

4. 3.2. 1-1 REACTOR PROTECTION SYSTEH INSTRUMENTATION 5ETPOINTS REACTOR PROTECTION SYSTEM INSTRUHENTATION.........

2-4 I

3/4 3-2 REACTOR PROTECTION SYSTEH

RESPONSE

TIHES..........

3/4 3-6 REACTOR PROTECTION SYSTEH INSTRUHENTATION SURVEILLANCE REQUIREHENTS..

3/4 3-7 ISOLATION SYSTEM INSTRUHENTATION RESPONSE TIHE....

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREHENTS 3/4 3-21 3/4 3-23 ISOLATION ACTUATION INSTRUHENTATION...............

3/4 3-11 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS.....

3/4 3-17 3.3. 3-1 W

3.3.3 2

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUHENTATION SETPOINTS 3/4 3-28 3/4 3-31

3. 3. 3-3

4. 3. 3. 1-1 3.3.4.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUHENTATION SURVEILLANCE REQUIREHENTS.........

A%VS RECIRCULATION PUHP TRIP SYSTEM INSTRUMENTATION 3/4 3-34 3/4 3-37 EMERGENCY CORE COOLING SYSTEM RESPONSE TIHES......

3/4 3-33 3.3.4.1-2 ASS RECIRCULATION PUMP TRIP SYSTEH INSTRUMENTATION SETPOINTS..

3/4 3-38 SUSQUEHANNA - UNIT 1 Amendment No.'72

LIST OF TABLES Continued INDEX TABLE PAGE 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE 3/4 8->

4.8.1.1.2-2 UNIT 1 AND CONTIN DIESEL GENERATOR LOADING TIMERS..

3/4 8-8 4.8.2.1-1 3.8.4.1-1 3.8.4. 2-1

3. 11.1. 1-1 BATTERY SURVEILLANCE REQUIREMENTS PRIMARY CONTAINMEHT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES........ ~........

~..

NTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION........

MAXIMUMPERMISSIBLE COHCEHTRATION OF DISSOLVED OR EHTRAINED HOBLE GASES RELEASED FROM THE SITE TO UNRESTRICTED AREAS IH LIQUID WASTE................

3/4 8-14 3/4 8-24 3/4 8" 29 3/4 11-2 4.11.1.1.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND AHALYSIS PROGRAM o

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4.11.2.1.2-1 RADIOACTIVE GASEOUS WASTE SAMPLIHG AND ANALYSIS P ROG RAM

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3/4 11-3 3/4 11-10

3. 12. 1-1

3. 12. 1-2

4. 12. 1-1

$3/4.4.6-1 5.7.1-1 6.2.2-1 REPORTIHG LEVELS FOR RADIOACTIVITYCOHCEHTRATIONS IH ENVIRONMENTAL SAMPLES..........................

DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS REACTOR VESSEL TOUGHNESS.......

3/4 12-9 3/4 12-10 I

B 3/4 4-6 COMPONENT CYCLIC OR TRANSIENT LIMITS..............

5-8 MINIMUM SHIFT CREW COMPOSITION 6-5 RADIOLOGICAL EHVIRONMEHTAL MONITORING PROGRAM.....

3/4 12-3 SUSQUEHANNA - UNIT 1 haendment No.

72

l. 0 DEF INITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved.

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

ACTIOM 1.1 ACTION sha'll be that part of a Specification which prescribes reaediai measures required under designated conditions.

AVERAGE EXPOSURE 1,2 The AVERAGE BUNDLE EXPOSURE shall be equal to the sum of the axially aver-aged exposure of all the fuel rods in the specified bundle divided by the number of fuel rods in the fuel bundle.

The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE

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

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 CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

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

CHANNEL CHECK

1. 5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where

possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST

1. 6 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.

SUS/UEHANNA " UNIT 1 Amendment No. 72

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal; relocation or movement of fuel, sources, or reactivity controls wi hin the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Normal-movement of the

SRMs, IRMs, TIPs or special moveable detectors is not considered a

CORi ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude comple-tion of the movement of a component to a safe conservative position.

CRITICAL PO'n'ER RATIO

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

DOSE E UIVALENT I-131

'.9 DOSE E(UIVALENT 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 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those. listed in Table III of TI0-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

AVERAGE DISINTEGRATION ENERGY

l. 10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS

RESPONSE

TIME

1. 11 The EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE

TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of per-forming its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.

Times shall include diesel generator starting and sequence loading delays where applicable.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.12 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM

RESPONSE

TItlE shall be that time interval to complete supression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initia! movement of the associated:

a.

Turbine stop valves, and b.

Turbine control valves.

This to'al system response time consists of two components, the instru-mentation response time and the breaker arc suppression time.

These times may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

SUS(UEHANNA - UNIT I, 1-2 Amendment No.36

DEFINITIONS FRACTION OF LIMITING POWER DENSITY 1.13 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the LHGR specified in Section 3.2 '

for that bundle type.

FRACTION OF RA1'ED THERMAL POWER

1. 14 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

1. 15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intel vals defined =in Table l. l.

GASEOUS RADWASTE TREATMENT SYSTEM

1. 16 A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:

a.

Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a collecting tank, or b.

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

LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME 1.18 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions.

Times shall include diesel generator starting and sequence loading delays where applicable.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMITING CONTROL ROD PATTERN

1. 19 A LIMITING CONTROL ROD PATTERN shall be a patter n which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.20 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit. length.

SUSQUEHANNA - UNIT 1 1-3 Amendment No.

72

DEFINITIONS LOGIC SYSTEM FUNCTIONAL TEST 1.21 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, ie., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.22 The MAXIMUM FRACTION OF LIHITING POWER DENSITY (HFLPD) shall be the highest value of the FLPD which exists in the core.

HEMBER S

OF THE PUBLIC 1.23 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant.

This category does not include employees of the utility, its contractors or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

HINIHUM CRITICAL POWER RATIO 1 ~ 24 The MINIMUM CRITICAL POWER RATIO (HCPR) shall be the smallest CPR which exists in the core for each class of fuel.

OFFSITE DOSE CALCULATION MANUAL 1-.25 The OFFSITE DOSE CALCULATION MANUAL (ODCH) shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints and in the conduct of the environmental radiological monitoring program.

'PERABLE - OPERABILITY 1.26 A system, subsystem,

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

train, component or device to perform its function(s) are also capable of performing their related support function(s):

OPERATIONAL CONDITION - CONDITION 1.27 An OPERATIONAL CONDITION, i.e.,

CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

SUS(UEHANNA - UNIT 1

2.1 SAFETY LIMITS BASES

2. 0 INTRODUCTION 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 oper ations 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 the limit specified in Specifications

2. 1.2 for both GE and Exxon fuel.

MCPR greater than the specified limit 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 mate-rials 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 condi-tions 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 con-dition intended by design for planned operation.

The MCPR fuel cladding inte-grity Safety Limit assures that during normal operation and during anticipated operational occurr ences, at least

99. 9X of the fuel rods in the core do not experience transition boiling (ref. XN-NF-524(A)).

2. 1. 1 THERMAL POWER Low Pressure or Low Flow The use of the XN-3 correlation is valid for critical power calculations at pressures greater than 580 psig and bundle mass fluxes greater than 0.25 x 10 lbs/hr-ft2.

For operation at low pressures or low flows, the fuel cladding integrity Safety Limit is established by a limiting condition on core THERMAL POWER with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to assure a

minimum bundle flow for all fuel assemblies which have a relatively high power and potentially can approach a critical heat flux condition.

For the ANF 9x9 fuel design, the minimum bundle flow is greater than 30,000 lbs/hr, For the ANF and GE Bx8 fuel, the minimum bundle flow is greater than 28,000 lbs/hr.

For all designs, the coolant minimum flow and maximum flow area is such that the mass flux is always greater than 0.25 x 10 lbs/hr-fthm.

Full scale cri-tical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 10 lbs/hr-ft is 3. 35 Mwt or greater.

At SUS(UEHANNA - UNIT 1 B 2-1 Amendment No. /2

2.1 SAFETY LIMITS BASES

2. l. 1 THERMAL POWER Low Pressure or Low Flow (Continued) 25K thermal power a bundle power of 3.35 Mwt corresponds to a bundle radial peaking factor of greater than 3.0 which is significantly higher than the expected peaking factor.

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

SUS(UEHANNA - UNIT 1 B 2-la Amendment No. 7>

SAFETY LIM175 BASES

2. 1.2 THERMAL POWER Hi h Pressure and Hi h Flow Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure.

However, the existence of critical power, or boiling transition, is not a directly observable parameter in an operating reactor.

Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.

The margin for each fuel assembly is characterized by the critical power ratio (CPR),

which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power.

The minimum value of this ratio for any bundle in the core. is the minimum critical power ratio (MCPR).

The Safety Limit MCPR assures sufficient conservatism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9X of the fuel rods in the core would be expected to avoid boiling transition.

The margin between calculated-

= boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detail-ed statistical procedure which considers the uncertainties in monitoring the core operating state.

One specific uncertainty included in the safety limit is the uncertainty inherent in the XN-3 critical power correlation.

XN-NF-524 describes

- the methodology used in determining the Safety Limit MCPR.

The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual critical power being estimated.

As long as the core pressure and flow are within the range of validity of the XN-3 correlation (refer to Sec-tion 8 2. l. 1), the assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounding high radial power fac-tors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition.

Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that during sustained operation at the Safety Limit MCPR there would be no transition boiling in the core.

If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not necessarily be compromised.

Significant test data accumulated by the U.S. Nuclear Regulatory Coaeission and private organiza-tions indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach.

Much of the data indicates that LWR fuel can survive for an extended period of time in an environment of boiling transition.

SUS(UEHANNA - UNIT 1 B 2-2 Amendment Ko. 7>

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITIHG CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE for GE fuel and AVERAGE BUNDLE EXPOSURE for ANF fuel shall not exceed the liaits shown in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3.

APPLICABILITY:

OPFRATIOHAL CONDITION 1, when THERMAL POWER is greater than or ACTIOR:

gith an AFLHGR exceeding the lieita of Figure 3.2. 1-1, 3.2. 1-2, or 3.2. 1-3 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 RE UIREMEHTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures

3. 2.1-1, 3.2.1-2, and 3. 2.1-3:

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 coapletion of a THERMAL POWER increase of at least 15% of RATED THERNL 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 LINITING CONTROL ROD PATTERN for APLNGR.

d.

The provisions of Specification 4.0.4 are not applicable.

<<See Specification 3.4.1.1.2.a for single loop operation requirewents.

SUSQUEHANNA - UNIT 1 3/4 2-1 haendment Ho. 72

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MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE GE FUEL TYPES 8CR233 (2.33o/o ENRICHEO) fIGURE 3.2.1-1

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MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ANF 8X8 FUEL FIGURE 3.2.1-2

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O MAXlMUMAVERAGE PLANAR LlNEAR HEAT GENERATION RATE (MAPLHGR) YERSUS AYERAGE BUNDLE EXPOSURE ANF 9X9 FUEL RGURE 3.2.1-3

POWER DISTRIBUTION LIMITS 3/4. 2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION

3. 2. 2 The APRM flow biased simulated thermal power-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:

Tri Set oint Al 1 owable Value

+ 59K)T SRB

< (0.58W + 50X)T SRB < (0.5BW + 53K)T 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 MAXIMUM FRACTION OF LIMITING POWER DENSITY.

Where:

a.

The FRACTION OF LIMITING POWER DENSITY (FLPD) for GE fuel is the actual LINEAR HEAT GENERATION RATE (LHGR) divided by 13.4 per Specifica-tion 3.2.4.1, and b.

The FLPD for ANF fuel is the actual LHGR divided by the LINEAR HEAT GENERATION RATE from Figure 3. 2. 2"1.

T is always less than or equal to 1.0.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is gr eater than or ACTIDN:

With the APRM flow biased simulated thermal power-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 value" 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 25K 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 />.

st L

greater than the FRTP during power ascension up to 90K of RATED THERMAL POWER, rather, than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100K times

MFLPD, provided that the adjusted APRM reading does not exceed 100K of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10K of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

See Specification 3.4.1.1.2.a for single loop operation requirements.

SUSQUEHANNA - UNIT 1 3/4 2-5 Amendment No. 72

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-upscale scram and flow biased neutron flux-upscale control rod block'rip 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 15K 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.

SUSQUEHANNA - UNIT 1 3/4 2-5a Amendment No.

72

POWER DISTRIBUTION LH".JTS 3/4.2.3 HINJHUH CRJTJCAL PO'VER RATIO LIHITING CONDJT ION FOR OPERATION 3.2.3 The HJNIHPl CRITICAL POVER RATIO (HCPR) shall be greater than or equal tO the greater Of the t~O ValueS determined frOm Figure 3.2.3-1 and Figure 3.2. 3

?

APPLICABILITY:

OPERATIONAL CONDITION 1, ~en THERHAL POSER is greater than or ACT I 0 M:

With KCPR less than the applicable KCPR limit determined above, initiate cor-rective action vithin 15 minutes and restore KCPR 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 THERHAL PStER to less than 25K of RATED THERHAL PONR within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREHENTS

4. 2. 3. 1 H"PR shall be determined to be greater than or equal to the applicable HCPR limit determined from Figure

3. 2. 3-1 and Figure 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 1? hours after completion of a THERNL POMER increase of at least 15% of RATED TKERHAL PStER, 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 /> ~n vitn a LJHJTING CONTROL ROD PATTERN for HCPR.

the reactor is operating d.

The provisions of Specification

4. 0.4 are not applicable.

SUSQUEHANNA - UNIT 1 3/4 2-6 Amendment Ho.

64

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10000 20000 30000 40000 Average Planar Exposure (MNfD/MT) 6000 LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE ANF FUEL FIGURE 3.2.2-1 Svsqvehanna

- Unit 1

3/4 2-7 Amendment No. 72

TM}5 VAGE 15 DELETED SUSQUEMAkkA'- Uk1T 1

3/C 2-6 Amendment ko.

6C

1.7 CURVE A: EOC-RPT inoperable; Main Turbine Bypass Operable CURVE B: Main Turbine Bypass inoperable; EOC-RPT Operable CURVE C: EOC-RPT and Main Turbine Bypass Operable 1.39 1.31 1.28 60 80 70 80 Total Core Flow (% OF RATED) 90

$00 FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-1

CURVE A: EOC-RPT Inoperable:

Main Turbine Bypass Operable CURVE 8: Main Turbine Bypass Inoperable; EOC-RPT Operable CURVE C: EOC-RPT and Main Turbine Bypass Operable 30 40 80 60 80 70 Core Power (% OF RATED)

REDUCED POWER MCPR OPERATING LIMIT Figure 3.2.3-2 90 100

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE ANF FUEL LIMITING CONDITION FOR OPERATION 3.2.4. 2 The LINEAR HEAT GENERATION RATE (LHGR) for ANF fuel shall not exceed the LHGR limit determined from Figures 3.2.4.2-1 and 3.2.4.2-2.

APPLICABILITY:

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

With the LHGR of any fuel rod exceeding its applicabl'e limit from Figure 3.2.4.2-1 or 3.2.4.2-2, 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 RE UIREMENTS 4.2.4.2 LHGRs for ANF fuel shall be determined to be equal to or less than the I

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 15K 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 on a LIMITING CONTROL ROD,PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUS(UEHANNA - UNIT 1 3/4 2-10a Amendment No.

72

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56

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIHITING CONDITION FOR OPERATION

.- 3.4.1. 1.2 One reactor coolant recirculation loop shall be in operation with the pump speed ( 80K of the rated pump speed, and a ~

the following revised specification limits shall be followed:

1.

Specification 2.1.2:

the HCPR Safety Limit shall be increased to 1.07.

2, Table 2.2.1-1:

the APRM Flow-Biased Scram Trip Setpoints shall be as follows:

Tri Set oint W +

Allowable Value 3.

Specification 3.2. 1:

The HAPLHGR limits shall be as follows:

GE fuel:

the limits specified in Figure 3.2.1-1 multiplied by 0.81.

ANF fuel:

the limits specified in Figures 3.2.1-2 and 3.2.1-3 multiplied by 0.0.

4.

Specification 3.2.2:

the APRH Setpoints shall be as follows:

Tri Set oint Allowable Value SRB < (0.58M + 46K)T SRB

< (0.58W + 49K)T 5.

Table

3. 3.6-2:

the RBM/APRM Control Rod Block Setpoints shall be as follows:

4 a.

RBM - Upscale Tri Set oint Allowable Value

+

b.

APRM-Flow Biased Tri Set oint Allowable Value

+

b.

APRM and LPRH*"* neutron flux noise levels shall be less than three times their established baseline levels when THERMAL PtNER is greater than the limit specified in Figure 3/4. 1.1. 1-1.

c.

Total core flow shall be greater than or equal to 42 million lbs/hr when THERMAL POWER is greater than the limit specified in Figure 3.4.1.1.1-1.

'APPLICABILITY:

OPERATIONAL CONDITIONS 1>> and 2", except during two loop operati on.f SUSQUEHANNA - UNIT 1 3/4 4-1c Amendment No. 72

3/4. 1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1. 1 SHUTDOWN MARGI N 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 inad-vertent criticality in the shutdown condition.

Since core reactivity values wi 11 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 sub-critical by at least R + 0.38K delta k/k or R + 0.28K delta k/k, as appro-priate.

The value of R in units of X delta k/k is the difference between the cal-culated beginning of cycle shutdown margin minus the calculated minimum shutdown margin in the cycle, where shutdown margin is a positive number.

The value of R must 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 analytically or by test.

The SHUTDOWN MARGIN is demonstrated by control 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 condition assures subcritica-lity 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 aust also be determined anytime a control rod is incapable of insertion.

3/4.1.2 Reactivit Anomalies Since the SHUTDOWN MARGIN requirement is small, a careful check on actual reactor conditions compared to the predicted conditions is necessary.

Any changes in reactivity from that of the predicted (predicted core k ff) can be determined from the core monitoring system (monitored core k ff).

In the absence of any deviation in plant operating conditions or reactivity anomaly, these values should be essentially equal since the calculational methodologies are consistent.

The predicted core k ff is calculated by a 3D core simulation code as a function of cycle exposure.

This is performed for projected or anticipated reactor operat-ing states/conditions throughout the cycle and is usually done prior to cycle operation.

The monitored core k ff is the k ff as calculated by the core monitor-eff.

ing system for actual plant conditions.

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

h ]X deviation in reactivity from that of the predicted is larger than expected for normal operation, and therefore should be throug'hly evaluated.

A deviation as large as lX would not exceed the design conditions of the reactor.

SUS(UEHANNA - UNIT 1 B 3/4 1-1 Amendment No. 72

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 accident analysis, and (3) limit the potential effects of the rod drop accident.

The ACTION statements permit variations from the basic re-quirements 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 pill be investigated on a timely basis.

Damage within the control rod drive mechanism 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 oper ation 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.

4~

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a gener ic 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 the limit specified iri Specification 2.1.2 during the core wide transient analyzed in the cycle specific transient analysis report, This analysis shows that the negative re-ac'tivity 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 the limit specified in Specification 2.1.2.

The occurrence of scram times longer then those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in orde~ 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 inoperable 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 inserte+ with normal drive water pressure.

Operability of the accu-mulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

SUSQUEHANNA " UNIT 1 8 3/4 1-2 Amendment No.

57

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3/4.2 POWER DISTRIBUTION LIMITS BASES l

The specifications of this section assure that the peak cladding tempera-ture 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 following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

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.

For GE fuel, 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 dependent steady state gap conductance and rod-to-rod local peaking factor.

The Technical Speci-fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for GE fuel is this LHGR of the highest powered rod divided by its local peaking factor which results in a calculated LOCA PCT much less than 2200 F.

The Technical Specification APLHGR for ANF fuel is specified to assure the PCT following a postulated LOCA will not exceed the 2200'F limit.

The limiting value for APLHGR is shown in Figures

3. 2. 1-1,

3. 2. 1-2, and 3. 2. 1-3.

The calculational procedure used to establish the APLHGR shown on Figures 3.2. 1-1, 3.2. 1-2, and 3.2.1-3 is based on a loss-of-coolant accident analysis.

The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR 50.

These models are described in Reference 1 or XN-NF-80-19, Volumes 2, 2A, 2B and 2C.

3/4. 2. 2 APRM SETPOINTS The flow biased simulated thermal power-upscale scram setting and flow biased simulated thermal power-upscale control rod block functions of the APRM instru-ments limit plant operations to the region covered by the transient and accident analyses.

In addition, the APRM setpoints must be adjusted to ensure that

)1X plastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (AOO), including transients initiated from partial power operation.

For ANF fuel the T factor used to adjust the APRM setpoints is based on the FLPD calculated by dividing the actual LHGR by the LHGR obtained from Figure 3.2. 2-1.

The LHGR versus exposure curve in Figure 3. 2. 2-1 is based on ANF's Protection Against Fuel. Failure (PAFF) line shown in Figure 3.4 of XN-'NF-85-67(A), Revision 1.

Figure 3.2.2-1 corresponds to the ratio of PAFF/1.2 under which cladding and fuel integrity is protected during AOOs.

For GE fuel the T factor used to adjust the APRM setpoints is based on the FLPD calculated by dividing the actual LHGR by'he LHGR limit specified for GE fuel in Specification 3.2.4.1.

SUSQUEHANNA - UNIT 1 B 3/4 2-1 Amendment No. 72

PO'E'ER DlSTRI BUTIOH LIHITS BASES 3/4.2.3 HIHIHUH CRITICAL POSER RATIO I.

The required operating limit HCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit HCPR, and an analysis of abnormal operational transients.

For any abnorma1 operating transient analysis evaluation with the initial con-dition of the reactor being at the steady state operating limit, it is required that%he resulting HCPR does not decrease below the Safety Limit HCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POMER RATIO (CPR).

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature dectease.

The limiting transient yields the largest delta HCPR.

- Mnen added to the Safety Limit HCpR, the required minimum operating limit HCpR

~of Specification 3.2. 3. is obtained and presented in Figures 3.2.3-1 and 3.2. 3-2.

The evaluation of a given transient begins with the system initial param-eters shown in the cycle specific transient analysis report that are input to a

Exxon-cote dynamic behavior transient computer program.

The outputs of this program along winch the initial HCPR form the inpu'or further analyses of the

" " -..ally limiti.g bundle.

The codes and methodology t". evaluate pressuriza-

~~"" end non.pressurization events are described in XH-HF-1'9-71 and XH-HF-84-105.

The principal result of this evaluation is the reduction in HCPR caused by the transient.

Figure 3.2.3-1 defines core flov dependent HCPR operating limits which e<<ure that the Safety Limit HCPR will not be exceeded during a flow increase transient resulting from a motor-generator speed control failure.

The flov dependent HCPR is only calculated for the manual flow control mode.

Therefore, automatic flow control operation is aot permitted.

Figure 3.2.3-2 defines the power dependent HCPR operating limit which assures that the Safety Limit HCPR will not be exceeded.in the event of a feedwater controller fai lure initiated from a reduced power condition.

Cycle specific analyses are performed for the most limiting local and core wide transients to determine therma) margin.

Additional analyses are performed to determine the HCPR operating limit with either the Hain Turbine Bypass in-operable or the EOC-RPT inoperablc.

Analyses to determine thermal margin with both the EOC-RPT inoperable aqd Hain Turbine Bypass inoperable have not been pe~formed.

Therefore, operation in this condition is not permitted.

At THERHAL POMER levels less than or equal to 25K of RATED THERHAL POIER, 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 indi-cates that the resulting HCPR value is in excess of requirements by a consider-able margin.

During initial start-up testing of the plant, a

HCPR evaluation SUS(UEHAHHA - UNIT 1 B 3/4 2-2 Amendment Ho.

54

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4. 1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and found acceptable, provided that the unit is operated in accordance with Specification 3.4.1.1.2.

For single loop operation, the MAPLHGR limits for ANF fuel are multi-plied by a factor of 0 ~ 0.

This multiplication factor precludes extended opera-tion with one loop out of service.

For single loop operation, the RBM and APRM setpoints are adjusted by a 7X decrease in recirculation drive flow to account for the active loop drive flow that bypasses the core and goes up through the inactive loop jet pumps.

Surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive reactor vessel internals vibra-tion.

Surveillance on differential temperatures below the threshold limits on THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel

nozzles, recirculation pumps and the vessel bottom head during extended operation in the single loop mode.

The threshold limits are those values which will sweep up the cold water from the vessel bottom head.

THERMAL POWER, core flow, and neutron flux noise level limitations are pre-scribed in accordance with the recommendations of General Electric Service Information Letter No. 380, Revision 1, "BWR Core Thermal Hydraulic Stability,"

dated February 10, 1984.

An inoperable jet pump is not, in itself, a sufficient reason to declare a

recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the cote; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump peiformance on a prescribed schedule for significant degradation.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two loop operation.

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during the loop operation, continued operation is permitted in the single loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50F of each other rior to startup of an idle loop.

The loop temperature must also be within 0

F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles, Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the tem-perature difference was greater than 145'F.

SUS(UEHANNA - UNIT 1 B 3/4 4-1

.Amendment No. 72

3/4.4 REACTOR COOLANT SYSTEM BASES Continued 3/4.4. 2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code.

A total of 10 OPERABLE safety-relief valves is required to limit reactor pressure to within ASME III allow" able values for the ~orst case upset transient.

Oemonstration of the safety/relief va)ve liftsettings will occur only during shutdown. and will be performed in accordance with the provisions of Specification 4. 0.5.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE

, 3/4.4.3.

1 LEAKAGE OETECTIOH SYSTEMS The RCS leakage detection systems required by this specification are

!"provided to monitr r and detect leakage from the reactor coolant pressure

.::,boundary.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been

'based on the predicted and experimentally observed behavior of cracks in

-..'.:pipes.

The normally expected background leakage due to equipment design and

,+ the detection capability of the instrumentation for determining system leakage

,."-was also considered.

The evidence obtained from experiments suggests that for

"'-'leakage somewhat greater than that specified for UHIOENTIFIED LEAKAGE the

, probability is small. that the imperfection or crack associated with such leakage

~ould grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are establiihed to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

SUSQUEHANNA - UNIT 1 B 3/4 4-2 Amendment No. 56

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FIGURE 5.1.3-1b NAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS SUSQUEHANNA - UNIT 1 5-5 Amendment No. 29

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES

5. 3. 1 The reactor core shall contain 764 fuel assemblies with each fuel assembly containing 62 or 79 fuel rods and two water rods clad with Zircaloy -2.

Each fuel rod shall have a nominal active fuel length of 150 inches.

The initial core loading shall have a maximum average enrichment of 1.90 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum average enrichment of 4. 0 weight percent U-235.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 control rod. assemblies, each consisting of a cruciform array of stainless steel tubes containing 143 inches of boron carbide, B4C, powder surrounded by a cruciform shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4. 1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

l.

1250 psig on the suction side of the recirculation pumps.

2.

1500 psig from the recirculation pump discharge to the jet pumps.

For a temperature of 575 F.

VDLUNE 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal T

of 528 F.

SUSQUEHANNA - UNIT 1 5-6 Amendment No. 72