Text

Amend 31 to License NPF-22,revising Tech Specs to Include Operational Control for Cycle 2 Operation
	ML20215E070
Person / Time
Site:	Susquehanna
Issue date:	10/03/1986
From:	Adensam E; Office of Nuclear Reactor Regulation
To:	;
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ML20215E074	List: ML20215E070; ML20215E076;
References
	NUDOCS 8610150017
	Download: ML20215E070 (61)
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o UNITED STATES 8

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NUCLEAR REGULATORY COMMISSION o

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.N WASHINGTON, D. C. 20556

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PENNSYLVANIA POWER & LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-388 SUS 0VEHANNA STEAM ELECTRIC STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 31 License No. NPF-22 1.

The Nuclear Regulatory Comission (the Connission or the NRC) has found that:

l A.

The application for the amendment filed by the Pennsylvania Power &

Light Company, dated April 30, 1986, as supplemented on June 19, July 25, September 16 and 25,1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission'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 l

Chapter I; 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 Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-22 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 31 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.

mo18888th8ljr P

2-3.

This amendment is effective upon startup following the Unit 2 first refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION Elinor G. Adensam, Director BWR Project Directorate No. 3 Division of BWR Licensino

Enclosure:

Changes to the Technical Specifications Date of Issuance:

October 3,1986 4

y.-..---,-----,-.,--.-,---.----.---.---_m--------------_--____,-,,,v-.-

ATTACHKENT TO LICENSE AMENDMENT NO. 31 FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Rep 1tce the follewing pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document ecmpleteness..

REMOVE INSERT i

i ii ii (overleaf) iii iii (overleaf) iv iv xxi xxi (overleaf) xxit xxii xxiii xxiii xxiv xxiv xxy xxv xxvi xxvi xxvii xxvii 1-1 1-1 1-2 1-2 (overleaf) 1-3 1-3 1-4 1-4 (overleaf)

B 2-1 B 2-1 B 2-2 B 2-2 B 2-3 B 2-3 (overleaf) l B 2-4 B 2-4 3/4 1-1 3/4 1-1 (overleaf) 3/4 1-2 3/4 1-2 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-6a l

l l

REMOVE INSERT 3/4 2-7 3/4 2-7 3/4 2-B 3/4 2-8 3/4 2-8a 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4 2-10a 3/4 2-10b 3/4 3-39 3/4 3-39 (overleaf) 3/4 3-40 3/4 3-40 3/4 3-53 3/4 3-53(overleaf) 3/4 3-54 3/4 3-54 3/4 4-lb 3/4 4-lb (overleaf) 3/4 4-Ic 3/4 4-Ic 3/4 7-29 3/4 7-29(overleaf) 3/4 7-30 3/4 7-30 B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 B 3/4 1-4 B 3/4 1-4 B 3/4 1-5 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-3 B 3/4 2-3 B 3/4 2-4 B 3/4 2-5 B 3/4 4-1 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2 (overleaf)

B 3/4 7-3 B 3/4 7-3 (overleaf) l B 3/4 7-4 B 3/4 7-4 5-5 5-5 (overleaf) 5-6 5-6

a INDEX DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 ACTI0N.....................................................

3-1 1.2 AVERAGE EXPOSURE...........................................

1-1 l

1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.....I...........

1-1 1.4 CHANNEL CALIBRATION........................................

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

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

1-1 1.7 CO R E A LT E RAT I O N............................................

1-2 1.8 CRITICAL POWER RATI0.......................................

1-2 1.9 DOSE EQUIVALENT I-131......................................

1-2.

1.10 E-AVERAGE DISINTEGRATION ENERGY............................

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

1-2 1.12 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME..

1-2 1.13 FRACTION OF LIMITING POWER DENSITY.........................

1-3 1.14 FRACTION OF RATED THERMAL P0WER............................

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

1-3 1.16 GASEOUS RADWASTE TREATMENT SYSTEM..........................

1-3 1.17 I D ENT I F I E D L EA KAG E.........................................

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

1-3 1.19 LIMITING CONTROL R0D PATTERN...............................

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

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

1-4 1.22 MAXIMUM FRACTION OF LIMITING POWER DENSITY.................

1-4 1.23 MEMBER (S) 0F THE PUBLIC....................................

1-4 1.24 MINIMUM CRITICAL POWER RATI0...............................

1-4 l

1.25 0FFSITE DOSE CALCULATION MANUAL............................

1-4 t

l SUSQUEHANNA - UNIT 2 i

Amendment No. 31 l

INDEX DEFINITIONS SECTION DEFINITIONS (Continued)

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

1-4 1.27 OPERATIONAL CONDITION - CONDITION..........................

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

1-5 1.29 PRESSURE BOUNDARY LEAKAGE..................................

1-5 1.30 PRIMARY CONTAINMENT INTEGRITY..............................

1-5 1.31 PROCESS CONTROL PR0 GRAM....................................

1-5 1.32 PURGE-PURGING..............................................

1-5 1.33 RATED THERMAL P0WER........................................

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

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

1-6 1.36 ROD DENSITY................................................

1-6 1.37 SECONDARY CONTAINMENT INTEGRITY............................

1-6 1.38 SHUTDOWN MARGIN............................................

1-7 1.39 SITE B0VNDARY..............................................

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

1-7 1.41 SOURCE CHECK...............................................

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

1-7 1.43 THERMAL P0WER..............................................

1-7 1.44 TURBINE BYPASS SYSTEM RESPONSE TIME........................

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

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

1-8 1.47 VENTILATION EXHAUST TREATMENT SYSTEM.......................

1-8 1.48 VENTING....................................................

1-8 i

l SUSQUEHANNA - UNIT 2 i i

.-,,.3_

___--.m w,.

INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow...................

2-1 THERMAL POWER, High Pressure and High Flow................

2-1 Reactor Coolant System Pressure...........................

2-1 Reactor Vessel Water Leve1................................

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

2-3 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 Reactor Coolant System Pressure...........................

B 2-5 Reactor Vessel Water Leve1................................

B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints........ B 2-6 SUSQUEHANNA - UNIT 2 iii

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION

_ PAG _E.

3/4.0 A P P LI C AB I L I TY.............................................

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

3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES...................................

3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability................................

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

3/4 1-6 Control Rod Average Scram Insertion Times..............

3/4 1-7 Four Control Rod Group Scram Insertion Times...........

3/41-8 Control Rod Scram Accumulators.........................

3/4 1-9 Control Rod Drive Coupling.............................

3/4 1-11 Control Rod Position Indication........................

3/4 1-13 Control Rod Drive Housing Support......................

3/4 1-15 3/4.1.4 CONTROL R0D PROGRAM CONTROLS Rod Worth Minimizer....................................

3/4 1-16 Rod Sequence Control System............................

3/4 1-17 Rod Block Monitor......................................

3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM..........................

3/4 1-19 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............

3/4 2-1 3/4 2.2 APRM SETP0INTS.........................................

3/4 2-5 3/4.2.3 MINIMUM CRITICAL POWER RATI0...........................

3/4 2-6 3/4.2.4 LINEAR HEAT GENERATION RATE............................

3/4 2-10 GE FUEL...........................................

3/4 2-10 ENC FUEL..........................................

3/4 2-10a SUSQUEHANNA - UNIT 2 iv Amendment No. 31

= -

INDEX a

ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PR0 GRAM..................................

6-23 6.14 0FFSITE DOSE CALCULATION MANUAL..........................

6-24 6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS.....

6-24 i.

f l

SUSQUEHANNA - UNIT 2 xxi I

.n.----

-_----,,.,-n-.

INDEX LIST OF FIGURES FIGURE PAGE 1

3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

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

3/4 1-21 3.1.5-2 SODIUM PENTABORATE SOLUTION CONCENTRATION.........

3/4 1-22 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, GE FUEL TYPE 8CR183 (1.83% ENRICHED)................

3/4 2-2 3.2.1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.- AVERAGE PLANAR EXPOSURE, GE FUEL TYPE 8CR233 (2.33% ENRICHED)................

3/4 2-3 3.2.1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGd) VS. AVERAGE BUNDLE EXPOSURE, EXXON 9x9 FUEL...................................

3/4 2-4 3.2.2-1 LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, EXXON FUE.........

3/4 2-6a 3.2.3-1 FLOW DEPENDENT MCPR OPERATING LIMIT................

3/4 2-8 3.2.3-2 REDUCED POWER MCPR OPERATING LIMIT...............

3/4 2-9 3.2.4.2-1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE, EXXON 9x9 FUEL..........

3/4 2-10b 3.4.1.1-1 THERMAL POWER LIMITATIONS..........................

3/4 4-lb 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE...........................

3/4 4-18 4.7.4-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST........

3/4 7-15 B 3/4 3-1 REACTOR VESSEL WATER LEVEL........................

B 3/4 3-8 B 3/4.4.6-1 FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE..........................

B 3/4 4-7 5.1.1-1 EXCLUSION AREA....................................

5-2 5.1.2-1 LOW POPULATION ZONE...............................

5-3 5.1.3-la MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS......................

5-4 5.1.3-lb MAP DEFINING UNRESTRICTED AREAS FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS......................

5-5 SUSQUEHANNA - UNIT 2 xxii Amendment No. 31

.~

w e

e

INDEX LIST OF FIGURES (Continued)

FIGURE PAGE 6.2.1-1 0FFSITE ORGANIZATION..............................'

6-3 6.2.2-1 UNIT ORGANIZATION.................................

6-4 f

a 4

SUSQUEHANNA - UNIT 2 xxiii Amendment No. 31

TNDFX LIST OF TABLES TABLE PAGE 1.1 SURVEILLANCE FREQUENCY NOTATION...................

1-9 1.2 OPERATIONAL CONDITIONS............................

1-10 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS.........................................

2-4 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.........

3/4 3-2 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES..........

3/4 3-6 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................

3/4 3-7 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION...............

3/4 3-11 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS.....

3/4 3-17 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME....

3/4 3-21 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................................

3/4 3-23 3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...................................

3/4 3-28 3.3.3-2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.........................

3/4 3-31 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES......

3/4 3-33 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-34 3.3.4.1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION...................................

3/4 3-37 3.3.4.1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS.........................

3/4 3-38 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-39 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION...................................

3/4 3-42 3.3.4.2-2 END-0F-CYCLE RECIRCULATION PUMP TRIP SETPOINTS....

3/4 3-43 3.3.4.2-3 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.....................................

3/4 3-44 SUSQUEHANNA - UNIT 2 xxiv Amendment No. 31 l

l

INDEX LIST OF TABLES (Continued)

TABLE PAGE 4.3.4.2.1-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS.........................

3/4 3-45 3.3.5-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...................................

3/4 3-47 3.3.5-2 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.........................

3/4 3-49 4.3.5.1-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3/4 3-50 3.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION.................

3/4 3-52 3.3.6-2 CONTROL R0D BLOCK INSTRUMENTATION SETPOINTS.......

3/4 3-54 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................................

3/4 3-55 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION..............

3/4 3-58 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................

3/4 3-60 l

3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION...............

3/4 3-62 4.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................

3/4 3-63 3.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION.........

3/4 3-65 4.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................

3/4 3-66 3.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION........

3/4 3-68 4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................

3/4 3-69 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION...............

3/4 3-71 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................................

3/4 3-73 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION....................

3/4 3-78 3.3.7.10-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION...................................

3/4 3-83 SUSQUEHANNA - UNIT 2 xxv Amendment No. 31

s INDEX LIST OF TABLES (Continued)

TABLE PAGE 4.3.7.10-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING 3/4 3-85 INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

3.3.7.11-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION...................................

3/4 3-88 4.3.7.11-1 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........

3/4 3-91 3.3.9-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION...................................

3/4 3-97 3.3.9-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.........................

3/4 3-98 4.3.9.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........

3/4 3-99 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...

3/4 4-9 3.4.4-1 REACTOR C00LANTsSYSTEM CHEMISTRY LIMITS............

3/4 4-12 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM...................................

3/4 4-15 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM--

WITHDRAWAL SCHEDULE................................

3/4 4-19 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES...............

3/4 6-19

~

3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS..................................

3/4 6-35 3.7.6.5-1 FIRE HOSE STATIONS.................................

3/4 7-26 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE....................

3/4 8-7 4.8.1.1.2-2 UNIT 1 AND UNIT 2 DIESEL GENERATOR LOADING TIMERS..

3/4 8-8 4.8.2.1-1 BATTERY SURVEILLANCE REQUIREMENTS.................

3/4 8-15 3.8.4.1-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES....................

3/4 8-26 3.8.4.2-1 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION......................................

3/4 8-31 3.11.1.1-1 MAXIMUM PERMISSIBLE CONCENTRATION OF DISSOLVED OR ENTRAINED NOBLE GASES RELEASED FROM THE SITE TO UNRESTRICTED AREAS IN LIQUID WASTE................

3/4 11-2 SUSQUEHANNA - UNIT 2 xxvi Amendment No. 31

INDEX e

LIST OF TABLES (Continued)

TABLE PAGE 4.11.1.1.1-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM...........................................

3/4 11-3 4.11.2.1.2-1 RADI0 ACTIVE GASEOUS WASTE SAMPLING ANDt< ANALYSIS PROGRAM...........................................

3/4 11-10 3.12.1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM.....

3/4 12-3 3.12.1-2 REPORTING LEVELS FOR,RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES..........................

3/4 12-9 4.12.1-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE

~

ANALYSIS..........................................

3/4 12-10 83/4.4.6-1 REACTOR VESSEL TOUGHNESS..........................

B 3/4 4-6 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..............

5-8 6.2.2-1 MINIMUM SHIFT CREW COMPOSITION....................

6-5 SUSQUEHANNA - UNIT 2 xxvii Amendment No. 31

1.0 DEFINITIONS e

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.

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

AVERAGE EXPOSURE 1.2 The AVERAGE BUNDLE EXPOSURE shall be equal to the sum of the axially averaged exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in t.he 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 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 spe-cified 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 CALIBRA-TION 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, overlap-ping or total channel steps such that the entire channel is calibrated.

CHANNEL CHEC'K 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.

l l

SUSQUEHANNA - UNIT 2 1-1 Amendment No. 31

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

Normal movement of the SRMs, IRMs. TIPS, or special movable detectors is not considered a CORE ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude 9

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

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

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, 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 Reactor Sites."

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, 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 perform-ing its safety function, i.e., the valves travel to their required posi-tions, 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.

4 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.12 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete supression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

j a.

Turbine stop valves, and b.

Turbine control valves.

This total 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.

1 SUSQUEHANNA - UNIT 2 1-2

l 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 3.2.2 for l

that bundle type.

FRACTION OF RATED THERMAL POWER 1.14 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

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

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 LE,AKAGE.

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 R0D PATTERN 1.19 A LIMITING CONTROL R0D PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.20 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 2 1-3 Amendment No. 31

DEFINITIONS i

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.

Tra 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 LIMITI:4G POWER DENSITY (MFLPD) shall be the highest value of the FLPD which exists in the core.

MEMBER (S) 0F THE PUBLIC 1.23 MEMBER (S) 0F 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, l

occupational or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.24 The MINIMUM CRITICAL POWER RATIO (MCPR) 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 (ODCM) 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.

OPERABLE - 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 l

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.

4 SUSQUEHANNA - UNIT 2 1-4

e 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 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 the limit specified in Specification 2.1.2 for both GE and Exxon fuel. MCPR greater than the specified limit represents a conser-vative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive paterials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Al-though 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 pro-duct 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 incre-mental 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.

The MCPR fuel cladding integrity Safety limit assures that during normal operation and during antici-I pated operational occurrences, at least 99.9% of the fuel rods in the core do not experience tiransition boiling (ref. XN-NF-524(A)).

2.1.1 THERMAL POWER. Low Pressure or Low Flow i

The use of the XN-3 correlation is not valid for ali critical power

)

calculations at pressures below 785 psig or core flows les; than 10% of rated flow.

Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis.

Since the pressure drop in the bypass region is essentially all elevation head, t 6 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 108 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 108 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 RATED THERMAL POWER.

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

SUSQUEHANNA - UNIT 2 8 2-1 Amendment No. 31

SAFETY LIMITS BASES 2.1.2 IHERMAL POWER, High Pressure and High 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 tran-sition 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 99.9% 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 de-tailed 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 prac-tical 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 criti-cal power being estimated.

The assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounding high l

radial power factors 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 sus-tained 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 Commission and private or-ganizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach.

Much of the data in-dicates that LWR fuel can survive for an extended period of time in an environ-ment of boiling transition.

l SUSQUEHANNA - UNIT 2 B 2-2 Amendment No. 31 m

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SUSQUEHANNA - UNIT 2 B 2-0 Amendment No.2do i

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SUSQUEHANNA - UNIT 2 B 2-4 Amendment No. 31

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:

0.38% delta k/k with the highest worth rod analytically determined, a.

or b.

0.28% delta k/k with the highest worth rod determined by test.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.

ACTION:

With the SHUTDOWN MARGIN less than specified:

a.

In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWN MARGIN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In OPERATIONAL CONDITION 3 or 4, immediately verify all insertable control rods to be inserted and suspend all activities that could reduce the SHUTDOWN MARGIN.

In OPERATIONAL CONDITION 4, establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c.

In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS and other activities that could reduce the SHUTDOWN MARGIN and insert all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Establish SECONDARY CONTAIN-MENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

l SURVEILLANCE REQUIREMENTS 4.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than specified at any time during the fuel cycle:

l a.

By measurement, prior to or during the first startup after each refueling.

b.

By measurement, within 500 MWD /T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit.

l c.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after detection of a withdrawn control rod that is l

immovable, as a result of excessive friction or mechanical inter-ference, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod.

l l

l SUSQUEHANNA -UNIT 2 3/4 1-1

4 REACTIVITY CONTROL SYSTEMS 3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 4

3.1.2 The reactivity difference between the monitored core k,ff and the predicted core k,ff shall not exceed 1% delta k/k.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With the reactivity difference greater than 1% delta k/k:

l Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months perform an analysis to determine and explain the a.

cause of the reactivity difference; operation may continue if the difference is explained and corrected.

b.

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

SURVEILLANCE REQUIREMENTS 4.1.2 The reactivity difference between the monitored core k and the pre-dicted core k,ff shall be verified to be less than or equal t8I$% delta k/k:

a.

During the first startup following CORE ALTERATIONS, and b.

At least once per 700 MWD /MT of core exposure during POWER OPERATION.

l SUSQUEHANNA -UNIT 2 3/4 1-2 Amendment No. 31

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAP HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION for GE fuel and AVERAGE BUNDLE EXPOSURE for Exxon fuel RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3.*

APPLICABILITY:

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

ACTION:

With an APLHGR exceeding the limits 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 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 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

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

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

d.

The provisions of Specification 4.0.4 are not applicable.

See Specification 3.4.1.1.2.a for single loop operation requirements.

SUSQUEHANNA - UNIT 2 3/4 2-1 Amer.dment No. 31

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SUSQUEHANNA - UNIT 2 3/4 2-3 Amendment No. 31

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!: : : 25,000; :. Q .... PERMISSABLE '. '.. :. : 3 g g M9 S s, i e G .C 7 10d00 15dOO 20 BOO 25dOO 30 BOO 35$00 40000 0 5000 Average Bundle Exposure (MWD /MT) i i MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE EXXON 9X9 FUEL FIGURE 3.2.1-3 SUSQUEHANNA - UNIT 2 3/4 2-4 Amendment No. 31

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: Trip Setooint Allowable Value# 5 < (0.53W + 59%)T S < (0.58W + 62%)T S 5 (0.58W + 50%)T S 5 (0.58W + 53%)T RB RB where: S and S are in percent of RATED THERMAL POWER, RB W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million 1bs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. Vnere: a. The FRACTION OF LIMITING POWER DENSITY (FLPD) for GE fuel is the actual LINEAR HEAT GENERATION RATE (LHGR) divided by 13.4 per Specification 3.2.4.1, and b. The FLPD for Exxon 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 greater than or equal to 25% of RATED THERMAL POWER. ACTION: With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the ficw biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or S as l abovedetermined,initiatecorrectiveactionwithin15minutesandadjustgB,nd/or a l S to be consistent with the Trip Setpoint value* within 2 hours or reduce RB THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. l SURVEILLANCE REQUIREMENTS 1 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 trip setpoints verified to be within the above limits or adjusted, as required: l

With MFLPD greater than the FRTP during power ascension up to 90% of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100% times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10% cf 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.

i SUSQUEHANNA - UNIT 2 3/4 2-5 Amendment No. 31

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 4.2.2 (Continued) a. At least once per 24 hours,' b. Within 12 hours after completion of a THERMAL POWER increase of at least IS% of RATED THERMAL POWER, and Initially and at least once per 12 hours when the reactor is operating c. with MFLPD greater than or equal to FRTP. d. The provisions of Specification 4.0.4 are not applicable. l l SUSQUEHANNA - UNIT 2 3/4 2-6 Amendment No. 31 l [

i . O e 18 .. 0,0,. O,W 16.0. w w g g E gg l a e.:... l .C. M am w Wp g j4 9.O c Q 9 W (O W S I gg .. 43,200 g f S 9 6 .c O 10 -- JL g y. 8 0 10000.. 20$00 30$00 40$00 50000 Average Planar Exposure (MWD /MT) LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE EXXON FUEL l FIGURE 3.2.2-1 i SUSQUEHANNA - UNIT 2 3/4 2-6a Amendment No. 31

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the greater of the two values determined from Figure 3.2.3-1 and Figure 3.2.3-2. 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 determined above, initiate correc-tive action within 15 minutes and restore MCPR to within the required limit with-in 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.3.1 MCPR shall be determined to be greater than or equal to the applicable MCPR limit determined from Figure 3.2.3-1 and Figure.3.2.3-2: a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operatinq with a LIMITING CONTROL ROD PATTERN for MCPR. d. The provisions of Specification 4.0.4 are not applicable. l SUSQUEHANNA - UNIT 2 3/4 2-7 Amendment No. 31

l ,5 1.7 C J E CURVE A: EOC-RPT inoperable; E Main Turbine Bypass Operable j 7 CURVE B: Main Turbine Bypass Inoperable: 1 1.s ' EOC-RPT Operable c5 CURVE C: EOC-RPT and Main Turbine Bypass Operable [ ,e E J i at 1.5 a .E lii \\ A-1.41 O 14 t e n. m o l e 2 B ) \\ 1.31 l 13 1.30 C i i .~ 40 50 60 70 80 90 100 Total Core Flow (% OF RATED) P. E FLOW DEPENDENT MCPR OPERATING LIMIT m FIGURE 3.2.3-1 9

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o POWER DISTRIBUTION LIMITS j 3/4.2.4 LINEAR HEAT GENERATION RATE GE FUEL LIMITING CONDITION FOR OPERATION 3.2.4.1 The LINEAR HEAT GENERATION RATE (LHGR) for GE fuel shall not exceed 13.4 kw/ft. 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 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.4.1 LHGRs for GE fuel shall be determined to be equal to or less than the limit: a. At least once per 24 hours, 4 b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating on a LIMITING CONTROL R0D PATTERN for LHGR. d. The provisions of Specification 4.0.4 are not applicable. SUSQUEHANNA - UNIT 2 3/4 2-10 Amendment No. 31

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE ENC FUEL LIMITING CONDITION FOR OPERATION 3.2.4.2 The LINEAR HEAT GENERATION RATE (LHGR) for ENC fuel shall not exceed the LHGR limit determined from Figure 3.2.4.2-1. 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 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS I 4.2.4.2 LHGRs for ENC fuel shall be determined to be equal to or eless than the limit: a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and In'itially and at least once per 12 hours when the reactor is operating c. on a LIMITING CONTROL ROD PATTERN for LHGR. d. The provisions of Specification 4.0.4 are not applicable. l i 6 l SUSQUEHANNA - UNIT 2 3/4 2-10a Amendment No.31 t

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TABLE 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS m5E CHANNEL 1 E CHANNEL FUNCTIONAL CHANNEL i E TRIP FUNCTION CHECK TEST CALIBRATION 5 e 1. Reactor Vessel Water Level - S M R g Low Low, Level 2 2. Reactor Vessel Steam Dome Pressure - High NA M Q R. O s a

INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle rceirculation pump trip (EOC-RPT) system instrumenta-tion channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip i setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER. ACTION: a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value. b. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within one hour. c. With the number of OPERABLE channels two or more less than required I by the Minimum OPERABLE Channels per Trip System requirement for one trip system and: 1. If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within one hour. 2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable. d. With one trip system inoperable, restore the inoperable trip system { to OPERABLE status within 72 hours or evaluate MCPR to be equal to or greater than the applicable MCPR limit without EOC-RPT within 1 hour

or take the ACTION required by Specification 3.2.3.

With both trip systems inoperable, restore at least one trip system e. l to OPERABLE status within 1 hour or evaluate MCPR to be equal to or j greater than the applicable MCPR limit without EOC-RPT within 1 hour

l or take the ACTION required by Specification 3.2.3.
If MCPR is evaluated to be equal to or greater than the applicable MCPR limit without EOC-RPT within I hour, operation may continue and the provisions of t

Specification 3.0.4 are not applicable. SUSQUEHANNA - UNIT 2 3/4 3-40 Amendment No. 31 .. _ ~

..-.-2. .a. e l l _T_ABLE 3.3.6-1 (Continued) E CONTROL ROD BLOCK INSTRUMENTATION ACTION ACTION 60 Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3. ACTION 61 With the number of OPERABLE Channels: a. One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour. b. Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 1 hour. 4 ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within I hour. NOTES With THERMAL POWER 2 30% of RATED THERMAL POWER. With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. Not required when eight of fewer fuel assemblies (adjacent to the SRMs ) are in the core. (a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of RATED THERMAL POWER. (b) This function shall be automatically bypassed if detector count rate is 1 100 cps or the IRM channels are on range 3 or higher. (c) This function is automatically bypassed when the associated IRM channels are on range 8 or higher. 1 i (d) This function is automatically bypassed when the IRM channels are on range 3 or higher. (e) This function is automatically bypassed when the IRM channels are on range 1. 1 SUSQUEHANNA - UNIT 2 3/4 3-53 Amendment No.16

TABLE 3.3.6-2 i CONTROL POD 0109' IN5TRL%TFVATTOM SETPOINTS g TRIP FUNCTION TRIP SE1 POINT ALLOWABLE VALUE 1. R00 BLOCK MONITOR m i g a. Upscale ## $_ 0.66 W + 42% < 0.66 W + 45% l l E b. Inoperative NA HA c. Downscale _ 5/125 divisions of full scale _ 3/125 of divisions full scale ?. APRM j ez Q a. Flow Biased Neutron Flux - Upscale 5 0.58 W + 50%* 5 0.58 W + 53%* m b. Inoperative NA NA c. Downscale i d. Neutron Flux - Upscale -> 5% of RATED THERMAL POWER -> 3% of RATED THERMAL POWER Startup i 12% of RATED THERMAL POWER 5 14% of RATED THERMAL. POWER 3. SOURCE RANGE MONITORS 1 l a. Detector not full in NA NA 5 5 i b. Upscale 5 2 x 10 cps 5 4 x 10 cp, w l } c. Inoperative NA NA d. Downscale > 0.7 cps ** > 0.5 cps ** w f h 4. INTERMEDIATE RANGE MONITORS i a. Detector not full in NA NA l b. Upscale 5 108/125 divisions of full scale 5 110/125 divisions of full scale i c. Inoperative NA NA 1, d. Downscale _ 5/125 divisions of full scale _ 3/125 d.ivisions of full scale 5. SCRAM DISCHARGE VOLUME a. Water Level - High 5 44 gallons 5 44 gallons l g 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW ] g a. Upscale 5 108/125 divisions of full scale 5 111/125 divisions of full scale g b. Inoperative NA NA g c. Comparator 5 10% flow deviation 5 11% flow deviation = I o "The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow l we (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.

- Provided signal-to-noise ratio is > 2.

Otherwise, 3 cps as trip setpoint and 2.8 cps for allowable value. I

1. See Specification 3.4.1.1.2.a for single loop operation requirements.

l 1 1 ~.

] l l Figure 3.4.1.1.1-1' THERMAL POWER LIMITATIONS SO .......)..........'...............j.........j..........j..........j..........)..........b.........j..........j......... .......q... ......l..........).........>..................(..........(......... ..........)............. ..... R EGION GREATER TNAN UMIT. ~.~.:,..... ~.~..:,........+.>.........+........v......... t .t.......... rg) ..................)...............,........?........{.........4.......... . 4..... as== O ...t.. ...y......... ..t.........t......... .3 .t.. ...t... ..t., ............ 4.... < gg() 4... .4.. .........y........ ....g..... j...<....t........ y........g... ...t..... ...4.. .....4........ ..). ,4.... 4........ .g... .....L .....I....... .........................J.........; .............4........J...... ..i.... ...J, ss RA (

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REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1.2 .One reactor coolant recirculation loop shall be in operation with the pump speed 5 90% of the rated pump speed, and a. the following revised specification limits shall be followed: 1. Specification 2.1.2: the MCPR 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: - Trip Setpoint Allowable Value 5 0.58W + 55% 5 0.58W + 58%. 3. Specification 3.2.1: The MAPLHGR limits shall be the limits specified in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3, multiplied by 0.0. l 4. Specification 3.2.2: the APRM Setpoints shall be as follows: Trip Setpoint Allowable Value S < (0.58W + 55%)T S < (0.58W + 58%)T S $ (0.58W + 46%)T S $ (0.58W + 49%)T RB RB 7 5. Table 3.3.6 2: the RBM/APRM Control Rod Block Setpoints shall be as follows: a. RBM - Upscale Trip Setpoint Allowable Value 1 0.66W + 37% 5 0.66W + 40% l 5.a.1 and 5.a.2 shall be used in conjunction with the MCPR limits specified in Figures 3.2.3-la and 3.2.3-1b, respectively, b. APRM-Flow Biased Trip Setpoint Allowable Value 1 0.58W + 46% 1 0.58W + 49% b. APRM and LPRM*** neutron flux noise levels shall be less than three times their established baseline levels when THERMAL POWER 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 1bs/hr.aen THERMAL PCr.iR is greater than the limit specified in Figure 3.4.1.1.1-1. APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*, except during two loop operation.# ACTION: a. With no reactor coolant system recirculation loops in operation, take the ACTION required by Specification 3.4.1.1.1. SUSQUEHANNA - UNIT 2 3/4 4-Ic Amendment No. 31 1

PLANT SYSTEMS (' SURVEILLANCE REQUIREMENTS (Continued) 4.7.7.2 Each of the above required fire doors shall be verified OPERABLE by: Verifying the position of each closed fire door at least once per a. 24 hours. ' b. Verifying that doors with automatic hold-open and release mechanism.thg) are free of obstructions at least once per 24 hours. c. Verifying the position of each locked closed fire door at least once per 7 days, d. Verifying the OPERABILITY of the fire door supervision system by performing a CHANNEL FUNCTIONAL TEST at least once per 31 days. Inspecting the automatic hold-open, release and closing mechanism e. and latches at least once per 6 months. o O l l SUSQUEHANNA - UNIT 2 3/4 7-29

PLANT SYSTEMS 3/4.7.8 MAIN TURBINE BYPASS SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 The main turbine bypass system shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITION 1. ACTION: With the main turbine bypass system inoperable, restore the system to OPERABLE status within 2 hours or evaluate MCPR to be equal to or greater than the applicable MCPR limit without bypass within 1 hour

or take the ACTION required by Specification 3.2.3.

SURVEILLANCE REQUIREMENTS 4.7.8 The main turbine bypass system shall be demonstrated OPERABLE at least once per: 7 days by cycling each turbine bypass valve through at least one a. complete cycle of full travel, and b. 18 months by: 1. Perforraing a system functional test which includes simulated automatic actuation and verifying that each automatic valve actuates to its correct position. 2.~ Demonstrating TURBINE BYPASS SYSTEM RESPONSE TIME to be less than or equal to 0.30 second. "If MCPR is evaluated to be equal to or greater than the applicable MCPR limit without bypass within 1 hour, operation may continue and the provisions of Specification 3.0.4 are not applicable. 4 SUSQUEHANNA - UNIT 2 3/4 7-30 Amendment No. 31

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN 4 A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical 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 appro-i priate. The value of R in units of % delta k/k is the difference between the beginning of cycle shutdown margin minus the 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 l to provide for the different methods of demonstration of the SHUTDOWN MARGIN. 1 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 j 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 subcriticality with the most reactive control rod fully withdrawn. i l This reactivity characteristic has been a basic assumption in the analysis i 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 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,77). In the absence of any deviation in olant 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 operating 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 monitoring system for actual plant conditions. SUSQUEHANNA - UNIT 2 8 3/4 1-1 Amendment No. 31

4 REACTIVITY CONTROL SYSTEMS BASES REACTIVITY ANOMALIES (Continued) Since the comparisons are easily done, frequent checks are not an imposi-tion on normal operation. A 1% deviation in reactivity from that of the pre-dicted is larger than expected for normal operation, and therefore should be thoroughly evaluated. A deviation as large as 1% would not exceed the design conditions of the reactor. 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 l 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 re-quirements 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 mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechan-ical interference, operation of the reactor is limited to a time period which is ressonable 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. l The number of control rods permitted to be inoperable could be more than j the eight allowed by the specification, 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. i l l 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 speci-fied in Specification 2.1.? during the core wide transient analyzed in the cycle specific transient analysis report. This analysis shows that the negative reactivity rates resulting froni the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR i 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 inter-val is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem. t 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 i l SUSQUEHANNA - UNIT 2 B 3/4 1-2 Amendment No. 31 i . _ _-_.--_ _,, m, _ _ _ __._.

1 1 REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) reactor scram and will isolate the reactor coolant system from the containment when required. Control rods with inoperable accumulators are declared inoperable and'Spe-cification 3.1.3.1 then applies. This prevents a pattern of inoperable accumu-lators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be in-serted 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. Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of dete 1 mining that a rod is properly coupled -and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a tackup to the initial demonstration. ~ { In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod l position indication system must be OPERABLE. l The control rod housing support restricts the outward movement of a control i rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small mount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing. The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components. 3/4.1.4 CONTROL R0D PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control. SUSQUEHANNA - UNIT 2 B 3/4 1-3 Amendment No. 31 E

i REACTIVITY CONTROL SYSTEMS j l BASES i l CONTROL ROD PROGRAM CONTROLS (Continued) The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted. Parametric Control Rod Drop Accident analyses have shown that for a wide j range of key reactor parameters (which envelope the operating ranges of these j variables), the fuel enthalpy rise during a postulated control rod drop acci-i dent remains considerably lower than the 280 cal /gm limit. For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient, effective delayed neutron fraction, and maximum four-bundle local peaking factor are compared with the inputs to the parametric analyses to de-termine the peak fuel rod enthalpy rise. This value is then compared against l the 280 cal /gm design limit to demonstrate compliance for each operating cycle. ~ If cycle-specific values of the above parameters are outside the range assumed in the parametric analyses, an extension of the analysis or a cycle-specific analysis may be required. Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are provided in XN-NF-80-19 Volume 1. 4 The RBM is designed to automatically prevent fuel damage in the event of i erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provi.ded. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the cperator for withdrawal of control rods. 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the. reactor from full power to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted. To meet this objective i it is necessary to inject a quantity of boron which produces a concentration i of 660 ppm in the reactor core in approximately 90 to 120 minutes. A minimum l quantity of 4587 gallons of sodium pentaborate solution containing a minimum i of 5500 lbs. of sodium pentaborate is required to meet this shutdown require-l ment. There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing. The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak i and the required pumping rate is 41.2 gpm. The minimum storage volume of the j solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel. The temperature requirement for the sodium penetrate solution l 1s necessary to ensure that the sodium penetaborate remains in solution. With redundant pumps and explosive injection valves and with a highly reliable control rod scrcm system, operation of the reactor is permitted to i [ continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable. SUSQUEHANNA - UNIT 2 B 3/4 1-4 Amendment No. 31 1 --*----~..-----.---~~--,w., -,.mwen.-w,...--,--w.,.w... c.,-,p-m.m,.,-e,+%---

REACTIVITY CONTROL SYSTEMS BASES STANDBY LIQUID CONTROL SYSTEM (Continued) Surveillance requirements are established on a frequency that assures a .high reliability of the system. Once the solution is established, boron con-centration vill not vary unless more boron or water i's added, thus a check on the temperature and volume once each 24 hours assures that the solution is available for use. Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges. SUSQUEHANNA - UNIT 2 B 3/4 1-5 Amendment No. 31

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature 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 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 Specification AVERAGE PLANAR LINEAR HEAT GENER/. TION 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 Specifi-cation APLHGR for Exxon fuel is specified to assure the PCT following a postu-lated 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 Fig-ures 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 con-sistent 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. 3/4.7.2 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 instruments limit plant operations to the region covered by the transient and accident analyses. In addition, the APRM setpoints must be adjusted to ensure that >1% plastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (A00), including transients initiated from partial power operation. For Exxon 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 Exxon's Protection Against Fuel Failure (PAFF) line shown in Figure 3.4 of XN-NF-85-67, Revision 1. Figure 3.2.2-1 corresponds to the ratio of PAFF/1.2 under which cladding and fuel integrity is protected during A00's. SUSQUEHANNA - UNIT 2 8 3/4 2-1 Amendment No. 31

. ~. .j POWER DISTRIBUTION LIMITS BASES i APRM SETPOINTS (Continued) i For GE fuel the T factor used to adjust the APRM setpoints is based ~on the FLPD calculated by dividing the actual LHGR by the LHGR limit specified for GE fuel in Specification 3.2.4.1. 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as speci-ified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial con-i dition 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 Specifica-tion 2.2. To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL 4 POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase i in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification-3.2.3 is obtained and presented in Figure 3.2.3-1 and 3.2.3-2. l The evaluation of a given transient begins with the system initial parameters shown in the cycle specific transient analysis report that are input to a Exxon-core dynamic behavior transient computer program. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally i limiting bun'dle. The codes and methodology to evaluate pressurization and non-pressurization events are described in XN-NF-79-71 and XN-NF-84-105. The princi-pal result of this evaluation is the reduction in MCPR caused by the transient. Figure 3.2.3-1 defines core flow dependent MCPR operating limits which assure that the Safety Limit MCPR will not be violated during a flow increase tran-sient resulting from a motor generator speed control failure. The flow depend-ent MCPR is only calculated for the manual flow control mode. Therefore, automatic flow control operation is not permitted. Figure 3.2.3-2 defines the power dependent MCPR operating limit which assures that the Safety limit MCPR will not be violated in the event of a feedwater controller failure initiated 3 from a reduced power condition. Cycle specific analyses are performed for the most limiting local core wide tran-l sients to determine thermal margin. Additional analyses are performed to determine i the MCPR operating limit with either the Main Turbine Bypass inoperable or the EOC-RPT inoperable. Analyses to determine thermal margin with both the E0C-RPT inoperable and Main Turbine Bypass inoperable have not been performed. Therefore, operation in this condition is not permitted. SUSQUEHANNA - UNIT 2 8 3/4 2-2 Amendment No. 31

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the re-sulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit. 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975. SUSQUEHANNA - UNIT 2 8 3/4 2-3 Amendment No. 31

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 are multiplied by a factor of 0.0. This multiplication factor precludes extended operat' ion with one loop out of service. For single loop operation, the RBM and APRM setpoints are adjusted by a 7% 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 vibration. Surveillance on differential temperatures below the threshold limits of THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel nozzles, recirculation pumps and the vessel bottom head during extended opera-tion 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 prescribed in accordance with the recommendations of General Electric Service Information Letter No. 380, Revision 1, "BWR Core Thermal Hydraulic Stability," dated Febru-ary 10, 1984. An inoperable jet pump is not, in itself, a sufficient reason to declare a re-circulation loop inoperable, but it does, in case of a design basis accident, increase the, blowdown area and reduce the capability of reflooding the core; thus, the requirement.for shutdown of the facility with a jet pump inoperable. ~ Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation. 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. l In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50 F of each other prior to startup of an idle loop. The loop temperature must also be within 50 F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. 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 temperature differ-ence was greater than 145 F. SUSQUEHANNA - UNIT 2 B 3/4 4-1 Amendment No. 31

REACTOR COOLANT SYSTEM BASES l 3/4.4.2 SAFETY / RELIEF VALVES l l The safety valve function of the safety / relief valves operate to prevent the J 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 allowable values for the worst case upset transient. Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specifica-tion 4.0.5. l l 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 1 1 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS l The RCS leakage detection systems required by this specification are l provided to monitor and detect leakage from the reactor coolant pressure l 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 UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would 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 established 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, a SUSQUEHANNA - UNIT 2 8 3/4 4-2 Amendment No. 26

PLANT SYSTEMS BASES SNUBBERS (Continued) To provide assurance of snubber functional reliability one of three functional testing methods is used with the stated acceptance criteria: 1. Functionally test 10% of a type of snubber with an additional 10% tested for each functional testing failure, or 2. Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7.4-1, or 3. Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation. Figure 4.7.4-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in Quality Control and Industrial Statistics" by Acheson J. Duncan. Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions. The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and asso-ciated installation and maintenance records (newly installed snubber, seal replaced, sp' ring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. 3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring or boron measuring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism. SUSQUEHANNA - UNIT 2 B 3/4 7-3

PLANT SYSTEMS BASES 3/4 7.6 FIRE SUPPRESSION SYSTEMS The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety related equipment is located. The fire suppression system consists of the water system, spray and/or sprinklers, CO, bility of the fire suppression systems is adequa'te to minimize potential systems, Halon systems and fire hose stations. The collective capa damage to safety related equipment and is a major element in the facility fire protection program. i In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression. The surveillance requirements provide assurances that the minimum ^ OPERABILITY requirements of the fire suppression systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying the weight and pressure of the tanks. In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. The requirement for a twenty-four hour i report to the Commission provides for prompt evaluation of the acceptability of the corrective measures to provide adequate fire suppression capability for j the continue,d protection of the nuclear plant. i 3/4.7.7 FIRE RATED ASSEMBLIES The OPERABILITY of the fire barriers and barrier penetrations ensure that fire damage will be limited. These design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY. r 3/4.7.8 MAIN TURBINE BYPASS SYSTEM The required OPERABILITY of the main turbine bypass system is consistent with the assumptions of the feedwater controller failure analysis in the cycle specific transient analysis. I flSQUEHANNA - UNIT 2 B 3/4 7-4 Amendment No. 31 rw%-, www--,.,----c,..,.--.,v,-,v-i---c--we m-- irvv-m-----v" v4 " " ^ - - *

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'~ 9. FIGURE 5.1.3-1b l MAP DEFINING UNRESTRICTED AREAS FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS i SUSQUEMAN'lA - UNIT 2 5-5 l< (

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. l 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 l 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, 8 C, powder surrounded by a cruciform shaped stainless steel 4 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: 1. 1250 psig on the suction side of the recirculation pumps. 2. 1500 psig from the recirculation pump discharge to the jet

pumps, c.

For a temperature of 575*F. VOLUME 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',y, of 528 F. SUSQUEHANNA - UNIT 2 5-6 Amendment No. 31 /}}

ML20215E070

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2.0 INTRODUCTION

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ML20215E070

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2.0 INTRODUCTION

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