ML17157C440

Amends 126 & 95 to Licenses NPF-14 & NPF-22,respectively, Changing TSs to Remove cycle-specific Parameter Limits Per GL 88-16 & Modify Section 5.3.1 Per GL 90-02

ML17157C440
Person / Time
Site:	Susquehanna
Issue date:	08/04/1993
From:	Boyle M Office of Nuclear Reactor Regulation
To:
Shared Package
ML17157C441	List: ML17157C440 ML17157C442
References
GL-88-16, GL-90-02, GL-90-2, NUDOCS 9308250352
Download: ML17157C440 (77)

Text

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

& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION UNIT 1

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

126 License No.

NPF-14 1.

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

A.

The application for the amendment filed by the Pennsylvania Power Light Company, dated December 18,

1992, as supplemented by telecopy dated January 28,

1993, and by letters dated March 25, and May 20,

1993, 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 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 Commission's regulations and all applicable requirements have been satisfied.

9308250352 930804 PDR

• DOCK 05000387 P

PDR

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-14 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

126 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

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

3.

This license amendment is effective as of its date of issuance and is to be implemented within 30 days after its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 4, 1993 Michael L.

Bo A

ing Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO.

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

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

The overleaf pages are provided to maintain document completeness.*

REMOVE INSERT ill lv V

Vl Xlx XX Xxl Xxll 1-1 1-2 1-3 1-4 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-4a 1

1 1*

111*

lv v*

Vl XlX*

XX Xxl XX11*

1'-1*

1-2 1-3 4*

3/4 2-1 3/4 2-5 3/4 2-Sa 3/4 2-6 3/4 2-7 3/4 2-2 3/4 2-3 3/4 2-4

UNIT 1 3/4 2-8 3/4 2-9 3/4 2-9a 3/4 2-9b 3/4 2-9c 3/4 2-10 3/4 2-10a 3/4 2-10b 3/4 4-lb 3/4 4-lc 8 3/4 1-1 8 3/4 1-2 8 3/4 1-3 8 3/4 1-4 8 3/4 2-1 8 3/4 2-2 8 3/4 4-1 8 3/4 4-la 5-5 5-6 3/4 2-5 3/4 4-lb 3/4 4-lc 8 3/4 l-l*

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

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

5 5*

5-6 6-20a 6-20b

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INDEX DEFINITION E TION 1.

DEFINITI N 1

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1.2 AVERAGE EXPOSURE..................,.. ~.......... ~....

1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE

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

1.5 CHANNEL CHECK 1.6 CHANNEL FUNCTIONAL TEST.............. ~................

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1.7 CORE ALTERATION 1.7A CORE OPERATING LIMITS REPORT 1.8 CRITICAL POWER RATIO 1.9 DOSE EQUIVALENT I-131 1.10 E-AVERAGE DISINTEGRATION ENERGY..... ~...................

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1-2 1-2 1-2 1-2 1-2 1.1 1

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME

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1-2 1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME

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1-3 1.14 FRACTION OF RATED THERMAL POWER.......:..............'...

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1-3 1.15 FREQUENCY NOTATION

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

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1-3 1.1 7 IDENTIFIED LEAKAGE........................... ~.........

1-3 1.1 8 ISOLATION SYSTEM RESPONSE TIME

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

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

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1-3 1.21 LOGIC SYSTEM FUNCTIONAL TEST

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1.22 MAXIMUMFRACTION OF LIMITINGPOWER DENSITY 1.23 MEMBER(S) OF THE PUBLIC............ ~... ~................

1.24 MINIMUMCRITICAL POWER RATIO 1.25 OFFSITE DOSE CALCULATIONMANUAL 1-4 1-4 1-4 1-4 1-4 SUSQUEHANNA - UNIT 1 Amendment No.'P, 126

DEFIHIl'IOHS SECT ION DEFIHITIONS (Continued)

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

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

1-4

1. 28 I.29 PHYSICS TESTS..............................................

1-5 PRESSURE BOUNDARY LEAKAGE..................................'-5 1'.

30 PRIMARY CONTAIIRENT INTEGRITY..............................

1-5 lo 31 PROCESS CONTROL PROGRAMo

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

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

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

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1. 36 RO0 DENS ITYo ~

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1. 37 SECONDARY CONTAINMENT INTEGRITY............................

1-6

1. 38 SHUTDOWN MARGIN............................................

1-7 1.39 SITE BOUNDARY...........................;...............

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

1-7 1.41 SOURCE CKECKo

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

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~ I 7 1.44 TURBIHE BYPASS SYSTEM RESPONSE TIME........................

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1. 45 UNIDENTIFIED LEAKAGE.......................................

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

1-8 1.47 VENTILATION EXHAUST TREATMEHT SYSTEM..................,....

1-8 1.48 VEHTING.

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SUSQUEHANNA - UNIT 1 Amendment No. 29

INOEX SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS SECTION

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

THERMAL POWER, High Pressure and High Flow........

Reactor Coolant System Pressure.

Reactor Vessel Water Level.

PAGE 2-1 2" 1 2-1

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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 P(WER, High Pressure and High Flow................

B 2-2 Reactor Coolant Systei Pressure...........................

B 2-3 Reactor Vessel Water Level................................

B 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instr~ntation Setpoints........

B 2-4 SUSQUEHANNA - UNIT 1 Aeenchent No.

82 AUG ~0 i988

t A

INDEX LIMITIN NDITI N FOR PERATI N AND RVEILLANCERE IREMENT

~ET~IN 4.0 PPLI ABILITY

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3 4.1 REA TIVITY NTR L

Y TEM 3/4.1.1 SHUTDOWN MARGIN 3/4.1.2 REACTIVITYANOMALIES 3/4.1.3 CONTROL RODS

..... 3/4 1-1

. 3/4 1-2 Control Rod Operability Control Rod Maximum Scram Insertion Tines Control Rod Average Scram Insertion Times...

Four Control Rod Group Scram Insertion Times Control Rod Scram Accumulators

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~ 3/4 1 9 Control Rod Drive Coupling Control Rod Position Indication

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

3/4.1.4 CONTROL ROD PROGRAM CONTROLS

. 3/4 1-15 Rod Worth Minimizer

. 3/4 1-16 Rod Sequence Control System Rod Block Monitor 3/4 1-17

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3/4 1-19 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................

3/4 2-1 3/4.2.2 APRM SETPOINTS 3/4.2.3 MINIMUMCRITICAL POWER RATIO 3/4.2.4 LINEAR HEAT GENERATION RATE 3/4 2-2

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3/4 2 4

. ~... 3/4 2-5 SUSQUEHANNA - UNIT 1 IV Amendment No. 99,

INDEX LIMITING CONOITIONS FOR OPERATION ANO SURVEILLANCE RE UIREHENTS SECTION 3/4. 3 INSTRUMENTATION PAGE 3/4. 3.

1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............

3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION...............

3/4 3-9 3/4. 3. 3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION................

3/4 3-27 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation..

End-of-Cycle Recirculation Pump Trip System Instrumentatson..................

3/4 3-36 3/4 3"40 3/4. 3. 5 3/4. 3. 6 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION,.....................................

CONTROL ROO BLOCK INSTRUMENTATION....................

3/4 3-46 3/4 3-51 3/4. 3. 7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation........

Seismic Monitoring Instrumentation...

Meteorological Monitoring Instrumentation............

Remote Shutdown Monitoring Instrumentation...........

Accident Monitoring Instrumentation..................

Source Range Monitors................................

Traversing In-Core Probe System..............

Chlorine Oetection System....................

Fire Detection Instrumentation.

Radioactive Liquid Effluent Monitoring nstrumentation......................................

I Radi oacti ve Gaseous Effluent Monitoring Instrumentation......................................

Loose-Part Oetection System..........................

3/4 3-57 3/4 3-61 3/4 3-64 3/4 3-67 3/4 3-70 3/4 3-74 3/4 3-75 3/4 3-76 3/4 3-77 3/4 3-81 3/4 3-86 3/4 3-93 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEH...................

3/4 3-94 3/4. 3. 9 FEEOMATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.......................................

3/4 3-95 SUS(UEHANNA - UNIT 1

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INDEX LIIVIITINGCONDITIONS FOR OPERATiON AND SURVEILLANCEREQUIREMENTS

~ECTI N

4.4 REA T R

LANT Y TEM 3/4.4.1 'ECIRCULATIONSYSTEM Recirculation Loops - Two Loop Operation Recirculation Loops - Single Loop Operation Jet Pumps Recirculation Pumps Idle Recirculation Loop Startup 3/4.4.2 SAFETY/RELIEF VALVES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems Operational Leakage 3/4.4.4 CHEMISTRY 3/4.4.5 SPECIFIC ACTIVITY 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System Reactor Steam Dome

~PA 3/4 4-1 3/4 4 Ib i

3/4 4-2 3/4 4-3 3/4 4-4 3/4 4-5 3/4 4-6 3/4 4-7 3/4 4-10

~. 3/4 4-13 3/4 4-16 3/4 4-21 3/4.4.7 MAINSTEAM LINE ISOLATION VALVES 3/4.4.8 STRUCTURAL INTEGRITY 3/4.4.9 RESIDUAL HEAT REMOVAL 3/4 4-22 3/4 4-23 Hot Shutdown Cold Shutdown 3 4.

EMER EN Y ORE C LIN SYSTEM 3/4 4-24

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~.., 3/4 5-1 3/4.5.2 ECCS - SHUTDOWN 3/4.5.3 SUPPRE SSION CHAMBER...

3/5 5-6

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~ 3/4 5 8 SUSQUEHANNA - UNIT 1 Vl Amendment No. N.,126

SECTION

6. 1 RESPONSIBILITY PAGE 6-1

6. 2 ORGANIZATION

6. 2. 1 OFFSITE

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6-1 6.2.2 UNIT STAFF..............................

6-1 6.2.3 NUCLEAR SAFETY ASSESSMENT GROUP Function

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L + 1 esponslbllltles.......................................

6-7 Authority..

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6-7 6.2.4 SHIFT TECHNICAL ADVISOR................................

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6. 3 UNIT STAFF UALIFICATIONS 6-7 6. 4 TRAINING..

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6-7 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) unctionl ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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F 6"8 COmpOSltlono...o...o..o.o....o..........'.oo.o

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6-8 lternates........................................

A 6-8 M t J

eetlng Frequency.........................

6-8 uorum.........................................

Q 6-8 Responsibilities Author 1 ty e

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6-10 ecords................................................

R 6-10 6.5.2 SUSQUEHANNA REVIEW COMMITTEE (SRC)

Functl on. ~.............

6-10 C

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6-11 lternates...............................

A 6-11 Consultants..............

Meeting Frequency....

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6-11 6-11 Quorum...

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6-11 SUSQUEHANNA - UNIT 1 Xlx Amendment No. 29

o o

ADMINI TRATIVE NTROLS ggQT~IN UEHANNAREVIEW MMITTEE R

) (Continued)

,'eview Audits Authority Records

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6 1 3 6.5.3 TECHNICALREVIEW AND CONTROL Activities Technical Review

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6 14 REP RTABLE EVENT A TI N

6-14

.7 AFETY LIMITVI LATI N 6.

PROCED RES AND PR GRAM REP RTIN REQ IREMENT 6-15

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6-15 Routine Reports Startup Reports Annual Reports Monthly Operating Reports Annual Radiological Environmental Operating Report Semiannual Radioactive Effluent Release Report Special Reports Core Operating Limits Report 6.1 REC RD RETENTI N

6.11 RADIATIONPR TE TION PRO RAM 6-17 6-17 6-17 6-18 6-18 6-19 6-20 6-20a 6-21 6-22

.12 HIGH RADIATI N AREA

.1 PR E

NTR L PR GRAM 6-22

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.15

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T RADI A TIVE WA TE TREATMENT SY TEM 6-25 SUSQUEHANNA - UNIT 1 XX Amendment No.,

126

INDEX LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS 3/4 1-21 3.1.5-2 3.4.1

~ 1.1-1 THERMALPOWER RESTRICTIONS SODIUM PENTABORATE SOLUTION CONCENTRATION 3/4 1-22 t

3/4 4-1b 3.4.6.1-1 MINIMUMREACTOR VESSEL METALTEMPERATURE VS.

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

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3/4 4-18 B 3/4 3-1 REACTOR VESSEL WATER LEVEL B 3/4.4.6-1 FAST NEUTRON FLUENCE (E) 1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE

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B 3/4 3-8 B 3/4 4-7 5.1.1-1 EXCLUSION AREA 5.1.2-1 LOW POPULATION ZONE 5-2 5-3 5.1.3-1a MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3-1b MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5-4 5-5 6.2.1-1 OFFSITE ORGANIZATION 6.2.2-1 UNIT ORGANIZATION 6-3 6-4 SUSQUEHANNA - UNIT 1 XXI Amendment No. Qg,126

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i. ).2. Ii1 RtACT4a )e0 tCT10II SYSTte lIISTCOenArl4II SET %1IIT5 REACT0R iROTECT10II SYSTEM lIISTRNIEIITATIOII ~........

2og I

3/i 3

2 REACT0R IROTECTIOII SYSTE3I RESPONSE TIICS.....,....

3/i 3 i ACT4a ei0~tCT10II SYSTfII llISTIPg>TAT10II SuiVE? LANCE REOUlREIIEIITS 150LITI0II ACTuar IOII INSTRQCIITATIOII...............

3/i )~7

)/i )i11 I50~TIOII ACtWTIOII lIISTROCNTATIOII 5NVCILLACE QJIRQIEIITS i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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Rf 3/i ) 23 150>TIOII ACTUATI0II 1NSTIQCIITAT10II SETIOIIITS.....

)/i ).0 ISOLATIOII $YSTQI !IISTRLNCNTATIl R55 OIQ TIlg..

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)3)2 MRCQCY CORt COOL,IIC SYSTOI iCTuaTINI IISTRQQIITATlOII 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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ENarQCY CORE COOilll0 SYSTOI ACTVaTIN IIISTIPCIITATl4IISKTNLIITS o ~ ~ ~ ~ ~ ~ oe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~

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i.).3.1 1 3.3.i.l 1 3.).i.l 2 IONEIICY COat COOI.IIC SYSTNI ACTIIATI INTNJgjITAT?Nl SN%luACK RE(gB!%ÃfS.........

1%5 RNCI CVLATINI N80 TRP SYSTQI 5TQNITATINI ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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ASS RNCICVLATI MO TRI~ SYSTDI ISIQCXTATINICPOIIIT5 o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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1. 0 DEFI NIT IONS 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 s

1. I 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 aver-aged exposure of all the fuel rods in the specified bundle divided by the number of fuel rods in the fuel bundle.

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

AVERAGE PLANAR LINEAR HEAT GENERATION RATE

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

CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

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

The CHANNEL CALIBRATION may be performed by any series of sequentiala, overlapping or total channel steps such that the en.i. e channel is calibrated.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavio~

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 chnnnel 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 pirforsed by any series of sequential, overlapping or total channel steps such that the entire channel is t~~t~d SUSQUEHANNA - UNIT 1 Amendtsent No. 72, s:C

DEFINITION ORE ALTERATI N 1.7 CORE ALTERATIONshall 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.

Susp'ension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

RE PERATIN LIMIT REP RT 1.7A The CORE OPERATING LIMITS REPORT is the Susquehanna SES Unit 1 specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.3.

Plant operation within these operating limits is addressed in individual specifications.

RITI AL WER RATI 1.8 The CRITICALPOWER 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.

D E E IVALENTI-1 1

1.9 DOSE EQUIVALENT1-131 shall be that concentration of l-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, l-132, I-133, 1-134, and 1-135 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."

6-AVERA E DISINTE RATI N ENER Y

1.10 6 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.

II EMER EN Y RE 0

LIN SY TEM E

RESPON E TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIMEshall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety functions, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable.

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

END-OF-Y LE RE IR LATION P MP TRIP SYSTEM RESPON E TIME 1.12 The END-OF<YCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

a.

Turbine stop valves, and b.

Turbine control valves.

This total system response time consists of two components, the instrumentation 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 is measured.

SUSQUEHANNA - UNIT 1 1-2 Amendment No. 126

DEFINITIONS FRA Tl N

F LIMITIN P

WER DEN ITY 1.13 The FRACTION OF LIMITINGPOWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the applicable LHGR for APRM Setpoint limit specified in the CORE OPERATING LIMITS REPORT for that bundle type.

FRA TION F RATED THERMAL P WER 1.14 The FRACTION OF RATED THERMALPOWER (FRTP) shall be the measured THERMALPOWER divided by the RATED THERMAL POWER.

FREQ EN Y NOTATION 1.16 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

A E

S RADW T..M NT Y TEM 4lA 1.16 A GASEOUS'O'AQWASTE.-TREATMENT SYSTEM shall be any system designed and installed to reduce radioa'ctiye gaseou7).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 LEAKA f/ e 1.17 IDENTIFIED LEAKAGEshall 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 operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

I LATION YSTEM RE P

N E 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.

LIMITIN NTR L R 0 PATTERN 1.19 A LIMITINGCONTROL ROD 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 GENERATI N RATE 1.20 LINEARHEATGENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod.

It is the integral of the heat flux over the heat transfer area associated with the unit length.

SUSQUEHANNA - UNIT 1 1-3 Amendment No. 7g, <26

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

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

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

MEMBER S

OF THE PUBLIC 1.23 HEMBER(S)

OF THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant.

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

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

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

HINIHUM CRITICAL POWER RATIO 1.24 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core for each class of fuel.

OFFSITE DOSE CALCULATION HANUAL 1.25 The OFFSITE DOSE CALCULATION MANUAL (ODCH) shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints and in the conduct of the environmental radiological monitoring program.

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 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.

SUSQUEHANNA - UNIT 1 1-4

t 3/4.2 P

WER DISTRIBUTION LIMITS 4.2.1 AVERA E PLANAR LINEAR HEAT ENERATI N RATE LIMITINGC NDITI N FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for all fuel shall not exceed the limit specified in the CORE OPERATING LIMITS REPORT.

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

A~TION:

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

SURVEItLANCE REQUIREIVIENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limit:

a.

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

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

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

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 1 3/4 2-1 Amendment No. Dg, 126

P WER DI TRIB TI N LIMIT

/4.2.2 APRM ETP INT LIMITING NDITI N 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 SETPOINT ~

S c (0.58W + 59%) T SR S (0.58W + 50%) T ALLOWABLEVALUE:4""

S c (0.58W + 62%) T SRB c (0.58W + 53%) T where:

S and S>B are in percent of RATED THERMAL POWER, W

=

Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million Ibs/hr, T

=

Lowest value of the ratio of FRACTION OF RATED THERMALPOWER divided by the MAXIMUMFRACTION OF LIMITINGPOWER DENSITY. The FLPD for SNP fuel is the actual LHGR divided by the LINEAR HEAT GENERATION RATE for APRM Setpoints limit specified in the CORE OPERATING LIMITS REPORT.

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 THERMALPOWER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flowbiased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or SRB, as determined above, initiate corrective action within 15 minutes and adjust S and/ or SRB to be consistent with the Trip Setpoint value within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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% of RATED THERMALPOWER, and a notice of the adjustment is posted on the reactor control panel

~

See Specification 3.4.1.1.2.a for single loop operation requirements, SUSQUEHANNA - UNIT 1 3/4 2-2 Amendment,NO.'.Qg, 126

P WER DI TRIB TI N LIMITS LIRVEILLANCEREO IREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flowbiased simulated thermal power-upscale scram and flowbiased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

a.

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

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

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the rector is operating with MFLPD greater than or equal to FRTP.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 1 3/4 2-3 Amendment No. 7g, 126

P WER DI TRIB TI N LIMIT 4.2.

MINIM M RITI AL P WER RATIO LIMITINGCONDITION FOR OPERATION 3.2.3 The MINIMUMCRITICAL POWER RATIO (MCPR) shall be greater than or equal to the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT.

25% RATED THERMAL POWER.

ACTION:

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

SURVEILLANCEREQUIREMENTS 4.2.3.1 MCPR shall be determined to be greater than or equal to the applicable MCPR limit.

a.

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

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

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

d.

The provisions 'of Specification 4.0.4 are not applicable.

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

QIHI, 126

~

~

E,

P WER DI TRIB TI N LIMIT 4.2.4 LINEAR HEAT ENERATI N RATE LIIVIITING NDITI N FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the applicable LHGR limit specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: OPERATIONALCONDITION 1, when THERMALPOWER 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 1 5 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMALPOWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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

a.

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

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMALPOWER increase of at least 15%

of RATED THERMAL POWER, and c.

Initiallyand at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

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

Egg 126

100 E

1 1

I D

W CC O

0 E

Cl I

O, CP'-

80 70 60 60

~~, ~

~

~

~ a

">DP V~

~ '

~

~

~

~

~

~

~

~ > t

~

V

~

~

~

~

~ ~ \\ Ap 0

30 36 40 46 60 56 60, 05~'0 Coro Row (% RATED)

Figure 3.4.4.1.1-4 THERMAL POWER RESTRIGTIONS SUSQUEHANNA - UNIT 1 3/4 4-1b Amendment No. fN 126

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITINGCONDITION FOR OPERATION 3.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed s 80% of the rated pump speed and the reactor at a THERMALPOWER/core flow condition outside of Regions I and II of Figure 3.4.1.1.1-1, 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 0.58W + 54%

AIIowabte,Value 0.58W + 57%

3.

Specification 3.2.2:

the APRM Setpoints shall be as follows:

Trip Setpoint S 6 (0.58W + 54%) T SRa 6 (0.58W + 45%) T

', -'llowable:Value:-

S g (0.58W + 57%) T SR8 5 (0.58W + 48%) T 4.

Specification 3.2.3: The MINIMUMCRITICALPOWER RATIO (MCPR) shall be greater than or equal to the applicable Single Loop Operation MCPR limit as specified in the CORE OPERATING LIMITS REPORT.

5.

Table 3.3.6-2:

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

a.

RBM - Upscale

b. APRM-Flow Biased Trip Setpoint 6 0.66W + 36%

Trip Setpoint:....

C Q58W+ 45%

Allowable Value 6 0.66W + 39%

Allowable Value s Q58W+ 48%

APPLICABILITY: OPERATIONAL CONDITIONS 1" and 2" +, except during two loop operation.¹ SUSQUEHANNA - UNIT 1 3/4 4-1 c Amendment No.""~~ 126

I 1

~

~

3/a. 1 RKACTIVITY CONTROL SYSTKHS bASKS a 3/i 1 1 Ieeehant No. 72 gg 9 $87 3/a. l.I SHUTDOWN NARGIN A sufficient SHUTDOWN HARGIN ensures that l) the reactor can be wde sut-critical free all operating conditions,

2) the reactivity transients associated vith postulated accident conditions are controllable vithin acceptable

limits, and 3) the reactor vill be maintained sufficiently subcritical to preclude inad-vertent criticality in the shut4ovn condition.

Since core reactivity values vill vary through core life as a function of fuel 4epletion and poison burnup, the deaonstration of SHUTDOWN INRGIN vill be perforeed in the cold, xenon. free condition an4 shall ster the core to be sub-critical by at least R + 0.3 de>ta k/k or R + 0.28K delta k/k, as appro-priate.

The value of R in units of X delta k/k is ttw difference betveen the cal-culated beginning of cycle shutdovn aergin ainus the calculated

~inimm shutdown aargin in the cycle, vhere shutdovn aargin is a positive nuber.

The value of R oust be positive or zero and aust be deterained for each fuel loading cycle.

Tvo different values are supplied in the Liaiting Condition for Operation to provide for the different aethods of deoanstration of the SHUTDOWN NRCIN.

The highest vorth rod ax~. be deteroined analytically or by test.

The SHUTDOWN NRCIN is demonstrated by cont~ol rod vith4raval at the beginning of life fuel cycle conditions, and, if necessary, at any future tice in the cycle if the first deoonstration indicates that the required aargin could be reduced as a function of exposure.

Observation of subcriticality in this condition assures subcritica-lity vith the aost reactive control ro4 fully vithdravn.

This reactivity characteristic has been a basic ass~tion in the analysis of plant perforeance and can be best deoonstrated at the tice of fuel loading, but the sargin aust also be detemined anytiae

~ control rod is incapable of insertion.

3/1.1.2 Reactivit Anoaalies Since the SHUTDOWN MARGIN ~uirment is small, a careful check on actual reacto~ conditions ~ared to the predicted con4itions

$ s necessary.

Any changes in reactivity free that of the predicted (predicted core keff) can be detareined f~ the core aeitoring systeo (mnitored core keff).

In the absence

~f any 4eviation

$n plant-operating conditions or reactivity aeaaly, these values should be essentially apel since the calculational eethodologies are consistent.

The predicted core k ff is calculated by a 3D core siwlation code as a function eff of cyc1e exposure.

This is performed for pro]ected or anticipated reactor operat-ing states/conditions throughout the cycle an4 is usually done prior to cycle

~perat$ on.

Ttw aeitored core keff is the keff as calculated by the core anitor

$ ng systea, for actual plant conditions.

Sinco the ~arisons are easily 4one, frequent checks are mt an iaosition on normal operation.

A 1% deviation in reactivity free that of Qe predicted is lamer than expected for normal operation, and therefore shou14 be throughly

~vaTuatad.

A deviation as larie as 1X vould not exceed the design conditions of the reactor.

SUQVKHNNA NIT 1

REACTIVITY NTR L

Y TENI BASES

/4.1.3 NTR L R D

The specification of this section ensure that (1) the minimum SHUTDOWN MARGINis maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) limitthe potential effects of the rod drop accident.

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

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

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

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

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

The number of control rods permitted to be inoperable could be more than the eight allowed by the 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.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the limit specified in Specification 2.1.2 during the core wide transient analyzed for the specific reload cycle.

The MCPR operating limits as specified in the CORE OPERATING LIMITS REPORT may be a function of average scram speed.

In such a case, the results of the required scram time testing (Specification 4.1.3.3) are used to adjust the MCPR operating limits to assure the validity of the cycle specific transient analyses.

This ultimately assures that MCPR remains greater than the limitspecified in Specification 2.1.2. The occurrence of scram times longer than those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

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

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

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

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

SUSQUEHANNA - UNIT 1 B 3/4 1-2 Amendment Nd. QS,126

REACTIVITY CONTROL SYSTEMS BASES CONTROL ROOS (Continued)

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 determining 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 backup to the initial demonstration.

In order to ensure that the control rod patterns can be followed and there-fore that other Parameters are within their limits, the control rod position indication system must be OPERASLE.

The control rod housing support restricts the outward movement of a con-tro1 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 amount 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 ROD 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 homo-

geneous, 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/~.

Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20K of RATED THERMAL POWER provides adequate control.

The RSCS and RWM logic automatically initiates at the low power setpoint (20% of RATED THERMAL POWER) to 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 range of key reactor parameters (which envelope the operating ranges of these variables),

the fuel enthalpy rise during a postulated control rod drop acci-dent remains considerably 10wer than the 280 cal/gia limit.

For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient, effactive delayed neutron fraction, and maximum four'bundle local peaking factor are compared with the inputs to the parametric analyses to deter-mine the peak fual rod enthalpy rise.

This value is then compared against the SUSgUEHANN - NIT 1 B 3/4 1-3, Asendsent No. 109

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REA TIVIT NTR L

YSTEIVI BASES 4.1.4 NTR L R D PRO RAM NTROL (Continued) 280 cal/gm design limitto demonstrate compliance for each operating cycle. Ifcycle-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 referenced in Specification 6.9.3.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels willblock erroneous rod withdrawal soon enough to prevent fuel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

4.1.5 TANDBYLIQUID C NTROL YSTEM The standby liquid control system provides a backup capability for bringing the reactor from full pow'er to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted.

To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the'reactor core in approximately 90 to 120 minutes.

A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown requirement.

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 and the required pumping rate is 41.2 gpm. The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the fillingof other piping systems connected to the reactor vessel.

The temperature requirement for the sodium pentaborate solution is necessary to ensure that the sodium pentaborate remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

Surveillance requirements are established on a frequency that assures a high reliability of the system.

Once the solution is established, boron concentration willnot vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals willassure that these valves will not fail because of deterioration of the charges.

SUSQUEHANNA - UNIT 1 B 3/4 1-4 Anendment No.'gs; 126

4.2 P

WER Dl TRIB Tl N LIIVIIT BASES 4.2.1 AVERA E LANAR LINEA HEAT ENERATI N RA This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limitspecified 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.

The Technical Specification APLHGR for SNP fuel is specified to assure the PCT following a postulated LOCA willnot exceed the 2200'F limit.The limiting value for APLHGR is specified in the CORE OPERATING LIMITS REPORT.

The calculational procedure used to establish the APLHGR specified in the CORE OPERATING LIMITSREPORT is based on a loss-of-coolant accident analysis.

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

These'odels are part of the approved methodology referenced in Specification 6.9.3.

4.2.2 APRM SETPOINTS The flow biased simulated thermal power-upscale scram setting and flow biased simulated thermal power-upscale control rod block functions of the APRM 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 a1% plastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (AOO), including transients initiated from partial power operation.

For SNP 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 the LHGR for APRM Setpoints Curve specified in the CORE OPERATING LIMITS REPORT.

The LHGR for APRM Setpoints Curve

'pecified in the CORE OPERATING LIMITSREPORT is based on SNP's Protection Against Fuel Failure (PAFF) line which was developed using the approved methodology referenced in Specification 6.9.3. The LHGR for APRM Setpoints Curve specified in the CORE OPERATING LIMITSREPORT corresponds to the ratio of PAFF/1.2 under which cladding and fuel integrity is protected during AOOs.

SUSQUEHANNA - UNIT 1 8.3/4 2-1 Amendment N5. Egg 126

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WER D TRIB Tl N LIMIT BASES 3/4.2.

MINIM M RITICALP WER RATIO The required operating limit MCPRs at steady state operating conditions as specified in the CORE OPERATING LIMITS REPORT are derived from the established fuel cladding integrity Safety Limit MCPR and analyses of abnormal operational transients.

For any abnormal operational transient analysis with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip settings given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limitingtransients have been analyzed to determine which result in the largest reduction in CRITICALPOWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCPR. When added to the Safety LimitMCPR, the required minimum operating limitMCPR specified in the CORE OPERATING LIMITS REPORT is obtained.

The required MCPR operating limits as functions core power, core flow, and plant equipment availability condition are specified in the CORE OPERATING LIMITS REPORT.

The cycle specific transient analyses to determine the MCPR operating limits were performed using the NRC approved methods referenced in Specification 6.9.3.

The MCPR operating limits as specified in the CORE OPERATING LIMITS REPORT may be specified as a function of average scram speed.

In such a case, the results of the required scram time testing (Specification 4.1.3.3) are used to adjust the MCPR operating limits to assure the validity of the cycle specific transient analyses.

This ultimately assures that MCPR remains greater than the limit specified in Specification 2.1.2 for all anticipated operational occurrences.

The CORE OPERATING LIMITS REPORT specifies core flow dependent MCPR operating limits which assure that the Safety Limit MCPR will not be exceeded during a flow increase transient resulting from a motor-generator speed control failure.

'The flow dependent MCPR is only calculated for the manual flow control mode.

Therefore automatic flow control operation is not permitted.

The CORE'OPERATING LIMITS REPORT specifies the power dependent IVICPR operating limits which assure that the Safety Limit MCPR will not be exceeded in the event of a Feedwater Controller Failure, Rod Withdrawal Error, or Load Reject without Main Turbine Bypass Operable initiated from a full power or reduced power condition.

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

Additional analyses are performed to determine the MCPR operating limit with either the Main Turbine Bypass inoperable or the EOC-RPT inoperable.

Analyses to determine thermal margin with both the EOC-RPT inoperable and Main Turbine Bypass inoperable have not been performed.

Therefore, operation in this condition is not permitted.

At THERMAL POWER levels less than or equal to 25% of RATED THERMALPOWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content willbe very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.

During initial start-up testing of the plant, a MCPR evaluation SUSQUEHANNA - UNIT 1 B 3/4 2-2 Anendment No. Xgg 126

3 4.4 REACTOR COOLANT SYSTEM BASES 3 4.4.1 RECIRCULATIONSYSTEM 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.

LOCA analyses for two loop operating conditions, which result in Peak Cladding Temperatures (PCTs) below 2200 F, bound single loop operating conditions.

Single loop operation LOCA analyses using two-loop MAPLHGRlimits result in lower PCTs. Therefore, the use of two-loop MAPLHGR limits during single loop operation assures that the PCT during a LOCA event remains below 2200 F.

The MINIMUMCRITICALPOWER RATIO (MCPR) limits for single loop operation assure that the Safety Limit MCPR is not exceeded for any Anticipated, Operational Occurrence (AOO).

In addition, the MCPR limits for single-loop operation protect against the effects of the Recirculation Pump Seizure Accident. That is, for operation in single-loop with an operating MCPR limit) 1.30, the radiological consequences of a pump seizure accident from single-loop operating conditions are but a small fraction of 10 CFR 100 guidelines.

For single loop operation, the RBM and APRM setpoints are adjusted by a 8.5% 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 on THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel nozzles, recirculation pumps and the vessel bottom head during extended operation in the single loop mode. The threshold limits are those values which will sweep up the cold water from the vessel bottom head.

Specifications have been provided to prevent, detect, and mitigate core thermal hydraulic instability events. These specifications are prescribed in accordance with NRC Bulletin 88-07, Supplement 1, "Power Oscillations in Boiling Water Reactors (BWRs)," dated December 30, 1 988.

I LPRM upscale alarms are required to detect reactor core thermal hydraulic instability events.

The criteria for determining which LPRM upscale alarms are required is based on assignment of these alarms to designated core zones.

These core zones consist of the level A, B and C alarms in 4 or 5 adjacent LPRM strings.

The number and location of LPRM strings in each zone assure that with 50/o or more of the associated LPRM upscale alarms OPERABLE sufficient monitoring capability is available to detect core wide and regional oscillations.

Operating plant instability data is used to determine the specific LPRM strings assigned to each zone.

The core zones and required LPRM upscale alarms in each zone are specified in appropriate procedures.

SUSQUEHANNA - UNIT 1 B,3/4 4-1 Amendment No. TIN~

126

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4. 4. 1 RECIRCULATION SYSTEM (Continued)

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

recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of refloodfng 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.

In order to prevent undue stress on the vessel nozzles and bottom head

region, the recirculation loop temperatures shall be within 504F of each other prior to startup of an idle loop.

The loop temperature must also be within 504F of the reactor pr essure vessel coolant temperature to prevent thecal shock to the recirculation pump and-recirculation nozzles.

Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the tem-perature difference was greater than 145'F.

SU5gUfHANNA - UNIT 1 B 3/4 4-la mndwent No. >O>

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>OR MD?OACTlVK GASEOUS AND LlOUl0 EFFLUKNTS SVSQJBQtlNA NIT J

~~geht No. 29

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 764 fuel assemblies.

Each assembly consists of a matrix of Zircaloy clad fuel rods with an initial composition of non-enriched or slightly enriched uranium dioxide as fuel material and water rods.

Limited substitutions of Zirconium alloy filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved'codes and methods, and shown by test or analyses to comply with all fuel safety design bases.

A limited number of lead use assemblies that have not completed representative testing may be placed in non-limiting core regions.

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

Reload fuel shall have a maximum average enrichment of 4.0 weight percent U-235.

CONTROL ROD ASSEIVIBLIES 5.3.2 The reacto'r core shall contain 185 cruciform shaped control rod assemblies.

The control material shall be boron carbide powder tB4C), and/or Hafnium metal.

The control rod shall have a nominal axial absorber length of 143 inches.

Control rod assemblies shall be limited to those control rod designs approved by the NRC for use in BWRS.

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 withthe 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

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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>>, of 528 F.

SUSQUEHANNA - UNIT 1 5-6 Amendment No. flS, 126

ADMINI TRATIVE NTR L

RE PE I

LIMIT REP RT 6.9.3 The CORE OPERATING LIMITSREPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

6.9.3.1, Core operating limits shall be established prior to the startup of each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

a, The Average Planar Linear Heat Generation Rate (APLHGR) for Specification 3.2.1.

b.

The Linear Heat Generation Rate for Average Power Range Monitor (APRM) Setpoints for Specification 3.2.2.

c.

The Minimum Critical Power Ratio (MCPR) for Specification 3.2.3 and 3.4.1.1.2.

d.

The Linear Heat Generation Rate (LHGR) for Specification 3.2.4.

e.

The Thermal Power Restrictions for Specification 3.4.1.1.1 and 3.4.1.1.2.

And shall be documented in the CORE OPERATING LIMITS REPORT.

6.9.3.2 The analytical methods used to determine the core operating limits shall be those topical reports and those revisions and/or supplements of the topical report previously reviewed and approved by the NRC, which describe the methodology applicable to the current cycle.

For Susquehanna SES the topical reports are:

1

~

PL-NF-87-001-A, "Qualification of Steady State Core Physics Methods for BWR Design and Analysis," July, 1988.

2.

PL-NF-89-005-A, "Qualification of Transient Analysis Methods for BWR Design and Analysis," July, 1992.

3.

PL-NF-90-001-A, "Application of Reactor Analysis Methods for BWR Design and Analysis," July, 1992.

4.

XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads,"

Exxon Nuclear Company, Inc., June 1986.

5.

XN-NF-85-67(P)(A), Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, Inc., September 1986.

6.

PLA-3407, "Proposed Amendment 132 to License No. NPF-14: Unit 1

Cycle 6 Reload," Letter from H. W. Keiser (PPSL) to W. R. Butler (NRC), July 2, 1990.

7.

Letter from Elinor G. Adensam (NRC) to H. W. Keiser (PP&L), "Issuance of Amendment No. 31 to Facility Operating License No.,NPF Susquehanna Steam Electric Station, Unit 2," October 3, 1986.

8.

PLA-3533, Revised Proposed Amendment 67 to License No. NPF-22: Unit 2 Cycle 5 Reload," Letter from H. W. Keiser (PPSL) to W. R. Butler (NRC), March 7, 1991.

SUSQUEHANNA - UNIT 1

. 6-20a Amendment No.

126

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ADMINISTRATIVECONTROLS RE PERATIN LIMITS REPORT (Continued) 9.

XN-NF-84-97, Revision 0, "LOCA-Seismic Structural Response of an ENC 9x9 Jet Pump Fuel Assembly," Exxon Nuclear Company, Inc., December 1984.

10.

PLA-2728, "Response to NRC Question: Seismic/LOCA Analysis of U2C2 Reload,"

Letter from H. W. Keiser (PPSL) to E. Adensam (NRC), September 25, 1986.

11.

XN-NF-82-06(P)(A), Supplement 1, Revision 2, "Qualification of Exxon Nuclear Fuel for Extended Burnup Supplement 1 Extended Burnup Qualification of ENC 9x9 Fuel,"

May 1988.

12.

XN-NF-80-19(A), Volume 1, and Volume 1 Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors: Neutronic Methods for Design and Analysis,"

Exxon Nuclear Company, Inc., March 1983.

13.

XN-NF-524(A), Revision 1, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors,"

Exxon Nuclear Company, Inc., November 1983.

14.

XN-NF-512-P-A, Revision 1 and Supplement 1, Revision 1, "XN-3 Critical Power Correlation," October, 1982.

15.

XN-NF-80-19(A), Volumes 2, 2A, 2B, and 2C, "Exxon Nuclear Methodology for Boiling Water Reactors:

EXEM BWR ECCS Evaluation Model,"

Exxon Nuclear Company, Inc., September 1982.

16.

XN-NF-CC-33(A), Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon Nuclear Company, Inc., November 1 975.

17.

XN-NF-82-07(A), Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., November 1982.

18.

XN-NF-84-117(P), "Generic LOCA-Break Spectrum Analysis: BWR 3 and 4 with Modified Low Pressure Coolant Injection Logic," Exxon Nuclear Company, Inc.,

December 1984.

6.9.3.3 19.

XN-NF-86-65, "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for 9x9 Fuel,"

Exxon Nuclear Company, Inc., May 1986.

I'he core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

SUSQUEHANNA - UNIT 1

.6-20b Amendment No, >26

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA POWER EE LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-388 SUS UEHAN A STEAM ELECTRIC STATION UNIT 2 AMENDMEN TO FACI ITY OPERATING LICENSE Amendment No. 95 License No. NPF-22 1.

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

A.

The application for the amendment filed by the Pennsylvania Power Il Li'ght Company, dated December 18,

1992, as supplemented by telecopy dated January '28,

1993, and by letters dated March 25, and May 20,

1993, 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 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated 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 S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

95 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

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

3.

This license amendment is effective as of its date of issuance and is to be implemented within 30 days after its date of issuance.

FOR THE NUCLEAR R ULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 4, 1993 ichael L.

Bo A

ing Director Project Direc rate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO.

95 FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

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

The overleaf pages are provided to maintain document completeness.*

REMOVE INSERT 111lv V

Vl Xlx XX Xxl Xxll 1-1 1-2 1-3 1-4 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4

'/4 2-5 3/4 2-5a 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-Sa 3/4 2-Sb 3/4 2-9 1

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XX XXl*

Xxl 1 1*

1-2 1-3 4*

3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-5

UNIT 2 3/4 2-10 3/4 2-10a 3/4 4-lb 3/4 4-lc 8 3/4 1-1 8 3/4 1-2 8 3/4 1-3 8 3/4 1-4 8 3/4 2-1 8 3/4 2-2 5-5 5-6 6-19 6-20 3/4 4-lb*

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

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

8 3/4 1-4 8 3/4 2-1 8 3/4 2-2 5-5*

5-6 6-19*

6-20 6-20a 6-20b

INDEX DEFINITIONS

/ECTION 1.

OEFINITI N 1.1 ACTION....................... ~........ ~............

1-1 1.2 AVERAGE EXPOSURE.... ~............ ~...................

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE

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1.4 CHANNEL CALIBRATION... ~.......

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

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

1.7 CORE ALTERATION... ~..... ~....................

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1.7A CORE OPERATING LIMITS REPORT........ ~.......

1.8 CRITICAL POWER RATIO...... ~............

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1 2 1.9 DOSE EQUIVALENT I-131 1.10 E-AVERAGE DISINTEGRATION ENERGY

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1.1 1 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.13 FRACTION OF LIMITING POWER DENSITY

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1 2 1-3 1.14 FRACTION OF RATED THERMAL POWER 1.15 FREQUENCY NOTATION

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1.16 GASEOUS RADWASTE TREATMENT SYSTEM 1.1 7 IDENTIFIED LEAKAGE

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1 3 1.1 8 ISOLATION SYSTEM RESPONSE TIME............. ~...

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1.20 LINEAR HEAT GENERATION RATE..............................

1-3 1.21.

LOGIC SYSTEM FUNCTIONAL TEST 1.22 MAXIMUMFRACTION OF LIMITINGPOWER DENSITY 1.23 MEMBER(S) OF THE PUBLIC 1.24 MINIMUMCRITICAL POWER RATIO 1.25 OFFSITE DOSE CALCULATIONMANUAL 1-4 1-4

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SUSQUEHANNA - UNIT 2 Amendment No. B 95

DEFINITIONS INDEX SECTION DEFINITIONS (Continued) 1.26 OPERABLE " OPERABILITY..................

1.27 OPERATIONAL CONDITION - CONDITION..........................

1.28 PHYSICS TESTS...........................

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

1. 30 PRIMARY CONTAINMENT INTEGRITY....

1.31 PROCESS CONTROL PROGRAM.....

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

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

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PAGE 1-4 1-4 1-5 1-5 1-5 1-5 1-5 1"6 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME....................

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

1.36 ROD DENSITY...................................

1"6 1-6

1. 37 SECONDARY CONTAINMENT INTEGRITY............................

1-6

1. 38 SHUTDOWN MARGIN.....

1.39 SITE BOUNDARY.................

1.40 SOLIDIFICATION.......................

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

1"7 1-7 1-7 1-7 1.42 STAGGEREO TEST BASIS.......

1"7 1.43 THERMAL POWER..............

1.44.'TURBINE BYPASS SYSTEM RESPONSE TIME..

1.45 UNIDENTIFIED LEAKAGE.............

1.46 UNRESTRICTED AREA.........................

1.47 VENTILATION EXHAUST TREATMENT SYSTEM.

1.48 VENTING................

1"7 1-7 1-7 1-8 1-8 1-8 SUSQUEHANNA - UNIT 2

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AND LllHITIN AFETY Y TEM ETTIN 1

AF THERMAL POWER, Low Pressure or Low Flow THERMAL POWER, High Pressure and High Flow Reactor Coolant System Pressure Reactor Vessel Water Level........ ~......

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IN TM IN Reactor Protection System Instrumentation Setpoints

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F MIT THERMAL POWER, Low Pressure or Low Flow THERMAL POWER, High Pressure and High Flow Reactor Coolant System Pressure Reactor Vessel Water Level................

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YT IN Reactor Protection System Instrumentation Setpoints...............

8 2-4 SUSQUEHANNA - UNIT 2 Amendment No.91 OtlT 3 3 t99P

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INDEX LIMITIN CONDITI NS FOR OPERATION AND SURVEILLANCE REQUIREMENT

!~ETIAM 4.0 APPLI ABILITY 4.1 REA TIVITY NTR L

Y TEM

~PA

. 3/4 0-1 3/4.1.1 SHUTDOWN MARGIN 3/4.1.2 REACTIVITYANOMALIES 3/4.1.3 CONTROL RODS

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3/4 1-1

... 3/4 1-2 Control Rod Operability Control Rod Maximum Scram Insertion Tines Control Rod Average Scram Insertion Times Four Control Rod Group Scram Insertion Times Control Rod Scram Accumulators Control Rod Drive Coupling Control Rod Position Indication

. 3/4 1-3

. 3/4 1-6

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. 3/4 1-11 3/4 1-13 Control Rod Drive Housing Support

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3/4.1.4 CONTROL ROD PROGRAM CONTROLS

.. 3/4 1-15 Rod Worth Minimizer Rod Sequence Control System Rod Block Monitor

.. 3/4 1-16

. 3/4 1-17

~ 3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM 3 4.2 P

WER DI TRIBUTI N LIMIT 3/4 1-19 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 3/4 2-1 3/4.2.2 APRM SETPOINTS

. 3/4 2-2 3/4.2.3 MINIMUMCRITICAL POWER RATIO 3/4.2.4 LINEAR HEAT GENERATION RATE 3/4 2-4

.. 3/4 2-5 SUSQUEHANNA - UNIT 2 IV Amendment No.

$8, 95

INDEX LIHITING CONDITIONS FOR. OPERATION AND SURVEILLANCE RE UIREHENTS SECTION 3/4. 3 INSTRUMENTATION PAGE 3/4. 3. 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............

3/4 3-1 3/4. 3. 2 ISOLATION ACTUATION INSTRUMENTATION.......

3/4 3-9 3/4. 3. 3 EMERGENCY CORE COOLING SYSTEH ACTUATION INSTRUHENTATION..

3/4 3-27 3/4. 3. 4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation..

End-of-Cycle Recirculation Pump Trip System Instrumentation..

3/4 3"36 3/4 3-40 3/4. 3. 5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.....

3/4 3-46 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION....................

3/4 3-51 3/4. 3. 7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.....,...........

Seismic Monitoring Instrumentation...................

Meteorological Monitoring Instrumentation...

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Remote Shutdown Monitoring Instrumentation...........

Accident Honitoring Instrumentation..................

Source Range Monitors........

Traversing In-Core Probe System......................

Chlorine Detection System....

Fire Detection Instrumentation..

Radioactive Liquid Effluent Monitoring Instrwentation......................................

Radioactive Gaseous Effluent Monitoring Instroaentation.............

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3/4 3-57 3/4 3-61 3/4 3-64 3/4 3"67 3/4 3-70 3/4 3-74 3/4 3-75 3/4 3-76 3/4 3-77 3/4 3-82 3/4 3-87 3/4 3-94 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEH...................

3/4 3-95 3/4. 3. 9 FEEDMATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUHENTATION.............

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I3/4 3-96 SUS(UEHANNA - UNIT 2

INDEX LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS

!~ET~IN 4.4 R A T R

LANT Y TEM

~PA 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops - Two Loop Operation Recirculation Loops - Single Loop Operation Jet Pumps Recirculation Pumps Idle Recirculation Loop Startup 3/4.4.2 SAFETY/RELIEF VALVES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4 4-1 3/4 4-1b 3/4 4-2 3/4 4-3 3/4 4-4 3/4 4-5 Leakage Detection Systems Operational Leakage 3/4.4.4 CHEMISTRY 3/4.4.5 SPECIFIC ACTIVITY 3/4.4.6 PRESSURE/TEMPERATURE LIMITS 3/4 4-6 3/4 4-7

~.. ~... 3/44-10 3/4 4-13 Reactor Coolant System Reactor Steam Dome 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES 3/4.4.8 STRUCTURAL INTEGRITY 3/4.4.9 RESIDUAL HEAT REMOVAL 3/4 4-16 3/4 4-20 3/4 4-21

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3/4 4-22 Hot Shutdown Cold Shutdown 3/4 4-23

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. 3/44-24 4.5 EMER EN Y RE LIN Y TEM 3/4.5.1 ECCS - OPERATING 3/4.5.2 ECCS - SHUTDOWN 3/4.5.3 SUPPRESSION CHAMBER

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3/5 5-7 3/4 5-9 SUSQUEHANNA - UNIT 2 Vl Amendment No.

)III, 95

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ADMINISTRATIVE CONTROLS SECTION

6. 1 RESPONSIBILITY

6. 2 ORGANIZATION INDEX PAGE 6"1 6.2. 1 OFFSITE......................................

6.2.2 UNIT STAFF.....................................

6.2.3 NUCLEAR SAFETY ASSESSMENT GROUP Function.

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Composition.

Responsibilities Authority 6.2.4 SHIFT TECHNICAL ADVISOR.

6. 3 UNIT STAFF UALIFICATIONS

6. 4 TRAINING.

G. 5 REVIEW AND AUDIT 6-1 6-1 6-7 6-7 6"7 6"7 6-7 6-7 6-7 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC; Function.

Composition.

6-8 6-8 Alternates Meeting Frequency

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ResponsibiTities...

Authority.......

Records SUSQUEHANNA REVIEW Function.

COMMITTEE (SRC) 6-8

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Alternates.....

Consultants.

Meeting Frequency..............

Quorum.

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6eeg 1 6-11 6-11 6-11 SUSQUEHANNA - UNIT 2 Xix

INDEX ADMINI TRATIVE ONTROL

~E~TIQN EHANNA REVIEW MMITTEE R

) (Continued)

Review Audits Authority Records

~PA 6-11 6-12 6-13 6-13 6.5.3 TECHNICAL REVIEW AND CONTROL Activities Technical Review REP RTABLE EVENT A TI N

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.7 AFETY LIMITVl LATI N PR ED RE AND PR RAM

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6 1 5 6-15 REPORTIN RE IREMENTS Routine Reports Startup Reports Annual Reports Monthly Operating Reports Annual Radiological Environmental Operating Report Semiannual Radioactive Effluent Release Report Special Reports Core Operating Limits Report

.10 RE RD RETENTI N

6-17 6-17 6-17 6-18 6-18 6-19 6-20 6-20 6-20b

.11 RADIATI N PROTE Tl N PROGRAM

.12 HIGH RADIATI N AREA 6-22 6-22 SUSQUEHANNA - UNIT 2 XX Amendment No.

95

AO~INISTRATIVE CONTROLS

6. 13 PROCESS CONTROL PROGRAM.

g g3

g. I4 OFS'SITE OOSE CALCULATION NAGUAL.......................

5 Rl 1

$ 5

~D'OR CHANGES TO RAOIOACTIVE 'WASTE TREATMENT SYSTEHS.....

6-g4 SUgUKHANNA UNIT 2

LIST OF FIGURES FIGURE INDEX PAGE 3.1.5-1 3.1.5-2 SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS SODIUM PENTABORATE SOLUTION CONCENTRATION 3/4 1-21 3/4 1-22 3.4.1.1.1-1 THERMALPOWER RESTRICTIONS 3/4 4-1b 3.4.6.1-1 MINIMUMREACTOR VESSEL METALTEMPERATURE VS.

REACTOR VESSEL PRESSURE 3/4 4-18 4.7.4-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONALTEST...............

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

. B3/43-8 B 3/4.4.6-1 FAST NEUTRON FLUENCE (E) 1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE

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/XII Amendment No.Q, >E

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1. 0 DEFINITIONS ie following terms are defined so that uniform interpretation of these

.pecifications may be achieved.

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

I 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 divided by the number of fuel rods in the fuel bundle.

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

'lANNEL CALIBRATION L.4 A CHANNEL CALIBRATION shall be the adjdstment, 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 tota'l channel steps such that the entire channel is calibrated.

CHANNEL CHECK I

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.

SUSQUEHANNA - UNIT 2 Amendment No. 31 HOV n p )g.

DEFINITIONS RE ALTERATI N 1.7 CORE ALTERATIONshall 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 completion of the movement of a component to a safe conservative position.

RE PERATIN LIMIT REPORT 1.7A The CORE OPERATING LIMITS REPORT is the Susquehanna SES Unit 2 specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.3.

Plant operation within these operating limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.8 The CRITICALPOWER 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.

DO E EQUIVALENT I-1 1

1.9 DOSE EQUIVALENTI-131 shall be that concentration of 1-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of l-131, l-132, I-133, I-134, and 1-135 actually present.

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

6-AVERAGE DISINTE RATION 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 I 5 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMER EN Y RE LIN Y TEM E

RE P

N E TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIMEshall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety functions, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable.

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

END-OF-CY LE RE IR LATI N P MP TRIP YSTEM RESP N

E TIME 1.12 'he END-OF<YCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

a.

Turbine stop valves, and b.

Turbine control valves.

This total system response time consists of two components, the instrumentation 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.

SUSQUEHANNA - UNIT 2 1-2 Amendment iR. 95

DEFINITION FRA TI N

F LIMITIN P

WER DEN ITY 1.13 The FRACTION OF LIMITINGPOWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the applicable LHGR for APRM Setpoint limit specified in the CORE OPERATING LIMITS REPORT for that bundle type.

FRACTI N

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

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

ASEO RADWA T

Y TEM e>=-.~ 4g 1,16 A GASEOUS gAtIM/ASTE-TJQA'TMENT:STSTEM shall be any system designed and installed tn reduce radioactfve-.gaseous'off luents by collecting primary coolant system offgases from the primary system'.'and" providiiIg for delay or holdup for the purpose of reducing the total radioactivity prior~to releaseqo,the environment.

IDENTIFIED LEAKA saT e.

-'.17 IDENTIFIED LEAKAGEshall 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 operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

I OLATI N Y TEMRE P

N ETIM 1.18 The ISOLATIONSYSTEM RESPONSE TIMEshall be t)1at 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 rrleasured.

LIMITIN ONTR L R D PATTERN 1.19 A LIMITINGCONTROL ROD 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 LINEARHEAT 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 Ho. Q, 95

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

The LOGIC SYSTEH 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 LIMITING POWER DENSITY (MFLPD) shall be the highest value of the FLPD which exists in the core.

MEMBER S OF THE PUBLIC 1.23 MEHBER(S)

OF THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant.

This category does not include employees of the utility, its contractors or vendors'lso excluded from this category are persons who enter the site to service equipment or to make deliveries.

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

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 liqpid effluent monitoring alarmltrip setpoints and in the conduct of the environmental radiological monitoring program.

v, L

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 that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

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

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

SUS(UEHANNA - UNIT 2 1-4

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WER DI TRIB TI N LIMIT

/4.2.1 AVERA E PLANAR LINEAR HEAT GENERATION RATE LIMITINGCONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for all fuel shall not exceed the limit specified in the CORE OPERATING LIMITS REPORT.

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

ACTION:

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

SURVEILLANCEREQUIREIVIENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limit:

a.

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

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

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

d.

The provisions of Specification 4.0.4 are not applicable.

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

$5, 95

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P WER DI TRIB TI N LIMIT 4.2.2 APRIVI ETP INT LIMITING ONDITI N 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'SETPOINT,>>,

S s (0.58W + 59%) T SRB C (0.58W + 50%) T

. ALLOWABLE'ALUE/)"

S M (0.58W + 62%) T SRB c (0.58W + 53%) T where:

S and SRB are in percent of RATED THERMALPOWER, V

W Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million Ibs/hr, T

=

Lowest value of the ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the MAXIMUMFRACTION OF LIMITING POWER DENSITY.

The FRACTION OF LIMITINGPOWER DENSITY (FLPD) for SNP fuel is the actual LHGR divided by the LINEAR HEAT GENERATION RATE for APRM Setpoints limit specified in the CORE OPERATING LIMITS REPORT.

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 THERMALPOWER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flowbiased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or SRB, as determined above, initiate corrective action within 15 minutes and adjust S and/ or SRB to be consistent with the Trip Setpoint value within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />'or reduce

'HERMALPOWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCEREQUIREMENTS 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 flowbiased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

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% of RATED THERMALPOWER, and a notice of the adjustment is posted on the reactor control panel.

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

SUSQUEHANNA - UNIT 2 3/4 2-2 Amendment No. 91, 95

P WER Dl TRIB TI N LIMIT

/4.2.2 APRM ETP INT LIMITINGCONDITI N FOR OPERATION 4.2.2 (Continued) a.

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

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

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the rector is operating with MFLPD greater than or equal to FRTP.

d.

The provisions of Specification 4.0.4 are not applicable.

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

$8, 9>

P WER DI TRIB TI N LIMIT 4.2.3 MINIM M RITI AL P WER RATIO LIMITINGCONDITION FOR OPERATION 3.2.3, The MINIMUMCRITICAL POWER RATIO (MCPR) shall be greater than or equal to the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT.

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

SURVEILLANCEREQUIREMENTS 4.2.3.1 MCPR shall be determined to be greater than or equal to the applicable MCPR limit:

a.

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

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

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

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 2-4 Amerxlment No. 9S, 95

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POWER Dl TRIB Tl N LIMIT 4.2.4 LINE R HEAT ENERATI N RATE LIIVIITINGCONDITI N FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the applicable LHGR limitspecified in the CORE OPERATING LIMITS REPORT.

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equal to 25% of RATED THERMAL POWER.

~ATION:

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

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

a.

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

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMALPOWER increase of at least 15/

of RATED THERMAL POWER, and c.

Initiallyand at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 2-5 Amendment No. 97, 95

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Figure SA.4.1.1-1 THERMAl. POWER RESTRICTIONS 3/4 4-1b Anerxhnent No. 91 QGT28 $92

REACTOR COOLANT SYSTEIVI RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITINGCONDITION FOR OPERATION r

3.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed 6 80%

of the rated pump speed and the reactor at a THERMAL POWER/core flow condition outside of Regions I and II of Figure 3.4.1.1.1-1, 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 0.58W + 54%

llowable Value;-.

s 0.58W + 57%

3.

Specification 3.2.2:

the APRM Setpoints shall be as follows:

Trip Setpoint S s (0.58W + 54%) T SRB 6 (0.58W + 45%) T

'.-,. 'Allowable:ya'Iue S 6 (0.58W + 57%) T Sas 6 (0.58W + 48%) T 4.

Specification 3.2.3: The MINIMUMCRITICAL POWER RATIO (MCPR) shall be greater than or equal to the applicable Single Loop Operation MCPR limit as specified in the CORE OPERATING LIMITS REPORT.

5.

Table 3.3.6-2:

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

a.

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APRM - Flow Biased Trip Setpoint

'.66W

+ 36%

A I

- Trip Setpoint 6 0.58W + 45%

. Allowable Value' 0.66W + 39%

'- 'Allowable Vi(ue-6 0.58W + 48%

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2" +, except during two loop operation.¹ ACTION:

In OPERATIONALCONDITION 1:

With a) no reactor coolant system recirculation loops in operation, or b)

Region I of Figure 3.4.1

~ 1.1-1 entered, or c)

Region II of Figure 3.4.1.1.1-1 entered and core thermal hydraulic instability occurring as evidenced by:

SUSQUEHANNA - UNIT 2 3/4 4-1c Amendment No.9E> 95

0 3/4. 1 REACTIVITY CONTROL SYSTEMS BASES 3/4. 1. 1 'HUTS'ARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritfcal from all operating conditions,

2) the reactivity transients associated with postulated accident conditions are controllable within acceptable lfmfts, and 3) the reactor will be maintained sufficiently subcrftical 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 burnvp, the dtmonstration of SHUTOOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.3'elta k/k or R + 0.2'elta k/k, as appro-priate.

The valve of R in units of 'X 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 art supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

The hfghest worth rod may be determined analytically or by test.

The SHUTOOWN MARGIN fs demonstrated by control rod withdrawal at the beginning of life fuel cycle conditions, and, if necessary, at any future time in the cycle if the first demonstration fndfcates 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.

This reactivity characteristic has been a basic assumption in the analysis of plant performance and can bc best demonstrated at the time of fuel loading, but thc 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 tht predicted conditions is necessary.

Any, changes in reactivity from that of the predicted (predicted core k ff} can be determined from the core monitoring system (monitored core k ff}. In the cff abscncc of any deviation fn plant operating conditions or reactivity anomaly, these values should bc essentially equal

. ince the calculational methodologies are consistent.

The predicted co> e k ff is calculated by a 3D core simulation eff

'ode as a function of cycle exposure.

This is performed for projected or anticipated reactor operating states/conditions throughout the cycle and is usually dont 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, l~( L ~

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Amendment No. 31

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REA TIVITY NTROL YSTEMS BASES REACTIVITYAN IVIALIES (Continued)

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

A 1% deviation in reactivity from*that of the predicted is larger than expected for normal operation, and therefore should be thoroughly evaluated.

A deviation as large as 143 would not exceed the design conditions of the reactor.

4.1.

NTR L RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGINis maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) limitthe potential effects of the rod drop accident.

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

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

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

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

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

The number of control rods permitted to be inoperable could be more than the eight allowed by the 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.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the limit specified in Specification 2.1.2 during the core wide transient analyzed for the specific reload cycle.

The MCPR operating limits as specified in the CORE OPERATING LIMITS REPORT may be a function of average scram speed.

In such a case, the results of the required scram time testing (Specification 4.1.3.3) are used to adjust the MCPR operating limits to assure the validity of the cycle specific transient analyses.

This ultimately assures that MCPR remains greater than the limitspecified in Specification 2.1.2. The occurrence of scram times longer than those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

Tive scram discharge volume is require'd to be OPERABLE so that it will be available when needed to accept discharge water from the coniiol rods during a SUSQUEHANNA - UNIT 2 B.3/4 1-2 Amendment No. 9l', 95

REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued) reactor scram and will isolate the reactor coolant systea 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 novel 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.

Con~1 rod coupling integrity is required to ensure coepliance with the analysis of the rod drop accident in the FSAR.

The overtravel position feature provides the only positive aeans of determining that a rod is properly coupled and therefore this check aust 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 backup 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 position indication system aust be OPERABLE.

The control rod housing support restricts the outward movement of a control 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 amount of rod withdrawal is less than a dorsal'withdra~al increment and will not contribute to any damage to the priaary coolant systea.

The support is not required when there is no

. pressure to act as a driving force to rapidly e)ect a drive housing.

The required surveillance intervals are adequate to determine that the

'rods ar'e OPERABLE and not so frequent as to cause excessive wear on the system coaponents.

3/4. 1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the aaxiata insequence individual control rod or control rod segments which are withdrawn at any tiae during the fuel cycle cauld not be worth enough to result in a peak fuel enthalpy greater tht:n 280 rnl/gn in the event of a control rod drop accident.

The specified sequence:

are ct..racterized by homogeneous, scattered patt'erns of control rod withdrawal.

(~~qn THERMAL PMER is greater than 20K of RATED THERMAL REER, 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 RM to be OPERABLE when THERMAL POMER is less than or equal to 20K of RATED THERMAL POQER provides adequate cont~ol.

SVSgUEHANNA - VNIT 2 8 )/4 1-3 Apendment No. 31

REACTIVITY NTROL SYSTEMS BASE NTR L R D PROGRAIVI CONTR LS (Continued)

The RSCS and RWM logic automatically initiates at the low power setpoint (20% of RATED THERMALPOWER) to 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 range of key reactor parameters (which envelope the operating ranges of these variables), the fuel enthalpy rise during a postulated control rod drop accident 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 determine the peak fuel rod enthalpy rise. This value is then compared against 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 referenced in Specification 6.9.3.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided.

Tripping one of the channels willblock erroneous rod withdrawal soon, enough to prevent fuel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

3 4.1.5 TANDBYLIQUID CONTROL SYSTEIVI 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 it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes.

A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown requirement.

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 and the required pumping rate is 41.2 gpm. The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the fillingof other piping systems connected to the reactor vessel.

The temperature requirement for the sodium pentaborate solution is necessary to ensure that the sodium pentaborate remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to 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. 91, 95

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WER Dl TRIBUTION LIIVIITS BASES 4.2.1 AVERA E 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.

The Technical Specification APLHGR for SNP fuel is specified to assure the PCT following a postulated LOCA willnot exceed the 2200'F limit.The limiting value for APLHGR is specified in the CORE OPERATING LIMITS REPORT.

The calculational procedure used to establish the APLHGR specified in the CORE OPERATING LIMITSREPORT is based on a loss-of-coolant accident analysis.

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

These models are part of the approved methodology referenced in Specification 6.9.3.

4.2.2 APRIVI ETP INT 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 (AOO), including transients initiated from partial power operation.

For SNP 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 the LHGR for APRM Setpoints Curve specified in the CORE OPERATING LIMITS REPORT.

The LHGR for APRM Setpoints Curve specified in the CORE OPERATING LIMITSREPORT is based on SNP's Protection Against Fuel Failure (PAFF) line which was developed using the approved methodology referenced in Specification 6.9.3. The LHGR for APRM Setpoints Curve specified in the CORE OPERATING LIMITSREPORT corresponds to the ratio of PAFF/1.2 under which cladding and fuel integrity is protected during AOOs.

SUSQUEHANNA - UNIT 2 B.3/4 2-1 Amendment No. B, 95

P WER Dl TRIBUTI N LIIVIIT BASES 4.2.

IVIINIIVI M RITI AL POWER RATI The required operating limit MCPRs at steady state operating conditions as specified in the CORE OPERATING LIMITS REPORT are derived from the established fuel cladding integrity Safety Limit MCPR, and analyses of abnormal operational transients, For any abnormal operational transient analysis with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip settings given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limitingtransients have been analyzed to determine which result in the largest reduction in CRITICALPOWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCPR. When added to the Safety LimitMCPR, the required minimum operating limit MCPR specified in the CORE OPERATING LIMITS REPORT is obtained.

The required MCPR Operating Limits as a function of core power, core flow, and plant equipment availability condition are specified in the CORE OPERATING LIMITS REPORT.

The cycle specific transient analyses to determine the MCPR operating limits were performed using the NRC approved methods referenced in Specification 6.9.3.

The MCPR operating limits as specified in the CORE OPERATING LIMITS REPORT may be specified as a function of average scram speed.

In such a case, the results of the required scram time testing (Specification 4.1.3.3) are used to adjust the MCPR operating limits to assure the validity of the cycle specific transient analyses.

This ultimately assures that MCPR remains greater than the limit specified in Specification 2.1.2 for all anticipated operational occurrences.

The CORE OPERATING LIMITS REPORT specifies core flow dependent MCPR operating limiis which assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor-generator speed control failure.

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

Therefore, automatic flow control operation is not permitted.

The CORE OPERATING LIMITS REPORT specifies the power dependent MCPR operating limits which assure that the Safety Limit MCPR will not be violated in the event of a

'Feedwater Controller Failure, Rod Withdrawal Error, or Load Reject without Main Turbine Bypass operable initiated from a full power or reduced power condition.

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

Additional analyses are performed to determine the MCPR operating limit=with either the Main Turbine Bypass inoperable or the EOC-RPT inoperable.

Analyses to determine thermal margin with both the EOC-RPT inoperable and Main Turbine Bypass inoperable have not been performed.

Therefore, operation in this condition is not permitted.

SUSQUEHANNA - UNIT 2 B 3/4 2-2 Amendment No. 97., 9S

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FIGURE 5.1.3 1b NP DKFINING UNKSTRICTED AREAS t:OR RADIOACTIVE GASEOUS AND LIQUID'EFFLUEMTS SUSgUEO!lL LNIT 2 5-5

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEIVIBLIES 5.3.1 The reactor core shall contain 764 fuel assemblies.

Each assembly consists of a matrix of Zircaloy clad fuel rods with an initial composition of non-enriched or slightly enriched uranium dioxide as fuel material and water rods.

Limited substitutions of Zirconium alloy filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by test or analyses to comply with all fuel safety design bases.

A limited number of lead use assemblies that have not completed representative testing may be placed in non-limiting core regions.

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

Reload fuel shall have a maximum average enrichment of 4.0 weight percent U-235.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemblies.

The control material shall be boron carbide powder (B4C), and/or Hafnium metal.

The control rod shall have a nominal axial absorber length of 143 inches.

Control rod assemblies shall be limited to those control rod designs approved by the NRC for use in BWRs.

5.4 REACTOR COOLANT SYSTEIVI DESIGN PRESSURE AND TEIVIPERATURE 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.

IJ VOLUME c.

For a temperature of 575'F.

5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal T,, of 528'F.

SUSQUEHANNA - UNIT 2 5-6 Amendment No. 95

ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT" 6.9.1.8 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be sub-mitted within 60 days after January 1 and July 1 of each year.

The period of the first report shall begin with the date of initial criticality.

The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility as outlined in Regulatory Guide 1.21, "Measuring, Evaluating; and Reporting Radioactivity in Solid Wastes and Releases of Radio-active Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants,"

Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Semiannual Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year.

This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direc-tion and atmospheric stability, and precipitation (if measured),

or in the form of joint frequency distributions of wind speed, wind direction, atmospheric stability.** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figures

5. 1.3-1a and 5. 1.3-1b) during the report period.

All assump-tions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports.

The assessment of radiation doses shall be performed in accordance with the methodology and parameters of the Offsite Dose Calculation Manual (ODCM).

The Semiannual Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBERS OF THE PUBLIC from reactor releases and other near by uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation.

Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide l. 109, Rev.

1, October 1977.

• A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common'o all units at the station;

however, for units with separate radwaste

systems, the submittal shall specify the releases of radioactive material from each unit.

• • In lieu of submission with the first half year Semiannual Radioactive Effluent Release Report.,

the licensee has the option of retaining this summary of re-quired meteorological data on site in a file that shall be provided to the NRC upon request.

SUS(UEHANNA " UNIT 2 6-19

ADIVIINISTRATIVECONTROLS EIVIIANN AL RADI A TIVE EFFL ENT RELEA E REP RT (Continued)

The Semiannual Radioactive Effluent Release Reports shall include the following information for each type, of solid waste (as defined in 10 CFR PART 61) shipped offsite during the report period:

1. Container volume,

2. Total curie quantity (specify whether determined by measurement or estimate),

3.

Principal radionuclides (specify whether determined by measurement or estimate),

4. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

5. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

6. Solidification agent or absorbent (e.g., cement; urea formaldehyde).

The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period,to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL(ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to

= Specification 3.12.2.

PE IAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office within the time period specified for each report.

CORE OPERATIN LIMITS REPORT 6.9.3 The CORE OPERATING LIMITSREPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

6.9.3.1 Core operating limits shall be established prior to the startup of each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

aO C.

d.

e.

The Average Planar Linear Heat Generation Rate (APLHGR) for Specification 3.2.1.

The Linea.

Heat G-neration Rate for Average Power Range Monitor (APRM)

Setpoints for Speciii "ation 3.2.2.

The Minimum Critical Power Ratio (MCPR) for Specification 3.2.3 and 3.4.1.1.2.

The Linear Heat Generation Rate (LHGR) for Specification 3.2.4.

The Thermal Power Restrictions for Specification 3.4.1.1.1 and 3.4.1.1.2.

And shall be documented in the CORE OPERATING LIMITS REPORT.

SUSQUEHANNA - UNIT 2 6-20 Amendment No. 95

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~

J ADIVIINISTRATIVECONTROLS RE OPERATIN LIMITS REPORT (Continued) 6.9.3.2 The analytical methods used to determine the core operating limits shall be those topical reports and those revisions and/or supplements of the topical report previously reviewed and approved by the NRC, which describe the methodology applicable to the current cycle.

For Susquehanna SES the topical reports are:

1.

PL-NF-87-001-A, "Qualification of Steady State Core Physics Methods for BWR Design and Analysis," July, 1988.

2.

PL-NF-89-005-A, "Qualification of Transient Analysis Methods for BWR Design and Analysis," July, 1992.

3.

PL-NF-90-001-A, "Application of Reactor Analysis Methods for BWR Design and Analysis," July, 1992.

4.

XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads,"

Exxon. Nuclear Company, Inc., June 1986.

5.

XN-NF-85-67(P)(A), Revision 1; "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, Inc., September 1986.

6.

PLA-3407, "Proposed Amendment 132 to License No. NPF-14: Unit 1 Cycle 6 Reload," Letter from H. W. Keiser (PP&L) to W. R. Butler (NRC), July 2, 1990.

7.

Letter from Elinor G. Adensam (NRC) to H. W. Keiser (PP&L), "Issuance of Amendment No. 31 to Facility Operating License No. NPF-22

- Susquehanna Steam Electric Station, Unit 2," October 3, 1986.

8.

PLA-3533, Revised Proposed Amendment 67 to License No. NPF-22: Unit 2 Cycle 5 Reload," Letter from H. W. Keiser (PP&L) to W. R. Butler (NRC), March 7, 1991.

9.

XN-NF-84-97, Revision 0, "LOCA-Seismic Structural Response of an ENC 9x9 Jet; Pump Fuel Assembly," Exxon Nuclear Company, Inc., December 1984.

10. PLA-2728, "Response to NRC Question: Seismic/LOCA Analysis of U2C2 Reload,"

Letter from H. W. Keiser (PP&L) to E. Adensam (NRC), September 25, 1986.

11. XN-NF-82-06(P)(A), Supplement 1, Revision 2, "Qualification of Exxon Nuclear Fuel for Extended Burnup Supplement 1 Extended Burnup Qualification of ENC 9x9 Fuel," May 1988.

12. XN-NF-80-19(A), Volume 1, and Volume 1 Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors:,Peutronic Methods for Design and Analysis," Exxon Nuclear Company, Inc., March 1983.

13. XN-NF-524(A), Revision 1, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors,"

Exxon Nuclear Company, Inc., November 1983.

SUSQUEHANNA - UNIT 2 6-20a Amendment No. ~

ADMINISTRATIVECONTROLS CORE OPERATING LIMITS REPORT (Continued)

14. XN-NF-512-P-A, Revision 1 and Supplement 1, Revision 1, "XN-3 Critical Power Correlation," October, 1982.

15. XN-NF-80-19(A), Volumes 2, 2A, 2B, and 2C, "Exxon Nuclear Methodology for Boiling Water Reactors:

EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, Inc., September 1982.

16. XN-NF-CC-33(A), Revision 1, "HUXY:A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon Nuclear Company, Inc., November 1975.

17. XN-NF-82-07(A), Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., November 1982.

18. XN-NF-84-117(P), "Generic LOCA Break Spectrum Analysis: BWR 3 and 4 with Modified Low Pressure Coolant Injection Logic," Exxon Nuclear Company, Inc.,

December 1984.

19. XN-NF-86-65, "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for 9x9 Fuel," Exxon Nuclear Company, Inc., May 1986.

6.9.3.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

'.10.1 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time interval at each power level..

SUSQUEHANNA - UNIT 2 6-20b Amendment No. 95