Text

Amends 32 & 12 to Licenses NPF-68 & NPF-81,respectively, Removing Limits Associated W/Reactor Physics That Are Fuel Cycle Specific & Placed in Separate Core Operating Limits Rept
	ML20058L368
Person / Time
Site:	Vogtle
Issue date:	07/30/1990
From:	Matthews D; Office of Nuclear Reactor Regulation
To:	; Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA
Shared Package
ML20058L373	List: ML20058L368; ML20058L374;
References
	NPF-68-A-032, NPF-81-A-012 NUDOCS 9008070131
	Download: ML20058L368 (51)
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UNITED STATES sj.

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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA V0GTLE ELECTRIC GENERATING PLAhT UNIT 1

/MENDMENT TO FACILITY OPERATING LICENSE Amendment No. 32 License No. NPF-68 4

1.

The Nuciear Regulatory Comission (the Comission) has found that:

The app (lication for amendment to the Vogtle Electric Generating Plant.

A.

Unit 1 the facility) Facility Operating License No. NPF-68 filed by the Georgia Power Company acting for itself. Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the Itcenseas) dated March 22, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

C.

There is reasonable assurancs (1) that the activities authorized by I

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 Comission's regulations set forth in 10 CFR j

Chapter I; l

D.

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

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

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Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPT-68 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specificat'ons contained in Appendix A, as revised through Amendment No. 32, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. GPC shall operate the fecility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amer.dment is effective as of its date of issuance and shall i

be implemented within 30 days of issuance.

FOR THE HUCLEAR REGULATORY COMMISSION L

e/

David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of nuclear Reactor Regulation

Attachment:

1 Technical Specification Changes Date of Issuance:

July 30, 1990

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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORG!A CITY OF DALTON, GEORGIA V0GTLE ELECTRIC CENERATING PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Agendment No.12 License No hPF-81 1.

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

A.

Theapp(licationforamendmenttotheVogtleElectricGeneratingPlant, Unit 1 the facility) Facility Operating License No. NPF-81 filed by the Georgia Power Company acting for itself Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensees) dated March 22, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as cinended (the Act), and the Comission's rules and regulations set i

forth in.10 CFR Chapter I; B.

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

Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health _and Jafety of the public, and (ii) that such activities will be conducted L

in compliance with the Comission's regulations set forth in 10 CFR p

Chapter It D.

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's %1ations and all applicable requirements have been satisfied.

I

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2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-81 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as rev n ed through Amendment No.

12, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. GPC 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 shall be implemented within 30 days of issuanca.

FOR THE NUCLEAR REGULATORY COMMIS$10M I

(

4 David B.

atthews, Director Project Directorate !!-3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation I

Attachment:

Technical Specification Changes Date of Issuance: July 30, 1990 l

l

ATTACHMENT TO LICENSE AMENDMENT NO. 32 FACILITY OPERAT1hG LICENSE NO. NPF-68 AND LICENSE AMENDMENT NO. 12 1

FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NOS. 50-424 AND 50-425 l

l Replace the following pages of the Appendix "A" Technical Specifications with the enclosed paces. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.

j Pemove Pages Insert Pages i

I and II I and !!

l III* and IV III and IV V and VI*

Y and VI XXIII and XXIV

XXIII and XXIV 1-1* and 1-2 1-1 and 1-2 1-3 and 1-4 1-3 and 1-4 1-5 and 1-6 1-5 and 1-6 1-7 and 1-8*

1-7 and 1-8 3/4 1-1 ana 3/4 1-2*

3/4 I-1 and 3/4 1-2 3/4 1-3 and 3/4 1-3a 3/4 1-3b and 3/4 1-3c 3/4 1-3d and 3/4 1-4**

3/4 1-3 and 3/4 1-4 3/4 1-5 and 3/4 1-6*

3/4 1-5 and 3/4 1-6 3/4 1-13* and 3/4 1-14 3/4 1-13 and 3/4 1-14 3/4 1-15 and 3/4 1-16*

3/4 1-15 arid 3/4 1-16 3/4 1-20 3/4 1-20 i

3/4 1-21 and 3/4 1-22 3/4 1-21 l

3/4 2-1 and 3/4 2-2 3/4 2-1 and 3/4 2-2 3/4 2-3 and 3/4 2-4 3/4 2-3 and 3/4 2-4 3/4 2-5 and 3/4 2-6 3/4 2-5 and 3/4 2-6 3/4 2-7 and 3/4 2-8 3/4 2-7 and 3/4 2-8 B 3/4 1-1 and B 3/4 1-2 B S/4 1-1 and B 3/4 1-2 B 3/4 2-1 and B 3/4 2-2 B 3/4 2-1 and B 3/4 2-2 B 3/4 2-5 and B 3/4 2-6*

B 3/4 2-5 and B 3/4 2-6 6-21 and 6-22*

6-21 and 6-21a l

6-22

0verleaf page containing no change, j

- Two versions of page 3/4 1-4 are included:

one version is imediately l

ef fective and shall be implemented with 30 days of issuance; the other version reflects changes issued by Amendnents 25 and 6 which beccine effective following shutdown from Unit 2 Cycle 1 operation, and the page is marked accordingly, i

INDEX DEFINITIONS StCTION PAGE 1.0 DEFINITIONS 1.1 ?CTI0N........................................................

1-1 1.2 ACTUATION LOGIC TEST..........................................

1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................

1-1 1.4 AXIAL FLUX DIFFERENCE.........................................

1-1 1.5 CHANNEL CALIBRATION...........................................

1-1

1. 6 CHANNEL CHECK.................................................

1-1 1.7 CONTAINMENT INTEGRITY.........................................

1-2

~

1. 8 CONTROLLED LEAKAGE............................................

1-2 1.9 CORE ALTERATIONS........................

1-2 1.10 CORE OPERATING LIMITS REP 0RT.................................

1-2 j

1.11 DOSE E0VIVALENT I-131...........,............................

1-2 1.12 E-AVERAGEDISINTEGRATIONENERGY...

1-2 1

1,13 ENCINEERED SAFETY FEATURES RESFONSE TIME...................-..

1-3 1.14 FREQUENCY N0TATION...............

1-3 1.15 GASEOUS WASTE PROCESSING SYSTEM..............................

1-3 1.16 IDENTIFIED LEAKAGE...........................................

1-3 1.17 MASTER RELAY TEST............................................

1-3 1.18 MEMBER (S) 0F THE PUBLIC......................................

1-3 1.19 0FFSITE DOSE CALCULATION MANUAL..............................

1-4 1.20 OPERABLE - OPERABILITY.......................................

1-4 1.21 OPERATIONAL MODE - M0DE......................................

1-4 1.22 PHYSICS TESTS................................................

1-4 1.23 PRESSURE BOUNDARY LEAKAGE....................................

1-4 1.24 PROCESS CONTROL PR0 GRAM......................................

1-4 1.25 PURGE - PURGING..............................................

3-4 1.26 QUADRANT POWER TILT RATI0....................................

1-5 1.27 RATED THERMAL P0WER..........................................

1-5 1.28 REACTOR TRIP SYSTEM RESPONSE TIME............................

3-5 1.29 REPORTABLE EVENT.............................................

5 1.30 SHUTDOWN MARGIN..............................................

1-5 1.31 SITE B0VNDARY................................................

1-5 V0GTLE UNITS - 1 & 2 I

Amendment No.

32 (Unit 1)

Amendment No.

12 (Unit 2)

b INDEX DEFINITIONS SECTION PAGE 1,32 SLAVE RELAY TEST.............................................

1-5 1.33 SOLIDIFICATION...............................................

1-5 1.34 SOURCE CHECK.................................................

1-6 1.35 STAGGERED TEST BASIS..........................................

1-6 1.36 THERMAL P0WER................................................

1-6 1.37 TRIP ACiUATING DEVICE OPERATIONAL TEST.......................

1-6

1. 3 8 ' U N I D E NT I F I E D L E A KAG E.........................................

1-6 1.39 UNRESTRICTED AREA............................................

1-6 1.40 VENTILATION EXHAUST TREATMENT SYSTEM.........................

1-6 l

1.41 VENTING......................................................

1-7 l-TABLE 1.1 FREQUENCY N0TATION......................................

1-8 TABLE 1.2 OPERATIONAL M00ES.......................................

1-9 l

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t-1.

t i

l 1.

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I i

l.

l' V0GTLE UNITS - 1 & 2 II Amendment No.

32 (Unit 1)

Amendment No.

12 (Unit 2) l I'l' a.

INDEh l

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS p

(-

SECTION PAGE 2.1 SAFETY LIMITS i

2.1.1-REACTOR C0RE................................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE...................-..........

2-1 FIGURE 2.1-l' REACTOR CORE SAFETY LIMIT............................

2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS i-l-

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS...............

2-3 TABLE 2.2-1 REACTOR TRIP. SYSTEM INSTRUMENTATION TRIP SETPOINTS....

2-4

.1 l

i l

-BASES SECTION

-PAGE l

l 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

B'2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS

-2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS..........-.....

B 2-3 l

.1 1

V0GTLE UNITS - 1 & 2 III i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - MODES 1 and 2.........................

3/4 1-1 Shutdown Margin - MODES 3, 4 and S.......................

3/4 1-3 Moderator Temperature Coefficient........................

3/4 1-4 Minimum Temperature for Criticality......................

3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Flow Path - Shutdown.....................................

3/4 1-7 Flow Paths - Operating...................................

3/4 1-8 Charging Pump - Shutdown.................................

3/4 1-9 Charging Pumps - Operating...............................

3/4 1-10 Borated Water Source - Shutdown..........................

3/4 1-11 Borated Water Sources - Operating........................

3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................

3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE

. EVENT OF AN IN0PERABLE CONTROL OR SHUTDOWN R0D...........

3/4.1-16 Position Indication Systems - Operating..................

3/4 1-17 Position Indication System - Shutdown....................

3/4 1-18 Rod Drop Time............................................

3/4 1-19 Shutdown Rod Insertion Limit.........

3/4 1-20 Control Rod Insertion Limits.............................

3/4 1-21 V0GTLE UNITS - 1 & 2 IV Amendment No. 32 (Unit 1)

Amendment No.

12 (Unit 2)

_INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE' REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE....................................

3/4 2-1 FIGURE 3.2-1 (DELETED) 3/4 2 l 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z).....................

3/4 2-4 9

FIGURE 3.2-2 (DELETED) 3/4 2 l' 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F 3/4 2-0 H''''''

3/4.2.4 QUADRANT POWER TILT P.ATI0................................

3/4 2-10 3/4.2.5 DNB PARAMETERS...........................................

3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3-1 TABLE.3.3-1, REACTOR TRIP SYSTEM INSTRUMENTATION...................

3/4 3-2 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................

?3-3/4.3.2 ENGINEERED SAFETY F2ATURES ACTUATION SYSTEM INSTRUMENTATION........................................

b/4 TABLE 3.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTR #iENTATION..........................................

3/4 3-17 L

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS...........................

3/4 3-28 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-36 L

3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................

3/4 3-45 TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS.....................................

3/4 3-46 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.....................

3/4 3-48 i.

Movable Incore Detectors.................................

3/4 3-49 Seismic Instrumentation (Common System)..................

3/4 3-50 V0GTLE UNITS - 1 & 2 V

Amendment No.

32 (Unit 1)

Amendment No.

12 (Unit 2)

INDEX

' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-5 SEISMIC MONITORING INSTRUMENTATION....................

3/4 3-51 n

L TABLE 4.3-4 SEISMIC MONITORING' INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................

3/4 3-52 Meteorological Instrumentation (Common System)...........

3/4 3-53 TABLE 3.3-6 METEOROLOGICAL MONITORING INSTRUMENTATION.............

3/4 3-54 Remote Shutdown System...................................

3/4 3 -

TABLE 3.3-7 REMOTE SHUT 00WN SYSTEM MONITORING INSTRUMENTATION.....

3/4 3-56 Accident Monitoring Instrumentation......................

3/4 3-58

.e TABLE 3.3-8' ACCIDENT MONITORING' INSTRUMENTATION...................

3/4 3-59 Chlorine Detection Systems (Deleted)....................,

3/4 3-63 i

1 Loose Parts Detection System (Deleted)..................

3/4 3-64 t

Radioactive Liquid Effluent Monitoring Instrumentation...

3/4 3-65 j

TABLE 3.3-9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-66 TABLE 4.3-5 RADI0 ACTIVE LIQUID EFFLUENT MONITORING 1

INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-68 Radioactive Gaseous Effluent Monitoring Instrumentation..

3/4 3-71 TABLE 3.3-10 RADIOACTIVE GASE0US EFFLUENT MONITORING l

l INSTRUMENTATION..........................................

3/4 3-72 TABLE 4.3-6 RADI0 ACTIVE GASE0US EFFLUENT MONITORING

' INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-76 High-Energy Line Break Isolation Sensors.................

3/4 3-79 TABLE 3.3-11 HIGH-ENERGY LINE BREAK INSTRUMENTATION...............

3/4 3-80 3/4.3.4 TURBINE OVERSPEED PR0TECTION..............................

3/4 3-81 V0GTLE UNITS - 1 & 2 VI

INDEX ADMINISTRATIVE CONTROLS SECTION i

4 6.4.2 SAFETY REVIEW BOARD (SRB)

Function..................................................

6-9 Composition...............................................

6-10 Alternates................................................

6-10 Consultants...............................................

6-10 Meeting Frequency.........................................

6-10 Quorum....................................................

6-10 Review....................................................

6-11 Audits...................................................

6-11 Records...................................................

6-12 6.5 REPORTABLE EVENT ACTI0N.....................................

6 6.6 SAFETY LIMIT VIOLATION......................................

6-13 6.7 PROCEDURES AND PR0 GRAMS.....................................

6-13 6.8 REPORTING REQUIREMENTS l

6.8.1 ROUTINE REP 0RTS...........................................

6-17 Startup Report............................................

6-17 Annual Reports............................................

6-17 Annual Radiological Environmental Surveillance Report.....

6-18 I

Semiannual Radioactive Ef fluent Release Report............

6-19 Monthly Operating Reports.................................

6-21 Core Operating Limits Report..............................

6-21 l

I 6.8.2 SPECIAL REP 0RTS...........................................

6-21 6.9 RECORD RETENTION............................................

6-22 V0GTLE UNITS - 1 & 2 XXIII Amendment No.

32 (Unit 1)

Amendment No.

12 (Unit 2)

["

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INDEX ADMINISTRATIVE' CONTROLS I

SECTION 6.10 RADIATION PROTECTION PR0 GRAM...............................

6-23 6.11! HIGH RADIATION AREA........................................

6-23

.l 6.12 PROCESS CONTROL PROGRAM (PCP)..............................

6-24 6.13 0FFSITE DOSE CALCULATION MANUAL (00CM).....................

6-25 6.14 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID j

F L

RADWASTE TREATMENT SYSTEMS.................................

6-25' I

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V0GTLE UNITS - 1 & 2 XXIV l

l

' 1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

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

ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the apriication of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output.

The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.

^

ANALOG CHANNEL OPERATIONAL TEST

1. 3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions.

The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the setpoints are within the required range and accuracy.

AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in flux signals (normalized to their full power sum and expressed as their percentage) between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.6 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.

V0GTLE UNITS - 1 & 2 1-1

.=

'~OEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

All penetrations required to be closed during accident cor.ditions a.

are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)-

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.-

b..

All equipment hatches are closed and sealed, Each air lock is in compliance with the requirements of Specification c.

3.6.1.3, d.

The containment leakage rates are within the limits of Specification L

3.6.1.2, and The sealing mechanism associated with each penetration-(e.g., welds,.

-e.

l bel. lows, or 0 rings) is OPERABLE.

I CONTROLLED LEAKAGE 1

1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor l

coolant pump seals.

CORE: ALTERATIONS 1.9. CORE ALTERATIONS shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and _ fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude. completion of i

movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORTS-1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-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 J

cycle in accordance with Specification 6.8.1.6.

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

DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131'(microcurie / gram) l which alone would produce the same thyroid dose as the quantity and isotopic i

r mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 'of NRC Regulatory Guide 1.109, Revision 1, October 1977.

l V0GTLE UNITS - 1 & 2 1-2 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

o

t

DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average, weighted in proportion to the concentration of l

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 the.

isotopes with half' lives greater than 14 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE' TIME shall be thdt time I

. interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travet to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance l

Requirements shall correspond to the intervals defined in Table 1.1.

GASEGUS WASTE PROCESSING SYSTEM 1.15 A GASE0US WASTE PROCESSING SYSTEM shall be any system designed and installed l

to reduce radioactive gaseous. effluents by collecting Reactor Coolant System offgases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

g a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump sesi or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphert from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or c.

Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST 1.17 A MASTER RELAY TEST shall be the energization of each master relay and l

verification of OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

V0GTLE UNITS - 1 & 2 1-3 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

i 1

' DEFINITIONS I

MEMBER (S) 0F THE PUBLIC 1.18 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-

\\

tionally associated with the plant.

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

Also excluded from this category 1

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 recre-ational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.19 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology l

and parameters used in the :alculation 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 Environ-mental Radiological Monitoring Program.

)

OPERABLE - OPERABILITY 1.20 A system, subsystem, train, component or device shall be OPERABLE er l

l have OPERABILITY when it is capable of performing its specified function (s),

j 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 MODE - MODE 1.21 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive j

combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.22 PHYSICS TESTS shall be those tests performed to measure the fundamental l

nuclear characteristics of the reactor core and related instrumentation:

(1)' described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.23 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube l

leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM 1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, l

sampling, analyses, tests, and determinations to be made to ensure that proces-sing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State V0GTLE UNITS - 1 & 2 1-4 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

DEFINITION $

regulations, burial ground requirements, and other requirements governing the disposal of radioactive waste.

PURGE - PURGING 1.25 PURGE or PURGING shall be any controlled process of discharging air or gas l

from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore l

detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.- With one excore detector inoperable, the remaining three detectors.

j shall be used for computing the average.

1 RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to l

the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from l

when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in l

Sections 50.72 and 50.73 of 10 CFR Part 50.

SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which l

the reactor is subcritical or would be subcritical from its present condition assuming all rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.31 The SITE B0UNDARY shall be the exclusion boundary line as shown in l

Figure 5.1-1.

SLAVE RELAY TEST 1.32 A SLAVE RELAY TEST shall be the energization of each slave relay and l

verification of OPERABILITY of each relay.

The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices, i

V0GTLE UNITS - 1 & 2 1-5 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

DEFINITIONS SOLIDIFICATION 1.33 SOLIDIFICATION shall be'the conversion of wet wastes into a form that

\\

meets shipping and burial ground requirements.

SOURCE CHECK 1.34 A SOURCE CHECK shall be the qualitative assessment of channel response l-when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS i

1.35 A STAGGERED TEST BASIS shall consist of:

l A test schedule'for n systems, subsystems, trains, or other a.

designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.36 THERMAL POWER shall be the total reactor core heat transfer rate to the l

reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.37 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the l

Trip Actuating Device-and verifying OPERABILITY of alarm, interlock and/or-trip functions.- The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary,' of the-Trip. Actuating Device such that-it actuates i

at the= required Setpoint within the required accuracy.

1 UNIDENT!FIED LEAKAGE 1$38 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE l

or. CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.39 An UNRES1RICTED AREA shall be any area at or beyond the SITE BOUNDARY l

access to which~is not controlled by the licensee for' purposes of-protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational' purposes.

V0GTLE UNITS - 1 & 2 1-6 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2) 1

DEFINITIONS VENTILATION EXHAUST TREATMENT SYSTEM i

1.40 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l

-installed to reduce gaseous radioiodine o* radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particu-lates from the gaseous exhaust stream prior to the release to the environment.

Such a system is not considered to have any effect on noble gas effluents.

Engineered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.41 VENTING shall be the controlled procets of discharging air or gas from a l

confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that-replacement air or gas is not pro-vided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

'1 J

V0GTLE UNITS - 1 & 2 1-7 Amendment No.

32 (Unit 1)

Amendment No.

12 (Unit 2)

.L

.1 TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY

+

S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D:

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M At least once per 31' days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

i N.A.-

Not applicable.

P Completed prior to each release.

i I

j V0GTLE UNITS - 1 & 2 1-8

3/4.1 REACTIVITY' CONTROL' SYSTEMS

-3/4.1.1 BORATION CONTROL-SHUTDOWN MARGIN - MODES 1 AND 2 I

LIMITING CONDITION FOR OPERATION 3.1.1.1' The. SHUT 00WN MARGIN shall be greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

MODES 1 and 2*.

1 ACTION:

With the SHUTOOWN MARGIN less than the limit specified in the COLR, immediately l

' initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUT 00WN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limit specified in the COLR.

I a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at~least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b.

When in MODE 1 or' MODE 2 with K,ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is 1

within the limits of Specification 3.1.3.6; j

With K,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor-c.

l:

criticality by verifying that the predicted critical control rod l-position'is within the limits of Specification 3.1.3.6; and j

i d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors below, with the control l

banks at the maximum insertion limit of Specification 3.1.3.6:

i

See Special Test Exceptions Specification 3.10.1.

il V0GTLE UNITS - 1 & 2 3/4 1-1 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

nr gin y-

REACTIVITY CONTROL SYSTEMS qg SURVEILLANCE REQUIREMENTS (Continued)

-r 1)

Reactor Coolant System boron concentration, i

2)

Control rod position, ik 3)-

Reactor Coolant System average temperature, i

4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% ok/k at least once per 31 Effective Full Power Days (EFPD).

This comparison shall consider at least those. factors stated in Specification 4.1.1.1.1d, above.

The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD e

after each fuel loading.

5 i

E N

M T

V0GTLE UNITS - 1 & 2 3/4 1-2 4

i iiiumm-sm-

mm-

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - MODES 3, 4 AND 5 i

-. LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limits

.specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

MODES 3, 4 AND 5.

ACTION:

With the SHUTDOWN MARGIN less than the required value, immediately initiate and-continue boration at greater than or equal to 30 gpm of a solution containing

. greater than or equal to-7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

i SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the required value:

a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If an inoperable control rod (s) is immovable or eintrippable, the SHUTDOWN MARGIN shall'be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1)

Reactor _ Coolant System boron concentration,

)

2)'

Control rod position, 3)

Reactor Coolant System average temperature, 4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

V0GTLE UNITS - 1 & 2 3/4 1-3 Amendment No.

32 (Unit 1)

Amendment No.

12 (Unit 2)

c REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be within the beginning of cycle life (BOL) limit and the end of cycle life (EOL) limit specified in the CORE OPERATING LIMITS REPORT (COLR).

The maximum upper limit shall be:

Unit 1:

Less positive than +0.7 x 10 4 ok/k/*F for power levels up to l

70% RATED THERMAL POWER with a linear ramp to 0 ak/k/*F at 100%

RATED THERMAL POWER; and Unit 2:

Less positive than 0 ok/k/*F.

I APPLICABILITY: BOL limit - MODES l'and 2* only.**

EOL limit - MODES 1, 2, and 3 only.**

ACTION:

a.

With the MTC more positive than the BOL limit specified in the COLR, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within the BOL limit specified in the COLR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and

With K,ff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

V0GTLE UNITS - 1 & 2 3/4 1-4 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

' REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be within the beginning of cycle life (BOL)-limit and the end of cycle life (E0L) limit specified-in the CORE OPERATING LIMITS REPORT (COLR).

The maximum upper limit shall be:

Lessipositive than +0.7 x 10 4 Ak/k/ F for power levels up to 70% RATED THERMAL POWER with a linear ramp to O Ak/k/ F at 100%

RATED THERMAL POWER; and APPLICABILITY:' BOL limit - MODES 1 and 2* only.**

L E0L limit - HODES 1, 2, and 3 only **

1 ACTION:

With the MTC more positive than the BOL limit specified in the COLR, a.

operation in MODES I and 2 may proceed provided:

l 1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within the BOL limit specified in the COLR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next L

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to the I

insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and

With K,ff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

V0GTLE UNITS - 1 & 2 3/4 1-4 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

This page becomes effective following shutdown from Unit 2 Cycle 1 operation.

l

i REACTIVITY CONTROL SYSTEMS L

SURVEILLANCE REQUIREMENTS 3.

A Special Report is prepared and. submitted to the Commission, pursuant to Specification 6.8.2, within 10 days, describing the

.value of the measured MTC, the-interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition, b.

With the MTC more negative than the EOL limit specified in the COLR, l

be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.1.1.3 The MTC shall be determined to be within its limits during each fuel

~

cycle as follows:

i a.

The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, j.

after each fuel loading; and

'b.

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn,

[

]

RATED THERMAL POWER condition) within 7 EFP0 after reaching an equili-brium boron concentration of 300 ppm.

In the event this comparison indicates the MTC is more negative than the 300 ppm surveillance limit

(

specified-in the COLR, the MTC shall be remeasured, and compared to j

the EOL MTC limit specified in the COLR at least once per 14 EFPD during the remainder of the fuel cycle, i

V0GTLE UNITS - 1 & 2 3/4 3-5 Amendment No.

32 (Unit 1)

Amendment No.

12 (Unit 2)

REACTIVITY CONTROL SYSTEMS

' MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 -The Reactor Coolant System lowest operating loop temperature-(TI-0412, TI-0422, TI-0432, TI-0442) (T,yg) shall be greater than or equal to 551*F.

APPLICABILITY:

MODES 1 and 2.*#

L-ACTION:

With a Reactor Coolant System operating loop temperature (T,yg) less than 551'F, restore T,yg to within its limit within 15 minutes or be in-HOT

^

STANDBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS I

4.1.1.4 The Reactor Coolant System temperature (T"V9) shall_be determined to be greater than or equal to 551*F:

i Within 15 minutes prior to achieving reactor criticality, and a.

b.- At-least once pe: 30 minutes when the reactor is critical and the Reactor Coolant System T,yg (TI-0412, TI-0422, TI-0432, TI-0442) is less than 561*F with the T,yg-T On 'adon Alarm not reset.

t ref i

With K,77 greater than or equal to 1.

ee Special. Test Exceptions Specification 3.10.3.

S V0GTLE UNITS - 1 & 2 3/4 1-6

' REACTIVITY CONTROL SYSTEMS

' LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued).

c..With a Sludge Mixing Pump Isolation Valve (s) inoperable, restore the valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or isolate the sludge mixing system by either closing the manual isolation valves or-deenergizing the OPERABLE solenoid pilot valve within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and' maintain closed.

SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:

l-a.

At least once per 7 days by:

1)

Verifying the boron concentration in the water, 2)

Veri'ying the contained borated water volume of the water source, and 3)

When the boric acid storage tank is the source of borated water and the ambient temperature of the boric acid storage tank room (TISL-20902, TISL-20903) is < 72 F, verify the boric acid storage tank solution temperature is > 65'F.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature (TI-10982) when the outside air temperature is less than 50 F.

c.

At least once per 18 months by verifying that the Sludge Mixing Pump Isolation Valves automatically close upon RWST low-level test signal.

l-V0GTLE UNITS - 1 & 2 3/4 1-13 r

4

' REACTIVITY CONTROL SYSTEMS 3/4.1.3-MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION l

3.1.3.1 All shutdown and control rods shall be OPERABLE and positioned within i 12 steps-(indicated position) of their group demand position.

APPLICABILITY:

MODES 1* and 2*.

ACTION:

a.

With one or more rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be l-...

untrippable, determine that the SHUTDOWN MARGIN requirement of l-Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With one rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand hei y t by more than : 12 steps (indicated position),

POWE,R OPERATION may continue provided that within 1 hour:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or l-i li m 2.

The rod is declared inoperable and the remainder of the rods in I

the group with the inoperable rod are aligned to within i 12 steps L

of the inoperable rod while maintaining the rod sequence and Ll insertion limits of Specification 3.1.3.6.

The THERMAL POWER

_l l

level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

The rod is declared inoperable and the SHUTOOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER l

OPERATION may then continue provided that:

l a)

A reevaluation of each accident analysis of Table 3.1-1 is performed withi, 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; l

b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

V0GTLE UNITS - 1 & 2 3/4 1-14 Amendment No. 32 (Unit 1)

Amendment No.

12 (Unit 2)

f REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION

- ACTION (Continued) c)

A power distribution map is obtained from the movable incore'detectorsandF(Z)andFfg are verified to be 9

within_their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d)

The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85%

of RATED THERMAL POWER.

c.

With more than one rod trippable but inoperable due to e ases other I

than addressed by ACTION a above, power operation may concinue provided that:

l 1.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the rods in the bank (s) with the l

inoperable rods are aligned to within i 12 steps of the inoper-able rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6.

The THERMAL POWER level shall l

be restricted pursuant to Specification 3.1.3.6 during subsequent i

l.

operation, and l

4 l

2.

The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

L d.

With more than one rod misaligned from.its group step counter demand L

height by more than i 12 steps (indicated position), be in HOT-1

-STANDBY within 6-hours.

SURVEILLANCE REQUIREMENTS l

l l

4.1.3.1.1 The position of each rod shall be determined to be within the group l

- demand limit by verifying the individual rod positions at least once per L

12 hows except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each rod not fully inserted'in the core shall be determined to be l

OPERABLE by movement of at.least 10 steps in any one direction at least once per 31 days.

V0GTLE UNITS - 1 & 2 3/4 1-15 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

TABLE 3.1-1

~

ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE CONTROL OR SHUT 00WN ROD Rod Cluster. Control Assembly Insertion Characteristics Rod Cluster-Control Assembly Misalignment Decrease in Reactor Coolant Inventory Inadvertent Opening of a Pressurizer Safety or Relief Valve Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundry That Penetrate Containment Loss-of-Coolant-Accidents Increase in Heat Removal by the Secondary System (Steam System Piping Rupture)

Spectrum-of Rod Cluster Control Assembly Ejection Accidents, i

L l

l l

i 1

l l

.V0GTLE UNITS - 1 & 2 3/4 1-16 L

l l

REACTIVITYCONTROL'SYSTE3 SHUT 00WN R00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

MODES 1* and 2* #.

ACTION:

With a maximum of one shutdown rod inserted beyond the insertion limit speci-fied in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a.

Restore the rod.to within the insertion limit specified in the COLR,

-l b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the COLR:

a.

Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and b.

At least-once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K,ff greater than or equal to 1.

V0GTLE UNITS - 1 & 2 3/4 1-20 Amendment No. 32(Unit 1)

Amendment No.12(Unit 2)

' REACTIVITY CONTROL SYSTEMS

. CONTROL' ROD-INSERTION LIMITS i

' LIMITING CONDITION FOR OPERATION 3.1.3.6 The: control banks shall be limited in physical insertion as specified-in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

MODES 1* and 2* #.

ACTION:

With'the control banks inserted beyond the insertion limits, except for sur-l' veillance testing pursuant to Specification 4.1.3.1.2:

r a.

Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion.using the insertion limits specified in the COLR, or l

c.

Be in at least HOT. STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

' SURVEILLANCE REQUIREMENTS 4.1.3.6 The position.of_each control bank shall be determined to be within the insertion ~ limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod insertion limit monitor is inoperable, then verify the. individual rod _ positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K,ff greater than or equal to 1.

V0GTLE UNITS - 1 & 2 3/4 1-21 Amendment No. 32 (Unit 1)

Amendment No.12 (Unit 2)

l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFEREh!E LIMITING CONDITION FOR OPERATION 3.2.1 The indicated (NI-0041B, NI-0042B, NI-00438, NI-0044B) AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (flux difference units) about the target flux difference.

The target band is specified in the CORE OPERATING LIMITS REPORT (COLR).

The indicated AFD may deviate outside the required target band at greater than or equal to 50% but less than 90% of PATED THERMAL POWER provided the indicated AFD is Within the Acceptable Operation Limits specified in the COLR and the l

cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD may deviate outside the required target band at greater than l

15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILIH:

MODE 1, above 15% nf RATED THERMAL POWER.* #

ACTION:

a.

With the indicated AFD outside of the required target band and with l

THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:

1.

Restore the indicated AFD to within the target band limits, or 2.

Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

E th the indicated AFD outside of the required target band for more l

thaa I hour of cumuir ve penalty deviatio' time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside LP.e Acceptable Operation limits specified in the COLR and with THERMAL POWER less than 90% e.t > qual to or greater than 50% of RATED THERMAL POWER, reduce:

1.

THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and 2.

The Power Range Neutron Flux * - High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions Specification 3.10.2.

Surveillance testing of the Power Range Neutron Flux Channel may be performed (below 90% of RATED THERMAL POWER) pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits speci-fied in the COLR.

A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.

V0GTLE UNITS - 1 & 2 3/4 2-1 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) c.

With the indicated AFD outside of the required target band for more l

than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER 1ess than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the required target band and the cumulative penalty devia-l tion has been reduced to less than I hour in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a-Monitoring the indicated AFD for each OPERABLE excore channel:

1)

At least once por 7 days when the AFD Monitor Alarm is OPERABLE, and 2)

At least once per hour until the AFD Monitor Alarm is updated after restoration to OPERABLE status, b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable.

The logged values of the indicated AFD shall be assumed to u.ist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore.hannels are indicating the AFD to be outside the target band.

Penalty deviation outside of the required target band shall be l

accumulated on a time basis of:

a.

One minute penalty deviation for each I minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b.

One-half minute penalty deviation for each I minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4. 2.1. 3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at leai,t once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Ju11 Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life.

The provi-sions of Specification 4.0.4 are not applicable.

V0GTLE UNITS - 1 & 2 3/4 2-2 Amendment No.

32 (Unit 1)

. Amendment No.

12 (Unit 2)

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..:V0GTLEr. UNITS - l'& 2;-

3/4 2-3 Amendment _ No. 32 (Unit __1)

.s, Amendment No. 12 (Unit 2)-

I

+'1 g

.m

4 POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

9 RTP F (Z) 5 f Q

o P

i F (Z) $ F g

0

[K(Z ? for P $ 0.5

0. 5 i

Where:

F RTP= the F limit at RATED THERMAL POWER (RTP) g g

specified in the CORE OPERATING LIMITS REPORT j

(COLR),

Where:

P _ THERMAL POWER

, and RATED THERMAL POWER K(Z) = the normalized F (Z) as a function of core height specified in the COLR. 9 APPLICABILITY:

MODE 1.

ACTION:

With F (li exceeding its limit:

9 a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds q

the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subseouent POWER OPERATION mdy proceed provided the Overpower AT Trip Setpoints have been reduced at least 1%

for each 1% F (Z) exceeds the limit; and g

b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be g

within its limit.

V0GTLE UNITS - 1 & 2 3/4 2-4 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

l l

FIGURE 3.2-2 (DELETED) l V0GTLE UNITS - 1 & 2 3/4 2-5 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F

shall be evaluated to determine if F (Z) is within its limit by:

xy 9

a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER b.' fore exceeding 75% of RATED THERhAL POWER following each fuel

loading, Increasing the measured F,y component of the power distribution map b.

by 3% to account for manufacturing tolerances and further increasing P

the value by 5% to account for nieasurement uncertainties, c.

Comparing the F computed (Fx ) obtained in Specification 4.2.2.2b.,

above to:

xy 1)

The F,y limits for RATED THERMAL POWER (FRTP) for the appropriate x

measured core planes given in Specification 4.2.2.2e. and f.,

below, and 2)

The relationship:

RTP F,

=F p pp (3 p) g x

l Where F is the limit for fractional THERMAL POWER operation RTP expressed as a function of F

, PF is the power factor x

xy multiplier for F specified in tne COLR, and P is the fraction xy of RATED THERMAL POWER at which "xy was measured, d.

Remeasuring F according to the fol'owing schedule:

xy FTP 1)

When F is greater than the F limit for the appropriate x

measured core plane but less than the F relationship, additional P

power distribution maps shall be taken dF compared to F x

and F h either:

x a)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERhAL C

POWER or greater, the THERMAL POWER at which F*Y was last determined, or b)

At least once per 31 Effective Full Power Days (EFPD,,

whichever occurs first.

V0GTLE UNITS - 1 & 2 3/4 2-6 Amendment No.

32 (Unit 1)

Amendment No.

12 (Unit 2)

J

' POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

C RTP 2)

When the F is less than or equal to the F limit for the x

x appropriate measured core plane, additional power distribution andFfatleast maps shall be taken and F compared to F x x

once per 31 EFPD.

e.

The F limits used in the Constant Axial Offset Control analysis for RA ED THERMAL POWER (FRTP) shall be specified for all core planes x

containing Bank "D" control rods and all unrodded core planes in the COLR per Specification 6.8.1.6; f.

The F limits of Specification 4.2.2.2e., above, are not applicable j

yy in the following core planes regions as heasured in percent of core height from the bottom of the fuel:

1)

Lower core region from 0 to 15%, inclusive, 2)

Uppe, core region from 85 to 100%, inclusivc.

3)

Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 2%,

and 74.9 1 2%, inclusive, and

)

4)

Core plane reg #ons within i 2% of core height [t 2.88 inches) about the bank demand position of the Bank "D" control rods.

WithFfexceedingF the effects of F on F (2) shall be evaluated g.

x x

xy 9

to determine if F (Z) is within its limits.

g 4.2.2.3 When F (Z) is measured for other than F determinations, an overall 9

xy measured F (Z) shall be obtained from a power distribution map and incusased g

by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

V0GTLE UNITS - 1 & 2 3/4 2-7 Amendment No. 32 (Unit 1)

Amendment No.12 (Unit 2)

POWER DISTRIBUTION LIMITS l

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR-F H

LIMITING CONDITION FOR OPERATION FfH shall be limited by the following relationship:

3.2.3 F gi FhPgy pp H(1-P)]

Where:

FRTP= The F limit at RATED THERMAL POWER (RTP) specified in H

the CORE OPERATING LIMITS REPORT (COLR).

AH=thePowerFactorMultiplierforifH specified in the PF COLR, and

^

THERMAL POWER p - RATED fMERMAL POWER APPLICABILITY:

MODE 1.

ACTION:

With F exceeding its limit:

H a.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1.

Restore F to within the above limit, or H

2.

Reduce THERMAL POWER to less than 50T of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limit, verify through incore flux mapping that F has been restored to within H

the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and c.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced 7HERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent POWER OPERATION may proceed provided that F is demonstrated, through incore flux H

mapping to be within its limit prior to exceeding the following THERMAL POWER levels:

1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL POWER, and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

V0GTLE UNITS - 1 & 2 3/4 2-8 Amendment No. 32 (Unit 1)

. Amendment No. 12 (Unit 2)

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN i

A sufficient SHUTDOWN MARGIN ensures that:

(1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to i

precluoe total loss of SHUTDOWN MARGIN in the shutdown condition.

1 SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg.

In MODES 1 and 2, the most restrictive condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a mini-mum SHUTDOWN MARGIN of 1.3% Ak/k is required to control the reactivity transient, Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condi-ion and is consistent with FSAR safety analysis assumptions.

In MODES 3, 4 and 5, the most restrictive condition occurs at BOL, associated with a boron dilution accident.

In the analysis of this accident, a minimum SHUTDOWN MARGIN as defined in Specification 3/4.1.1.2 is required to allow the operator 15 minutes l

from the initiation of the Source Range High Flux at Shutdown Alarm to total l

loss of SHUTDOWN MARGIN.

Accordingly, the SHUTDOWN MARGIN requirement is based l

upon this limiting requirement and is consistent with the FSAR accident analysis assumptions.

The required SHUTDOWN MARGIN is specified in the CORE OPERATING LIMITS REPORT (COLR).

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are k ovided to ensure that the value of this coefficient remains within the limit ng condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

i The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.

These corrections l

V0GTLE UNITS - 1 & 2 B 3/4 1-1 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

1 REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting EOL MTC value.

The 300 ppm surveillance limit MTC value represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting EOL MTC value.

l The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551'F.

This limitation is required to ensure:

(1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.

NDT 3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation.

The components required to perform this function include:

(1) borated water sources, (2) charging pumps, (3) separate flow paths, and (4) the boric acid transfer pumps.

With the RCS average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure functional capability in the event an assumed single failure renders one of the flow paths inoperable.

The beration capability of either flow path is sufficient to provide a SHUTDOWN V0GTLE UNITS - 1 & 2 B 3/4 1-2 Amendment No. 32 (Unit 1) 4 Amendment No.12 (Unit 2) l

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local teat 9

flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; F

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of g

the integral of linear power along the rod with the highest integrated power to the aserage rod power; and F*Y(Z)

Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation 2.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper 9

bound envelope of the F limit specified in the CORE OPERATING LIMITS REPORT q

(COLR) times K(Z) is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

V0GTLE UNITS - 1 & 2 B 3/4 2-1 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

PO R DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)

Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL PUWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.

This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the tar-get band) provided the time duration of the deviation is limited.

Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits specified in the COLR while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant.

The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F H

The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that:

(1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position; b.

Control rod banks are sequenced with a constant tip-to-tip distance between banks as defined by Figure 3.1-3.

V0GTLE UNITS - 1 & 2 B 3/4 2-2 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

The Radial Peaking Feetor, fxy(Z), is measured periodically to provide assurance that the Hot Channal Factor, F (2), remains within its limit.

The 9

F limit for RATED THERMAL POWER (F RTP) as specified in the COLR per Specifi-l yy cation 6.8.1.6 was determined from expected power control manuevers over the full range of burnup conditions in the core.

3/,4.2.4 QUADRANT POWER TILT RATIO

/

^

The QUADRANT POWER TILT RATIO limit assces that the radial power distribu-tion satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts.

A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing q

the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.

The two sets of four symmetric thimbles is a unique set of eight detector locations.

These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DNS PARAMETERS The limits on the DNB51ated parameters assure that each of the parameters are maintained within the nyrmal steady-state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.

The indicated T,yg value of 591'F and the indicated pressurizer pressure value of 2224 psig correspond to analytical limits of 592.5'F and 2205 psig respec-tively, with allowance for measurement ur. certainty, V0GTLE UNITS - 1 & 2 B 3/4 2-5 Amendment No. 32 (Unit 1)

Amendment No.12 (Unit 2)

POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS (Continued)

The 12-hour periodic surveillance of these parameters through instt ument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of the flow rate degradation on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

A change in indicated percent flow which is greater than the instrument channel inaccuracies and parallax errors is an appropriate indication of RCS flow degradation.

6 9

V0GTLE UNITS - 1 & 2 B 3/4 2-6

~

.' ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)

The Semiannual Radioactive Effluent Release Reports shall also include the following:

an explanation as to why the inoperability of liquid or paseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.9 or 3.3.3.10, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.

MONTHLY OPERATING REPORTS

6. 8.1. 5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a

~'

copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT

6. 8.1. 6 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

SHUTDOWN MARGIN LIMIT FOR MODES 1 and 2 for Specification 3/4.1.1.1, j

a.

b.

SHUTDOWN MARGIN LIMITS FOR MODES 3, 4 and 5 for Specification 3/4.1.1.2, Moderator temperature coefficient BOL and EOL limits and 300 ppm c.

surveillance limit for Specification 3/4.1.1.3, d.

Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, 1

e.

Control Rod Insertion Limits for Specification 3/4.1.3.6, f.

Axial Flux Difference Limits, and target band for Specification 3/4.2.1, g.

Heat Flux Hot Channel Factor, K(Z), the Power Factor Multiplier and F,R for Specification 3/4.2.2, h.

Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be those previously approved by the NRC in:

V0GTLE UNITS - 1 & 2 6-21 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

1

. ADM NISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY",

a.

July 1985 (W Proprietary).

{

(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coef ficient, 3.1.3.5 - Shutdown Bank Insertion Limit 3.1.3.6 -

Control Bank Insertion Limits, 3.2.1 - Axial Flux Dif ference, 3.2.2

- Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise i

Hot Channel Factor.)

b.

WCAP-8385, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES

- TOPICAL RFPORT", September 1974 (W Proprietary).

j (Methodology for Specification 3.2.1 - Axial Flux Difference

[ Constant Axial Offset Control).)

T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC) c.

January 31, 1980--

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package.

(Methodology for Specification 3.2.1 - Axial Flux Difference

[ Constant Axial Offset Control).)

d.

NUREG-0800 Standard Review Plan, U. S. Nuclear Regulatory I

Commission, Section 4.3, Nuclear Design July 1981.

Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981.

(Methodology for Specification 3.2.1 - Axial Flux Difference a

[ Constant Axial Offset Control).)

e.

WCAP-9220-P-A, Rev. 1, " WESTINGHOUSE ECCS EVALUATION MODEL-1981 VERSION", February 1982 (W Proprietary).

1 (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

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

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

SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

i L

V0GTLE UNITS - 1 & 2 6-21a Amendment No. 32 (Unit 1) l Amendment No.12 (Unit 2) 1

i

,' ADMINIST @ IVE CONTROLS 6.9 RECORD RETENTION 6.9.1 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.

l 6.9.2 The following records shall be retained for at least 5 years:

a.

Records and logs of plant operation covering time interval at each power level; b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety; c.

All REPORTABLE EVENTS; I

~

I d.

Records of surveillance activities, inspections, and calibrttions required by these Technical Specifications; Records of changes made to the procedures required by e.

l Specification 6.7.1; I

l f.

Records of radioactive shipments g.

Records of sealed source and fission detector leak tests and results; and i

l h.

Records of annual physical inventory of all sealed source material of record.

6.9.3 The following records shall be retained for the duratio.i of the plant l

Operating License:

1 a.

Records and drawing changes reflecting plant design modifications made

{

to systems and equipment described in the Final Safety Analysis Report; l

b.

Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories; c.

Records of radiation exposure for all individuals entering radiation control areas; d.

Records of gaseous and liquid radioactive material released to the environs; e.

Records of transient or operational cycles for those plant components identified in Table 5.7-1; f.

Records of reactor tests and experiments; i

g.

Records of training and qualification for current members of the plant staff; V0GTLE UNITS - 1 & 2 6-22 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

.-.

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