Text

Proposed Tech Specs Removing cycle-specific Limits & Placing in Separate Core Operating Limits Rept
	ML20012E013
Person / Time
Site:	Vogtle
Issue date:	03/22/1990
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20012E012	List: ML20012E011; ML20012E013; ML20012E014; ML20012E015;
References
	NUDOCS 9003290213
	Download: ML20012E013 (75)
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y DEFINIT!0NST 7 0FFSITE DOSE CALCULATION MANUAL. 1-1.kTheOFFSITEDOSECALCULATIONMANUAL(ODCM)shall.containthemethodologyz and parameters:usedLin the calculation of offsite doses-due to radioactive-gaseous and liquid effluents, in the calculation of gaseous and liquid-effluent monitoring Alam/ Trip Setpoints, and in the conduct of the Environ- l

. : mental Radiological Monitoring Program. i-

' OPERABLE - OPERABILITY' q I M A' system, subsystem, train, component or device shall be OPERABLE or .

.have 0PERABILITY 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 perfoming their related support function (s). y 1 OPERATIONAL MODE - MODE

1. N An OPERATIONAL' MODE (1.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average-reactor ,

coolant' temperature;specified in Table 1.2.

PHYSICS TESTS l

-1. PHYSICS TESTS shall be tho.se tests perfonned to measure the fundamental 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

.13 -

1. E ' PRESSURE-BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System ccaponent body, pipe wall, or vessel wall.

L PROCESS CONTROL PROGRAM

2. */

1.. W The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made:to ensure that processing 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 regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste.

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PURGE - PURGING 1 PURGE-or PURGING shall be.any controlled process of discharging air or gas l from a confinement to maintain temperature, pressure, humidity, concentration I or other ' operating condition, in such a manner that replacement air or gas is required to purify the confinement.

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INDEX + -.

DEFINITIONS 3 t '

1 SECTION' , PAGE/ i l'.0- DEFINITIONS )

. - 1.11 ACTION........................................................ 1-1~

1.21 ACTUATION LOGIC TEST..........................................- 1-1 1.3' ANALOG CHANNEL OPERATIONAL TEST............................... 1-1: r

- 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' i m' . .

,(

.1.8 ' CONTROLLED LEAKAGE............................................. 11-2 ;

1.9 CORE'AL ERN IIONS.,........kEIW...............................

.. 1-2 to ConEO FR+ T o ve- } t M 175 I IT ,. ,

I-A >

1. gDOSE'E UIVALENT:1-131.......................................

1,yJ ... 1 >

1.101E-AVERAGE:DISINTEGRATIONENERGY.............................. 1-2= ,

IMe3 ENGINEERED ~ SAFETY FEATURES RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . 1-3 -

1Mi d F REQUEN_CY NOT ATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

'l y trGASEOUS WASTE PROCESSING SYSTEM.............................. 1-3

- l'. IN4 I D E NT I F I E D L E AKAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . 1-3.

1M/ 7 MA ST E.R - R E LAY T E ST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . .

l-3 1

- 1.-1TItMEMBER(S) 0F THE PUBLIC...................................... 1-3

1. )st4 0FFSITE DOSE _ CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-4' .

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- 1.PD00PERABLE - OPERABILITY........................................ 1-4 1 1

- 1.76 80PERATIONAL MODE - M0DE...................................... 1-4

1. gn PHY S I C S T E ST S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.yb! PRESSURE BOUND ARY LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4

l 1.NfM PROCESS CONTROL PROGRAM. .................................... 1-4' L' 1.2p rPURGE - PURGING............................-..................

1-4

. 1. 54 QUADRANT POWE R TI LT RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 i l

l- I E.URATED THERMAL P0WER...................... ................... 1-5 o

1.7/1FREACTOR TRIP SYSTEM RESPONSE TIME............................ 1-5 L

1.3tf1fREPORTABLE EVENT............................................. 1-5

1. 7,00 S HUT DOWN MA RG I N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.263/SITEB0VNDARY................................................ 1-5

1. M31 SLAVE RELAY TEST........................ .................... 1-5 V0GTLE' UNITS - 1 & 2 I l.

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a - DEFINITIONS O I.SECTIONL PAGE. ,

4 71;)d3 50LI DI F I CAT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . =1-5 l; 1 - )!Pf SOURC E CHE C K ; . . . . . . . . . . . . . . . . . . 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 j? 1-6 i '

j I M Jf5TAGGERED TEST BASIS..............................-........... .

~1.ptfS$THE RMA L P0WE R . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . -. . . . . . . . . . . 6- ll 1.MI37 TRIP' ACTUATING DEVICE OPERATIONAL TEST...... ................ . 1 f 1;)'[#UN I DENT I F I ED LE AKAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . -. . . . . . 1-6 l

1.MNUNRESTRICTEDAREA...........................=...............'... 1-6 l 1-6 !

1.M? VENTI LATION EXHAUST- TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . .

1.MI4/ VENTING...................................................... 1-7

. TABLE 1.1

FREQUENCY N0TATION. .................................... 1-8 .,

1 TABLE ?L 2T OPERATIONAL M0 DES........................-.-..............- 1- 9' j

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.V0GTLE UNITS - 1 & 2 II

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' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABillTY............................................... 3/4 0 1

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3/4.1 REACTIVITY CONTROL SYSTEMS

~3/4.1.1- BORATION CONTROL

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. Moderator Temperature Coef f icient. . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-4 Minimum Temperature for Criticality...................... 3/41-6

-3/4.1.2 BORATION SYSTEMS F l ow P a t h - Shut down . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-7 Flow Paths'- Operating................................... 3/4 1-8 C ha rgi ng Pump - Shut down. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 W 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height............................................. 3/4 1*14 14LE3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE

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EVENT OF AN INOPERABLE CONTROL OR SHUTDOWN ROD........... 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 Limit 5............................. 3/4 1-21 V0GTLE UNITS - 1 & 2 IV

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS .

1 PAGE i SECTION I 7:0"" 0.1-3 ^^0 " '.?' N!!"T ! 0" L : ": T S "f "0"5 M ""' t . ^"' R . . . . . . . 3/4 1-22 l

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 FIGURE 3.2-1 AX:AL FL"X O!FF:^iN"" L "

. = == =. . . . ?u:.wTE "i) A T"""T:0" 0F1.d ..................... 3/4 2 3 ,

3/4.2.2 g i g 01 CHANNEL FACTOR - F g (Z)..................... 3/4 2-4 l FIGURE 3.2P ".(?) - 9"^1??ED F (2) ^.5

F""CTIO" 0F C^*E "E!0"T, 3/4 2-5 i 3/4.2.3 NUCLEAR ENTHALPY RISE H$T CHANNEL F ACTOR -3/4 2-8. . . . . . . . . . l Ffg.

- 3/4.2.4 QUADRANT POWEh TILT RATI0................................ 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 4.3-3 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE '

REQUIREMENTS............................................. 3/4 3-9 ,

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM I N ST R UM E NT AT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3 15 .

TABLE 3.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-17 TABLE 3.3-3 ENGINEERED SAFEfY 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 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-45

- I TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS..................................... 3/4 3-46

, TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT L OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3 48 Movable Incore Detectors................................. 3/4 3-49 Seismic Instrumentation (Common System).................. 3/4 3-50 L

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. V0GTLE UNITS - 1 & 2 V l

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' i ADMINISTRATIVE CONTROLS i f

SECTION ,

6.4.2 SAFETY REVIEW BOARD (SRB) l Funct;on.................................................. 6-9 !

Composition............................................... 6-10 l 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..................................... 6-13 6.6 SAFETY LIMIT V10LAT10N...................................... 6-13 ,

i 6.7 PROCEDURES AND PR0 GRAMS..................................... 6-13 6.8 REPORTING REQUIREMENTS ,

6.8.1 ROUTINE REP 0RTS........................................... 6-17 l, Startup Report............................................ 6-17 Annual Reports............................................. 6-17 l Annual Radiological Environmental Surveillance Report..... 6-18 Semiannual Radioactive Effluent Release Report............ 6-19 i Operating Reports................................. 6-21 C 0Rt C 7tWe t.144t1 g CRT Monthly)fc ESiel %!h . M'Fg F00t07 L]a..g[ n it........................ 6*21 6.8.2 SPECIAL REP 0RTS........................................... 6-21 6.9 RECORD RETENTION............................................ 6-22 t

e .1 V0GTLE UNITS - 1 & 2 XXIII 1

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.- L DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when: i

a. All penetrations required to be closed during accident conditions are either: i 4 j

1) Capable of being closed by an OPERABLE containment automatic 1 isolation valve system, or l 1

Closed by manual valves, blind flanges, or deactivated automatic

' Q 2) valves secured in their closed positions.

k b. All equipment hatches are closed and sealed, d) c. Each air lock is in :ompliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and

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b e. The sealing mechanism associated with each penetration (e.g., welds, T bellows, or 0-rings) is OPERABLE.

q CONTROLLED LEAKAGE 4

h 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor cociant pump seals.

CORE ALTERATIONS M 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

's !

l 3 the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of , ,

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movement of a component to a safe conservative position. -

DOSE EQUIVALENT I-131 1.y(dDOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram)

V which alone would produce the same thyroid dose as the quantity and isotopic

. mixture of I-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid , .

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

I - AVERAGE DISINTEGRATION ENERGY 11 )

f 1.E I shall be the average, weighted in proportion to the concentration of-I 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. .

V0GTLE UNITS - 1 & 2 1-2 ,

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@ RE 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 cycle in accordance with Specification 6.8.1.6. Unit operation within these operating j limits is addressed in individual specifications, j

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DEFINITIONS-a ENGINEEREDSAFETYFEATURESRESPONSETIM{

1.kTheENGINEEREDSAFETYFEATURES(ESF)RESPONSETIMEshallbethattim interval free when the monitored parameter exceeds its ESF Actuation Setpoint l,

at the channel sensor until the ESF equipment is capable of. performing its ,

safety function (i.e., the valves. travel to their required positions pump .

dischargepressuresreach,theirrequiredvalues,etc.).TimesshallInclude !

diesel generator starting and sequence loading delays where applicable. ;

FREQUENCY NOTATION

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The FREQUENCY NOTATION specified for the perfomance of Surveillance !

1.

Requirements shall correspond to the intervals defined in Table 1.1. j GASE0US WASTE PROCESSING SYSTEM s stem designed and installed 1.[AGASEOUSWASTEPROCESSINGSYSTEMshallbeafng to reduce radioactive gaseous offluents by collect eactor 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, j IDENTIFIED LEAKAGE 1.f!DENTIFIEDLEAKAGEshallbe:

a. - Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump ,

seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or j

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the -!

L r 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 .

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

MEMBER (S) 0F THE PUBLIC IV i

1. W 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-are persons who enter the site to service equipment or to make deliveries.

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

.V0GTLE UNITS - 1 & 2 1-3 l

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DEFINITIONS QUADRANT POWER TILT RATIO 1 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore-detector calibrated output to the average of the upper excore detector cali- !

brated outputs, or the ratio of the max' aum 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 shall be used for computing the average. l, RATED THERMAL POWER l 27 -

1.26' RATED THERMAL POWER shall be a total reactor core heat transfer rate to ;

the reactor coolant of 3411 Wt.

REACTOR TRIP SYSTEM RESPONSE TIME i

IMThe REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from ~

when the monitored parameter exceeds its Trip Setpoint at the channel sensor

until loss of stationary gripper coil voltage. .

REPORTABLE EVENT I M A REPORTABLE EVENT shall be any of those conditions specified in Sections 50.72 and 50.73 of 10 CFR Part 50. 1 SHUTDOWN MARGIN l

1. h SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which '

l the reactor is suberitical or would be suberitical 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.

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SITE BOUNDARY l

1. h The SITE BOUNDARY shall be the exclusion boundary line as shown in

. Figure 5.1-1. ,

SLAVE RELAY TEST 1.hASLAVERELAYTESTshallbetheenergizationofeachslaverelayand

, verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include I a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION

, IN SOLIDIFICATION shall M the conversion of wet wastes into a form that meets shipping and burial ground requirements.

V0GTLE UNITS - 1 & 2 1-5 0_________--_____-_________ _ _ _ _ _ _

DEFINITIONS 3 SOURCE CNECK I M A SOURCE CHECK shall be the qualitative assessment of channel response t when the channel sensor is exposed to a source of increased radioactivity. ,

STAGGERED TEST BASIS I M A STAGGERED TEST BASIS shall consist of: ;

a. A test schedule for.n systems, subsystems, trains, or other i designated components obtained by dividing the specified test interval into n equal subintervals, and 1

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

THERMAL POWER ,

4 1.E THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. >

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.hATRIPACTUATINGDEVICEOPERATIONALTESTshallconsistofoperatin :

Trip' Actuating Device and verifying OPERABILITY of alarm, interlock or and/g the :

trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include i adjustment, es necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

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l UNIDENTIFIED LEAKAGE 1 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE

> or CONTROLLED LEAKAGE. ,

UNRESTRICTED AREA 1 h An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY ,

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.

VENTILATION EXHAUST TREATMENT SYSTEM 1.h A VENTILATION EXHAUST TREA1NENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate .)

l 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-

l. 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 reatures Atmospheric Cleanup Systems are not considered "

to be VENTILATION EXHAUST TREATMENT SYSTEM components. ,

.V0GTLE UNITS - 1 & 2 1-6

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i DEFINITIONS i

~t r VENTING 1, k VENTING shall be the controlled process of discharging air er gas from a l 1

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 -

l a VENTING process. .

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.V0GTLE UNITS - 1 & 2 1-7

i 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SORATION CONTROL

$HUTDOWN MARGIN - MDDES 1 AND 2 LIMITING CONDITION FOR OPERATION j

+helbntYsycc.fkd 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equ 1 to 1..T. .: L/h.

s n +1,e CCRF OffM TIA!&- l-l A1175 R R PC R T (C O L G(), ;

APPLICABILITY: MODES 1 and 2*.

ACTION: .

& lk,Y . Sted Sie [ In th e CW- A With the SHUTDOWN MARGIN less than 1 7, eL/L, immediately initiate and continue l 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.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1. iE '" t h e i e'm lT specjfied io the CCM.

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, th2 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 withdrawai is within the limits of Specification 3.1.3.6;

c. 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 criticality by verifying that the predicted critical control rod -

position is within the limits of Specification 3.1.3.6; and ,

I

d. Prior to initial operation above 5% RATED THERMAL POWER after each i fuel loading, by consideration of the factors below, with the control banks at the maximum insertion limit of Specification 3.1.3.6:

See Special Test Exceptions Specification 3.10.1.

V0GTLE UNITS - 1 & 2 3/4 1-1 a: . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _ _ _ - _ - _ _ _ . _ _

i D

.m. -

2; l

l REACTIVITY CONTROL SYSTEMS ;

- }HtITDOWN MARGIN - MODES 3. 4 AND 5 LIMITING CONDITION FOR OPERATION f l

. 3.1.1.2- The SHUTDOWN MARGIN shall be greater than or equal to the limits shown- :

' da et;g.33 3.1 13 -g 3,3 3: (Lin gt :) : : ei;;7;; ,;-;; ;ng 3,3-;; ,

'" nit 2). S tecofred un +k, con opryrW& LIMITJ AEfoAT (Col.n) .

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.

SURVEILLANCE REQUIREMENTS t 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 untrippable, 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 L 6) Samarium concentration.

\

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

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l l

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(M/M V %) NIOWW NMOQinHS ymu t+ns , t m ,-n

r = _%.. -.a &-aw-ek---e eda 24 4e.4 m -4e.- daA--daw.ams.J e- =aE-.ae--a=a-Mh-DE'deln.4C ia nA,me----SM--J----v** *-,.Fu 4.iem-M 4 e<--a-w me---

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(M V%) NIDWVM M001ANS l

l YMTLE '3!!! !?2 0/0 1-;;

w % #c ks,,dy a col, lite (not) !l,,,Y l o nd +hr faJ or Cyel hfe (fet) l ' mil l REACTlvlTY CONTROL SYSTEMS specihed jo, the Con ofsgArap& 4/Ae/TF RFfoR T

/ /* I''"

MODERATOR TEMPERATURE COEF CIEN d' #/ de ! ,

i t LIMITING CONDITION FOR OPERATION 3.1.1.3 Themoderatortemperatirecoefficient(NTC)shallbe/ }

p' Unit 1: i Less positive than +0.7 x 10 4 Ak/k/*F '- "- ' "-J'"---

(Set),::nditi;Uofjo;y'jjyiliu'p't5"

-50;f- ' ; c' ey !: " f:

70% RATED THERMAL POWER with a linear ramp'to 0 ok/k/*F at 100% l RATED THERMAL POWER; and 1 Unit 2:

Less positive than 0 ok/k/*F,f:r th: :" :d: !thd = n, i;;ir d ;

c' eye'e e (SOL), het ::r: '"!'""L =^-^ _ diti;n, ;nd b! N 5' i ! I , $$ Yek;k "nkhen N'el b [ 5eb'I_ N ).'.Ib "" l

. . _ . . . , . . . .. s...,, ... . . . _ . . . . _ . . . . . . . . . . . .

60t- ls'o'slY .

APPLICABILIT_Y: Sp::i'i::ti:r 3.1.1.30. - MODES 1 and 2* only **

S;^:i'iratie- 3.1. 2. 35. - MODES 1, 2, and 3 only.**

2(A- /Imo'f ,

ACTION: ,

, BN' gecs(Od in Ae COL A

a. With the MTC more positive than the A limit +f-4pc:f't::ti:n 3.1.1.30.

b;ve, operation in MODES I and 2 may pro: eed provided: . .

e Get /mit ognfdd on f4: Cet R

1. Centrol rod withdrawal limits are e ished and maintained !

sufficient to restore the MTC to withi th: it::: li:!t 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

,___.,,__y.. ,

--- _.__..-9 .,,. , - . - . - , . ,_

t REACTIVITY CONTROL SYSTEMS ,

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.

to L. qec.hjed an +ke CQL R V the limit e 4;;ifi;etie. 3.1.1.0t.

l

b. With the MTC more negative than ebeve, 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: nei,fied in tb C C/.R. -

a. The MTC shall be measured and compared to the BOL limit Ef ;;;i

red e- ?.1.2.32., de'fe, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and

b. The MTC shall be measured at any THERMAL POWER and compared to 3.1 10 ' at/i/*f (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concen- 4 tration of 300 ppm. In the event this comparison indicates the MTC is more negative than 0.1 10 ' it/k/*f, the MTC shall be remeasured, and comp to the EOL MTC limi 3.1.1.05., at le onceper14EFPDduring[t-f;::!'i: remainder oftie- the fuel cycle. -

gegged in 4Ae (0/-N

-t he no pp m s a rve illa n c.e IthN spec lfsd m +4r Cot-R ,

V0GTLE UNITS - 1 & 2 3/4-1-5

. - -~ -- . - - - . - . - - - . . _ .

..- j l

)

i REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ,

GROUP HEIGHT LIMITING CONDITION FOR OPERATION I

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

\

APPLICABILITY: MODES 1* and 2*.

l 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 untrippable, determine that the SHUTDOWN MARGIN requirement of ;

Specification 3.1.1.1 is satisfied within I hour and be in HOT t STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . i t

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 height by more than 1 12 steps (indicated position), ,

POWER OPERATION may continue provided that within 1 hour: '

1. The rod is restored to OPERABLE status within the above alignment requirements, or .

g. 6 ;

2. The rod is declared inoperable nd the remainder of the rods in the group with the inoperabl rod are aligned to within i 12 steps of the inoperable rod whil aintaining the rod sequence and insertion limits ;f N ; .; 0.'. 2. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or ,

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER >

OPERATION may then continue provided that; a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; 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 UNIIS - 1 & 2 3/4 1-14

7 l

i REACTIVITY CONTROL SYSTEMS i i LIMITING CONDITION FOR OPERATION ;

A,QTION(Continued) ,

, c) A power distribution map is obtained from the movable !

incoredetectorsandF(Z)andF[H q are verified to be j within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and l 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 causes other j than addressed by ACTION a above, power operation may continue heuficakten J. l. 3. 6 i

1. Within I hour, the remainder of the rods in the bank (s) with the !

inoperable rod are aligned to within i 12 steps of the inoper- !

able rods whil maintaining the rod sequence and insertion limits of fi;;.n 0.13. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and

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 . ,

d. With more than one rod misaligned from its group sten counter demand height by more than i 12 steps (indicated position), be in HOT '

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

t SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each rod shall be determined to be within the group demand limit by verifying the individual rod positions 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 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 , j j

4.1.3.1.2 Each rod not fully inserted in the core shall be determined to be 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 1

1;:

l t

q. ;

REACTIVITY CONTROL $YSTEMS j

$NUTDOWN R0D IN$ERTION LIMIT I LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be "'"- -- ^^

^^"'" ;~;'t"- *'" "

i

? '- sa limlie d on pays,'ca l inerh'or, es spcifie d on +A

ccite otreArave Limors Arrear ccesia) 1 ApPLICAt!LITY: MODES la and 2* y :

ACTION:

j4 4, f,;jf gg ;

With a maximum of one shutdown rod inserted O 5$$br.10 : th-. **? c'- l t

except for surveillance testing pursuant to Speci;fication 4.1.3.1.2, within; ,

I. .

I hour Rufere either: de ec) lc d/lih 4}e Iaxeloo'., IkoYyx}fied tis $ W % w

4. "'thdrz th: 70 ' ' 0 ;::'t' r. ;r::t:r th:r r ::: 1 '- *** O' ;:, Or l

b. Declare the rod to be inoperable and apply Specification 3.1. 3.1. i l '

! _$URVEILLANCE REQUIREMENTS I t

.: w;W +Ae onser ho n t;> o't s7t0hh

':. in fkne coe. A :

'I 4.1.3.5 Each shutdown rod shall be determined to be "'* " - - '- - --- "' -

r: :ter th-- ^- ^^ut! '^ *** t:;::

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.

i l

See special Test Exceptions specifications 3.10.2 and 3.10.3.

0 With K,ff greater than or equal to 1.

V0GTLE UNITS - 1 & 2 3/4 1 20 Amendment No.29 (Unit 1)

Amendment No.10 (Unit 2)

- i REACTIVITY CONTROL SYSTEMS l CONTROL ROD INSERTION LIMITS f

LIMITING CONDITION FOR OPERATION J

3.1.3.6 The control banks shall be limited in physical insertion as d:r ' :

' Y,= :.1 :. Specife'e d in de CORR oggnies timin AEMnr(coon)

APPLICABILITY: MODES 1* and 2* #. l ACTION:

With the control banks inserted beyond the e r: insertion limits, except for i surveillance testing pursuant to Specification 4.1.3.1.2: ;

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 abar " - , r- Inserb /ibMr Jf ec.!/,;/ ,' -t/c cot g er j

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 .

l l ,

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

ij i

' <m.,m o.., i

. NmM4 i If I ,

1 /- y 1 m\ E E

l

\ / sANKB

/

/ I

/

/I E 1

I E

,, E A I I / I / I E E Il N 1 / / I E E/ I N I / / tioo.ies E

,, E co. isi) \l / /I i /E

' 3, E I

\ / sANKC/ I l/ ]

IN/ I/l 1

/ E-S E

E' E

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/ IV I

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/

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I/ I E i / I / \l / I E

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I I 1

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ll l ,, E / / I II i\ i E l-!

" F io.4ei / - i i / ' '

Ni i E

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_,h :

E ll E /i . I / o \ E r

R/ I i / i ! i ! \ E

<no.. a or/ I I i <

ummmmmmmhmmmmmmmmmmmmmmuum o 20 40 so so 1 RELATIVE POWER (pwoont) l I

[ IIOUR 2.1-2 I ":: !?= :N::ni::'; L:M:i: v:n::: i;;:=; t0;;;r.

v Y

fo n : .r:Ts ::: 3." : :: n . c, r.; n. :: (vat u v we ry y , _ ,..

ggy e.y vvy i n i n.. . es

-v 3Wy VW W f g g g- S C - T d - N d M O _ p yg g g n m yg y ' I: 90 Os ate 'E O

t

., t i

3/4.2 POWER DISTRIBUTION LIMITS t

3/4.2.1 AXIAL FLUX DIFFERENCE i LIMITING CM DITION FOR OPERATION ,

3.2.1 The indicated (NI-00418, N1-00428, NI-00438, NI-00448) AX1AL FLUX i DIFFERENCE (AFD) shall be maintained within the ' " r'; target band (flux ,

l about the targe flux d ffer difference unitf)91'f*.f!*!M .).!!'F..pCff&gnceFTAc %pf Jm/ n .rm,MJ

~

8a f4f(OM87 t_.,u.. u.. ..__...,m

!^^^ S/"TL'; ;r.d

-E ^ 5, 1 % ';r ;^r: ; ; ;;; :::r . , i:d L77 r ^' er;;t;r ^.h.o ^000 .

MlM 4esMe/ m de Cot.g i The indicated AFD may deviate outside the above required target and at greater :

than or equal to 50% but less than 90% of RATED THERMAL POWER rovided the indi- f cated AFD is within the Acceptable Operation Limits :' ff;; . 0.01 and the cumu-

lative 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 obove required target band at greater than 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 . ,

APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.*

ACTION:

a. With the indicated AFD outside of the ;t:x required target band and i with-THERMAL POWER greater than or equal to 90% of RATED THERMAL -

l 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,

b. With the indicated AFD outside of the :t: x required target band for more 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 or outside the Acceptable Operation Limits.e4.Specified m.

14e COL & Ti;;r; 2.0-1 and with THERMAL POWER less than 90% but equal 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 i 2. The Power Range Neutron Flux * - High Setpoints to less than or L equal to $5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l

! *See Special Test Exceptions Specification 3.10.2.

Surveillance testing of the Power Range Neutron Flux Channel any 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 ++ & ceifer'/

jo 4Ae C044rt;.7; 3. -1. 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

.g .w- , - - - .p., , y-pp., y,--,.,-- -, -+sgy,-,---W-. e-gy.--m-,- _,,gwvw-. v

r

?

l POWER DISTRIBUTION LIMITS ;

L LIMITING CONDITION FOR OPERATION ACTION (Continued) .

l

c. With the indicated AFD outside of theweben required target band for more than I hour of cumulative penalty deviation time during the lI previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less 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.eb m required target band and the l cumulative penalty deviation has been reduced to less than I hour in i 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. W itoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, ,

and f

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 exist during the interval preceding each logging. ;

1 4.2.1.2 The indicated AFD shall be coisidered outside of its tLrget band when two or more OPERABLE-excore thannels are indicating the AFD to be outside the target band. Penalty deviation outside of the-+bave required target band shall I ;

be accumulated on a time basis of:

a. One minute penalty deviation for each I minute of POWER OPERATION i outside of the target band at THERMAL POWER levels equs1 to or above 50% of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each 1 minute of POWER OPERATION I l 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 least 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 Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the 1 most recently measured value and 0% at the end of the cycle life. The provi- I sions of Specification 4.0.4 are not npplicable. !

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

. - - . ..- _ _ . _ _ - . . . . - . .~. - . .. -. . - _ - .

e

/ !

I k

l I

\ / l tm

\ / !

1

\ T ussancertant.

OPtftATICII l

( 11, e0) (11,e0) 1 i

\ .

s se l / / T :

l - e. . ..

l o

( 31, so) (31, so) 3 m

, / \

/ \

?

9 e

-se -es -se -se -te e to se se se se FLUX DIFFERE9CE, Al (pumant)

FIGURE 3.2-1 [ DEL 6 TE D)

Ax: AL rLux 0 rrt^txt: L:":T; A: A re";TIO" Or Pt.TED THEP".". P05'ER V0GTLE UNITS - 1 & 2 3/4 2-3

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

[K(Z)] for P > 0.5 WA'*' t F,#f de F, /sief FA (Z) ~< 'T~~ a} MA190 THEMMAL u Fg (Z) < 60- ( K(Z)] f or P < 0. 5 Para (47h3cuyie/en

/

L, ~ _A ~

}ftconi c!!AATIM o, r , and t.im r s arps47 Whe re *- P = THERMAL POWER RATED THERMAL POWER (cvA4);

K(Z) = the norulosed F m as a resd*n *f fur.; tier, ;.4

,t:ir.:d fret. Figur: 3.2-2 for : gi :r core height lecetfea. jf.ar/c/ ,,, v4 coe A.

APPLICABILITY: MODE 1.

ACTION:

With Fq (2) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds 9

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 ; subsequent POWER OPERATION may proceed provided the :'

Overpower AT Trip Setpoints have been reduced at least 1%

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

b. Identify and correct the cause of the out-of-limit condition ,

prior to increasing THERMAL POWER above the reduced limit re-p quired by ACTION a,, above; THERMAL POWER may then be increased

provided Fq(Z) is demonstrated through incore mapping to De '

within its limit.

i L

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

i- a l

. l l

l l

I l g 1

i s

1.5 '

1.00 , ,

I E 0.M s w

gg OORE '

ME60HT K(Z)

O.000 1, 0.000 10235 .030 ,

0.5 X 1L000 052 0 / 2.0 0.0 8.0 10.0 1't.

0 4.0 CORE HEIGHT m)

FIGURE 3.2-2

(~ 0 E t 6 TE D)

X(2) - NOP",AL12E0 FQ (2) AS A T' NCTION J OF 00Pi "E!OMT V0GTLE Uh!TS - 1 & 2 3/4 2-5

o e

POWER DISTRIBUTION LIMITS k

SURVEILLANCE REQUIREMENTS i

4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

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

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER

, before exceeding 75% of RATED THERMAL POWER following each fuel loading,

b. Increasing the measured F,y component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by ET tn account for measurement uncertainties,

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

above to:

1) The F limits for RATED THERMAL POWERRT(F ,P) for the appropriate xy measured core planes given in Specification 4.2.2.2e. and f.,

below, and 2)' The relationship: pg pg y 3 gje p,,, 4,/,r y L ,p RTP g3 (3,py3* ( uu,lhjlire se hy speco%'t/ ,

xy ' xy in + lie c o t %

Where F l is the limit for fractional THERMAL POWER operation I

Pd expressed as a function of F is the fraction of RATED THERMAL POWER at which F was measured.

yy

d. Remeasuring F xy according to the following schedule:

1) When F is greater than the F N limit for the appropriate measured core plane but less than the F relationship, additional LowerdistributionmapsshallbetakenafdF C compared to F RTP x yy and F either:

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

POWER or greater, the THERMAL POWER at which F*# ,,, ),,g determined, or .

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

whichever occurs first.

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

'4 s

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

2) When the F x is less than or equal to the F RTP limit for tM appropriate measured core plane, additional power distribution P

maps shall be taken and F x compared to F and F x at least once per 31 EFPD.

e. The F limits used in the Constant Axial Offset Control analysis for RA ED THERMAL POWER (F P)shallb[fi$kE:Iforallcoreplanes containing Bank "D" control rods and all unrodded core planes in-e-

-the cola-Aadi:1 P::k%g F cter Limit P.: pert per Specification 6.81.6;

f. The F I '"'I t s f Specification 4.2.2.2e., above, are not applicable xy in the following core planes regions as measured in percent of core height from the bottom of the fuel:

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

2) Upper core region from 85 to 100%, inclusive,

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 regions within 1 2% of core height [: ? 88 inches]

about the bank demand position of the Bank "D" control rods.

g. With F x exceeding F x the effects of F,y on F9 (Z) shall be evaluated to determine if F 9 (Z) is within its limits.

4.2.2.3 When Fg (Z) is measured for other than F determinations, an overall measured gF (Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

V0GTLE UNITS - 1 & 2 3/42-7 l

y. ;

e- j u ,, , 1 J

POWER DISTRIBUTION LIMITS- ..

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR-F g LIMIT]NG CONDITION FOR OPERATION l

N 3.2.3 F.

AHsga,1)belimitedbythefollowingrelationship:

Fu % g y bm,s a f IVA T F 8 N

F3g 1 ++6- [1 + +.+(1-P)] [f,r.4 H :: T Ac Fu '

T HER ML POWfA (A Y P) St r$4 E Wherej= &

THERMAL POWER

### I ##D l ~

p

M 8#4 T (C O A S) ,

RATED THERMAL POWER APPLICABILITY: MODE 1. /

f fH b

= s 9, w .,. F u h e A l Q l 4 .-

H ACTION: '

N for Yl8 Sr co M in N C04 A W$ #

With F 3g exceeding its limit:

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

1. Restore F g to within the above limit, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip 5etpoint 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 throughincorefluxmappingthat'Ffg has been restored to within 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 l c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit >

l.' required by ACTION a.2. and/or b. , above; subsequent POWER OPERATION l

may proceed provided that F g is demonstrated, through incore flux mapping to be within its limit prior to exceeding the following THEMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and i

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

m

. .7 3/4.1 REACTIVITY CONTROL SYSTEMS

~ BASES

-t 3/4.1.1 BORATION CONTROL b ' 3/4.1.1.1 and 3/4.1.1.2' SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients asso- r ciated with postulated accident conditions are controllable within acceptable ;

y limits, and (3) the' reactor will be maintained sufficiently suberitical to' J preclude total loss of SHUTDOWN MARGIN in the shutdown condition.

. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg. In MODES I 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 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 M etted er a functier of 905 berer

cntr;ti

7 spcifi'ed on +he CORS CWflATING-l-I A I TS AFAORT (COMf 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT .

The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses, i

( 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.

'. 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. Th'ese corrections i

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

.~- _. . . _ . .

.. I 1

REACTIVITY CONTROL SYSTEMS l BASES I

' 8CDERATOR 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 soderator donsity with temperature at RATED THERMAL PE ER conditions. This value of the IOC was then transformed into the limitin TRTC value c '.0 10 ' ; U J^i. The% UC value M -3,1 : 10 ' ih/k/*T 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 MTC value,ef 4,0 10 ' ok/k/*i.

h '

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.

L p 3/4.1-1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made ca al with the Reactor Coolant System average temperature less than 551" 's limitation is required to ensure: (1) the moderator temperature cit t is within its analyzed temperature range (2) the trip instrument a .ithin its normal operating range, (3) the pressurizer is capable of being *c ,

OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT NDT temperature.

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 1

an assumed single failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN '

I' 1

l' V0GTLE UNITS - 1 & 2 B 3/4 1-2 e

p v

b V.

r !b:

3/4.'2 POWER DISTRIodi10N LIMITS

(

BASES

.+

L 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, limitin~g the peak linear power density during Condition I events provides assurance that the.

initial conditions assumed for the LOCA ana' lyses 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:

F9 (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat 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; Ffg Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio .of the integral of linear power along the rod with the highest integrated power to the average rod power; and Fxy(Z)

Radial Peaking Factor, is defined as the ratio of peak power density J l

to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX OIFFERENCE the Q /Ad siwGd in Ne COM ONR47/4Te Al*!rs itsknr (coin)

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

l bound envelope during either of -G-&9 normal' dimes operation the or in n:r::li:

the eventd of:::[i:1 xenonp:: redistribution kins, f::t:r is not exceeded following l

power changes. L g(z)

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 The periodic updating of the target flux fractional THERMAL POWER level.

difference value is necessary to reflect core burnup considerations.

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

n. .

Ljp ,

a.

POWER DISTRIBUTION LIMITS L

t ,

in BASES 41 1

AXIAL FLUX DIFFERENCE (Continued) i

! !Although it is intended that the plant will be operated with the AFD l

within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER 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-fi ' ciently to change the envelope of peaking factors which may be reached on a l

p subsequent return to RATED THERMAL POWER (with the AFD within the target band) l provided the time duration of the deviation is limited. Accordingly, a 1-nour 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 of ";r: 3.2-1 while at THERMAL POWER 1evels between 50% and 90% of RATED THERMAL POWER. 77u,, rn For2,f./

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.

I 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

,l 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 During operation at THERMAL POWER levels than 90% of RATED THERMAL POWER.

L between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer i l . outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 fiOTCHANNELFACTOR-Fh The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the desi .'

and minimum DNBR are not exceeded and (2)gn in the limits event of aon peak-local LOCA the peakpower fuel density 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 grou) move together with no individual rod insertion differing by more t'ian

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.

L l

V0GTLE UNITS - 1 & 2 B 3/4 2-2 Amendment No.29 (Unit 1) l Amendment No.10 (Unit 2) 1 l -

_ . - ~_ .

. q

~ POWER DISTRIBUTION LIMITS BASES l

' HEAT FLUX HOT CHANNEL FACTOR ~and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR Gentinued) .

The Radial. Peaking. Factor, F,y(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), rompips within its limit. The.

u as inthePldNP::hbg 'l F

xy limit for RATED THERMAL POWER (F x S:ter L hit hp:rt per Specification 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 OUADRANT POWER ~ TILT' RATIO' The' QUADRANT POWER TILT RATIO limit assures that the radial power distribu- !

tion satisfies-the design v' alues used in the power capability analysis.

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

The limit:of 1.02, at which corrective action is required, provides DNB j 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. y 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 Fqis reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.

l L For purposes of monitoring QUADRANT-POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that i

the normalized symmetric power distribution is consistent with the QUADRANT

POWER TILT RATIO. The incore detector monitoring is done with a full incore L ' 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 DNB PARAMETERS

.The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in p, '

the transient and accident analyses. The limits are consistent with the t 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 l 2224 psig correspond to analytical limits of 592.5'F and 2205 psig respec-tively, with allowance for measurement uncertainty. ,

1 l

l s V0GTLE UNITS - 1 & 2 B 3/4 2-5

m-

- ~

.-.a Y ADMINISTRATIVE CONTROLS

+ SEMIANNUAL RADIDACTIVE 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_ gaseous effluent monitoring instrumentation was not corrected within the time specified -i' 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 folloving the calendar month covered by the report.

"CI AL P:"wG WTOD L!uF EP^@- 3, t "A" ;

-___ _. .. . .. . . . - . . . - . . . . _ - , , . RTPs _m_, ,u s.m, o o. 4. o um r.xy . ' "" " ' " ' "' ' ' " ' " "" ' "* " " xy . ' " " " ' ' " ' ' " " " ' ' ' " "

.. for least each' reload core.and sh'all-be maintained available in the C ol D

Room. limits shall be established and implemented on a time sca consistent w' normal procedural changes.

l-L The~ analytical.metho . used to generate the F yy limits 1 be those <

L previously reviewed and ap ed by the NRC*. .If anges to these methods are-L deemed necessary they will be e uated in a dance with 10 CFR 50.59'and

! _ submitted to.the NRC for review an r . prior to their use'if the change is determined to involve an unreview ty question or if such a change would require amendment of previ y subm1 documentation.

A report containing the xy limits for all core plan containing Bank "D" control rods and unrodded core planes along with the p of predicted axial core height (with the limit envelope for compariso shall be

-fhPreded to the NRC Document Control desk with copies to the Regional Adm 's-pr

. f tn :nd th: 5:ident Insp::t: _it P 20 d;y; ;f th;i- i:pi;;;nt;tien.

SNCIAL REPORTS 6.8.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time per'iod specified for each report.

N :!" 020; "P;.= Di',t;itation Centrol and L;;d ielle ing Precedores" and WCAP

' 272. A " Westinghouse Releed Safety Evoluetion "ethedclegy."

V0GTLE UNITS - 1 & 2 6-21

l g 4 4 1

1 INSERT "A" -

= 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:

a. SHUTDOWN MARGIN LIMIT FOR MODES I and 2 for Specification 3/4.1.1-1.

b. SHUTDOWN MARGIN limits for MODES 3, 4 and 5 for Specification 3/4.l'.1.2

c. Moderator temperature coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

d. Shutdown Rod Insertion Limit for Specification 3/4.1.3.5,

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 RTP and Fxy, 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:

a. WCAP-9272-P-A, " WESTINGHOUSE RELOAD. SAFETY EVALUATION .

METHODOLOGY", July 1985 (W Proprietary).

(Methodology for Specifications 3.1.1.3 --Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - i

- Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference,

3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy L Rise Hot Channel Factor.)

b. WCAP-8385, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING 3 PROCEDURES - TOPICAL REPORT", September 1974 (W Proprietary).

(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)

1 :-

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

3-- a

' ^'

.(

1-L INSERT "A" '

1

~

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

I Commission, Section 4.3, Nuclear Design, July 1981. ' Branch -

.l

!' Technical ~ Position CPB 4.3-1, Westinghouse Constant Axial Offset- i Control- (CAOC), Rev.2, July 1981. - -l (Methodology for Specification 3.2.1 - Axial Flux Difference !

-(Constant Axial Offset' Control).) -l

.e. WCAP-9220-P-A,-Rev. 1, " WESTINGHOUSE ECCS EVALUATION M00EL-1981 ,

VERSION", February 1982 (W Proprietary).

(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 .

~1imits, nuclear limits such as shutdown margin, and transient and accident 1 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.

i i

. _ _ _ _ _ _ _ _ _ _ . -. .- G.-

W Ws M w; , ,, m: y n +up

-u ,

4 j s 8: >

l e +

5;' -

OM.)

c 3"-

g _m cy ,L;- y +

a: -g p *

--li n, ,.

= 'c '

'^;(

wo s yt: r - -

f W- l n a- f,,2_ _ _

us ,

gyngg; -

sn N}b' (k y , -j-y.

y ~

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=

r-4 Y DEFINil10NS -

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9; SECTION- ,

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RAGE. ,

pig i ' -

I ll0L DEFINITIONS: d W m..

~

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bliliiAC110N........'.'.........c.....................e.s......i......

. - - :1 le ;

, 3 g

11'2: TACTUATIONJLOGIC TEST....'.. .................c................. L1 l! :-,)

~

~t W JM,,

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1. 3 > ; AN Al0G CH ANNE L O PE R AT ION AL ^ T EST . . . . . . . .. . . .. . . . . . . . . '. . . . '. . . . . . . .: -

.. 1:I - ei yS :;:0. .

J1.4?.AX1AL FLUX DIFFERENCE........,.............. 4 .'.i........',..i',' 11-11 g -1. 55 ;CHA'NNEL cad BRAT 10N. ...........,_.................'....i...;.....-

~

1 -1 l i;f * ~

,x y

a ' 1:. 6 : CHANNEL CHECK...................'............;.................. - -

l?

ep s_ r i

' l '. 7 CONTAINMENT INTEGRITY..................

- . ..........;............ 1 -2 '

y_ .

l.8 - I ' CONTROLLED LEAKAGE'.......:....,....................."..~........- .

.. ~ 1 -2 : -,'

'L y

L1.9 CORE ALTERAT10NS'.......-.......................................

2; 31310 CORE OPERATING LIMITS REP 0RT.................................. . L1-2 .

D1. l l ' DOSE E QU I VALE Ni t i -131. . v . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E1L21 4

"J'

1.12 -:: L-AVERAGE DISINTEGRAT ION ENERGY. . . . . . . . . . . . . ... .............. 1~ 3

$ #2 l.131 ENGINEERED SAFETY FEATURES RESPONSE TIME... ............,:.... 1 -3 i W l.14 FREQUENCY-NOTATION.,.........,...... ......, ............... . 1 -3 e 1.15c GASEOUS WASTE' PROCESSING ~ SYSTEM.....,..........'............... [l-3 L

?l .16 SIDENTIFIED EEAKAGE.'.-.......................................... 1 E l1.17:MA'STERRELAY;1EST...................i....................... /1-3'-

-pr -

.- . - x

^.

%, 1.18- MEMBER (S):0F THE.PUBLIC...................... ................

11-4

1.19c 0FFS!)E DOSE CALCULAT10N' MANUAL........ ................ ..... il 11.20_ OPERABLE---OPE'RABILITY..;. ................. ...... ......... 144:

y .l.21 'OPERAlI0NAL MODE'~' MODE.. .. .... . . .....

. - ................. ' 1 -4 :

ri

... . .l.22: PHYSICS TESTS.,. ....... ...

m w,' . ............. . ... .......'.... . 11 -4 '

l.23 _ PRESSURE BOUNDARY LEAKAGE......... ........ . ..... ....... .. 4

[ , ,

1.24 : PROCESS C0NTROL' PROGRAM;........ .. ..... ... .... .. ..... 15-T. 1.25 . PURGE 4 PURGING................... .... ,. . ... . 1 -5 .

I 1.26...QUADRANI POWER TILT RATIO. . . .. . . I4 m

1.27.- RATED THERMAL POWER., . ... . . . . ... .. . . ... 15 1.28- REACTOR TRIP SYSTEM RESPONSE TIME. ... .. .. 1 -5 ;

i 1.29 REPORTABLE EVENT. .. ... . .. . 1-5

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,y V0 GILE UNiiS -~1 & 2 I /

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DJ' 4 .

D ,

INDEX-i" '

k DEFINITIONS s.

r SECTION . , P AGJ, :

', l1'.30 SHUTOOWN MARGIN .............................................. 1-5 1.31 SITE 80VNDARY................................................. 11-6 1,32 SLAVE RELAY TEST.............................................. 1-6 1.33 ' SOLIDIFICATION. .............................................. 1 1. 3 4 ' S OU R C E C H E C K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.35' STAGGERED TEST BASIS.......................................... 'l-6 1.36' THERMAL PCWER................. ............................... .1-6 1.37 TRIP. ACTUATING DEVICE OPERATIONAL TEST...............-......... 1-6 1.38 UNIDENTIFIED LEAKAGE.......................................... 1-6 ,

1,39 UNRESTRICTED AREA........................... ................, 1-7 1.40 VENTILATION EXHAUST TREATMENT-SYSTEM.......................... 1-7 1.41 VENTING....................................................... 1-7 TABLE 1.1 FREQUENCY NOTAT10N....................................... 'l-8 TABLE 1.2 OPERATIONAL MODES... ....... ......... .................. 1-9 '

s V0 GILE UNITS - 1 & 2 11

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p -

(tv

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e p* i INDEX

[ -LIMITING CONDITIONS FOR-OPERATION AND SURVEILLANCE ~RE00fREMENTS L ,

E SECT 10N' . pag, r-3/4.0' APPLICABILITY.~............................................... 3/40-1 p.

t 3/4.1~ REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL .

~

Shutdown Margin - MODES 1 and 2...-......................... 3/4 1-11

)

Shutdown Ma rgin - MODES 3, 4 and 5. . . . . . . . . . . . . . . . . . . . . .. . .

. 3/41-3:

Moderator Temperature Coeff.icient..........................

~ h~

= 3/4 .1-4 :

Minimum Temperature for Criticality........................ '3/4 1-6.

3/4.1.2 BORAT10N SYSTEMS 1-Flow Path - Shutdown....................................... '3/4 1-7 Flow Paths - Operating..................................... 3/4-1-8 Cha rg i ng Pump - Shu td own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 3 /4 1 -9 Charging Pumps - Operating................................. 3/4 l'10 Borated Water Source - Shutdown............................ 3/4 1-11 Borated Water Sourc es - Operating. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-l2' 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 INOPERABLE CONTROL OR SHUTDOWN R00............. 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 i

i l

V0 GILE UNITS - 1 & 2 IV

r. , ,

( - '

~e ,

1 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS' SECTION' PAGE 3/4.2 POWER DISTRIBUTION LIMITS- -

3/4.2.1- AXIAL FLUX 01FFERENCE..................................... 3/4 2-1 FIGURE 3.2-1 (DELETE 0)............................................. 3/4 2 l 3/4.2.2 HEAT FLUX HOT CHANNEL FAC10R - Fn(7)....................., 3/4 2-4 FIGURE'3.2-2 (DELETE 0)............................................. 3/4 2-5 ll N

, 3/4.2,3 NUCLEAR ENTHALPY RISE H01 CHANNEL F ACTOR '- F aH. . . . . . . . . . . . 3/4 2-8 3/4.2.4 00ADRANT POWER TILT RAT 10................................. 3/4 2-10 3/4.2.5- DNB PARAMETERS............................................ 3/4 2-13 4

3/_4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................... 3/4-3-1 t TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION.................... 3/4 3-2 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................. 3/4 3-9:

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...................... .................. 3/4 3-15 TABLE 3.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............. ............................. 3/4 3-17.

TABLE-3.3-3 ENGINEERED SAFETY FEATURES ACIUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS............................ 3/4 3-28 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVE!LLANCE REQUIREMENTS................. 3/4 3-36 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations... ..... . ..... 3/4 3-45 TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS........ ........... .. ........... 3/4 3-46 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS. . . . .. . . . .. 3/4 3-48 Movable Incore Detectors. .. ..... . . . . .. . ... 3/4 3-49 Seismic Instrumentation... ..... ... . . . . . 3/4 3-50 t

V0GTLE UNITS - 1 & 2 V

m .

"c -

- 4, 4 ;;

jeu

(-

Us.?

s INDEX. 4

~~i-ADMINISTRATIVE CONTROLS -i 6

5 3 -SECTION- , Apag 6.4.2 SAFETY REVIEW BOARD (SRB)

Function................................'..................... 6-9 Composition.................................................. 6 Alternates................................................... . 6 Consultants.................................................. 6-10 Meeting Frequency............................................ 6-10 I

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

Review....................................................... 6 ,

Audits....................................................... 6-11 Records...................................................... 6-12 6.5 REPORTABLE EVENT ACTION....... ................................ 6-13 6.6 SAFETY LIMIT VIOLAT10N......................................... '6-13. '

6.7 PROCEDURES AND PR0 GRAMS......................

................. 13 6.8 REPORTING REOUIREMENTS 6.8.1 ROUTINE REP 0RTS. ............................................ 6-17 Startup Report............................................... 6-17 Annual Report................................................ 6-17

-Annual Radiological Environmental Surveillance Report........ 6-18 Semiannual Radioactive Effluent Release Report............... 6-19 4

Monthly Operating Reports...... ................ .... ... ... 6 Core Operating Limits Report.............. .................. 6-21 l 6.8.2 .SPECIAL REPORTS..... .. ..... .............................. 6-21a1 4

6.9 RECORO RETENTION....... .. ..... ..... ...... ................. 6-22 V0GTLE UNITS - 1 & 2 XXill

yp&q, y; a ,

1.' .

o

,qd. y m

Y ;iL

' DEFINITIONS' -

' CONTAINMENT INTEGRITY .

4

--a v ..

a cc ~ 1.7;ECONTAINMENTJINTEGRITY shall exist when:

.a. Allipenetra't' ions required to.be. closed during accident conditions

, are.either:

m 1)' CapableLof being closed by an OPERABLE containment automatic

'I lisolation valve system, or

2).: Closed by manual valves, blind flanges,-or deactivated. automatic--

valves secured _in their closed positions. Ei t

b...All equipment hatche's'are closed and sealed, L !

c. Each air lock is in compliance with the requirements of Specification

> 3.6.1.3,

d. The containment. leakage rates are within the limits of Specif.ication 3.6.1.2, and ,

1

e. The sealing mechanism associated with each penetration-(e.g., welds,. .t bellows, or-0 rings) is OPERABLE. -

' CONTROLLED LEAKAGE 1.8 ~ CONTROLLED LEAKAGE"shall be that seal water flow supplied to the reactor.

coolant = pump seals. ..

CORE-ALTERATIONS

~

1.9 : CORE . ALTERATIONS shall be the movement or ma'nipulation of any component- l withiri the' reactor pressure -vessel.with the vessel head removed and fuel in ~

j the vessel. Suspension'of CORE ALTERATIONS shall~not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORTS 1.1'0. .The. CORE.0PERATING~ LIMITS REPORT (CQLR) 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

cycle; in accordance with Specification 6.8.1.6. Unit operation within these-4 . 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) which alone would' produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133,1-134, 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.

V0GTLE UNITS & 2 1-2

_. I

p . .,;

4 DEFINITIONS' E - AVERAGE DISINTEGRATION' ENERGY f

~

-1.'l2 E shall b'e the average,' weighted in proportion to the concentration of I 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. y

~

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY. FEATURES (ESF) RESPONSE TIME shall be that time- l interval from when the monitored parameter exceeds its ESF Actuation Setpoir.t at the channel. sensor until the ESF equipment is capable.of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures. reach their required values, etc.). Times shall include-diesel generator starting and sequence loading delays where applicab,le.

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.

1 GASEOUS WASTE PROCESSING SYSTEM i

1.15 A GASEOUS WASTE PROCESSING SYSTEM shall be any system designed and installed l l to reduce' radioactive gaseous effluents by collecting Reactor Coolant System l 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 j environment. ~i IDENTIFIED LEAKAGE

'1.16 IDENTIFIED LEAKAGE shall be: l

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

j or collecting tank, or ;

q

b. Leakage into the containment atmosphere 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 l

L LEAKAGE, or

li

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 i a continuity check of each associated slave relay.

V0GTLE UNITS - 1 & 2 1-3 l

l

, q .

.1 y

DEFINITIONS MEMBER ('S)' 0F THE PUBLIC

'1'.18- MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa- 'l' tionally associated-with the plant. This category does not include employees

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

OFFSITE U0SE CALCULATION MANUAL 1.19 The DFFSITE DOSE CALCULATION MANUAL-(0DCM) shall contain the methodology -l 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 /Trlp 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 OPERABL'E or l E[ 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 MODE - MODE .

I 1.21 An OPERATIONAL MODE (i.e. , MODE) shall correspond to any.one-. inclusive l 4

. 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 fundam' ental l nuclear characteristics of the reactor core and related instrumentation:

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

l PRESSURE BOUNDARY LEAKAGE i

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

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

V0GTLE UNITS - 1 & 2 1-4

77, t

4 r

DEFINITIONS PROCESS CONTROL PROGRAM 1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the. current formulas, sampling,' analyses, tests, and determinations to be made to ensure that' l processing 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 E

Federal and State. regulations, burial ground requirements, and other require-

'ments 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' j 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

-shall be used for computing the average.

RATED THERMAL POWER

'I.27- RATED' THERMAL POWER shall be a total reactor core heat transfer rate to. I the reactor coolant of 3411 MWt.

-REACTOR TRIP SYSTEM-Resp 0NSE 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 i

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 P

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 l assumed to be fully withdrawn.

V0GTLE UNITS - 1 & 2 1-5

1

( y

-s.

DEFINITIONS

~ SITE BOUNDARY l'.31- 'The SITE BOUNDARY shall be the exclusion boundary lineL as shown.in Figure 5.1-1.- l .

SLAVE RELAY TEST -

m

1.'32 A SLAVE' RELAY TEST shall be the energization of each. slave relay and ll verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include l a continuity check, as a minimum, of-associated testable actuation devices.. 3 SOLIDIFICATION g,

1.33 SOLIDIFICATION shall be the conversion of' wet wastes into a form that- .l meets shipping and burial ground requirements. ,

-SOURCE CHECK 31.34 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 11.35 A STAGGERED TEST BASIS shall cont.ist of: !

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

designated components obtained by dividing the specified test

. interval into n equal subintervals,' and i

b. .The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval. 1 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'0PERATIONAL TEST shall consist of operating the -l Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or -

l trip functions. .The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment,. as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE

, 1.38 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE l or CONTROLLED LEAKAGE.

,V0GTLE UNITS - 1 & 2 1-6

+

- ~

~ . .

?

e

[

DEFINITIONS

UNRESTRICTED AREA

~

.1.39 'An. UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY l accessxto 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 B0UNDARY used for. residential-quarters or forLindustrial, '

commercial,-institutional, and/or recreational purposes. *

. VENTILATION EXHAUST TREATMENT SYSTEM 1.40 A VENTILATION EXHAUST TREATMENT SYSTEM.shall be any system designed-and ' l-installed to reduce-ga:enus radiciodine or radioactive material in particulate

' form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/orLHEPA filters for the purpose of removing iodines or particu-lates from the gaseous exhaust stream prior to the release to the environment, c

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 process 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.

t 0

V0GTLE UNITS - 1 & 2 1-7

m Ici t

3/4.1 REACTIVITY CONTROL-SYSTEMS 3

3/4.1.1 BORATION CONTROL SHUTOOWN MARGIN - MODES 1 AND 2 '

LIMITING-CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to the limit specified in the CORE OPERATING LIMIlS REPORT (COLR).

APPLICABILITY: MODES 1 and 2*.

ACTION:

'With the SHUTDOWN MARGIN less than the ' limit specified in the COLR, immediately initiate and continue boration at greater than cr equal to 30 gpm of a solution l containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTOOWN 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.

l_

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months af ter 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 thereaf ter 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 Kerr 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 within the limits of Specification 3.1.3.6;

c. With Keff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6; and

d. Prior to initial operation above 5% RATED THERMAL POWER af ter each fuel loading, by consideration of the factors below, with the control banks at the maximum insertion limit of Specification 3.1.3.6:

See Special Test Exceptions Specification 3.10.1.

V0GTLE UNITS - 1 & 2 3/41-1

V y

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - MODES 3. 4 AND 5 LIMITING CONDITION FOR OPERATION l

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.

SURVEILLANCE REOUIREMENTS 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 -l least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter while the rod (s) . is inoperable.

If an inoperable control rod is immovable or untrippable, 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, f

2) Control rod position,

3) Reactor Coolant System average temperature, I

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

l V0GlLE UNITS - 1 & 2 3/4 1-3 1

L: - - -

t REACTIVITY CONTROL SYSTEMS S

MODERATOR TEMPERATURE COEFFICIENT ,

LIMITING CONDITION FOR OPERATION ,

3.1.1.3 The moderator temperature coef ficient (M1C) shall be wi' ,the

. Beginning of Cycle Life (BOL) limit and the End of Cycle-Life ' .) limit specified in the CORE OPERAllNG LIMlls REPOR1 (COLR). The e- .ium upper limit shall be

Unit 1: I Less positive than + 0.7 x 10-* Ak/k/'F for power levels:up to l 70% RATED THERMAL POWER with a linear ramp to O Ak/k/*F at 100%

RA1ED THERMAL. POWER; and Unit 2:

Less positive than 0 Ak/k/*F.

APPLICABILITY: BOL limit - MODES 1 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 MIC 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 M1C has been restored to within its limit for the all rods withdrawn condition; and

With Kef t greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

V0G1LE UNilS - 1 & 2 3/4 1-4

m ,

2, r

f. '

p.-

REAC1JVITY CONTROL SYSTEMS SURVEllLANCE REQUIREMENTS

3. A Special Report is prepared and submitted to the Commission, pursuant to Specification 6.8.2, within .10 days, describing the i value of the measured MIC, the interim control rod. withdrawal -;

limits, and the predicted average core burnup necessary'for .

restoring the positive MIC to within its limit for the all-rods withdrawn condition,

b. With the MTC more negative than the.EOL limit specified in'the COLR, be.in H01 SHU100WN 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 f uel ;

cycle as follows: '

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 POWlR, I 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 l

withdrawn.. RATED THERMAL POWER condition) within 7 EFP0 after reaching an equilibrium 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 'ae remeasured and compared to the EOL MTC limit specified in the COLR, at least once per.14 EFPD during the remainder of the fuel cycle, 1

V0 GILE UNilS - 1 & 2 3/41-5 1

9;.

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ,

l GROUP HEIGHT LIMITING CONDITION FOR OPERATION

~ 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*.

AQTION:

a. With one or more rods inoperable due to being immovable as a result of excessive f riction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within -1 hour 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 height by more than i 12 steps (indicated position),

POWER OPERATION may continue provided that within 1 hour:

1. The rod is restored to OPERABLE status'within the above alignment requirements, or

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER l level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

i

' a) A reevaluation of each accident analysis of Table 3.1-1 is.

performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents l remain valid for the duration of operation under these conditions; 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 l

1

f_

i' '

L i

REACTIVITY = CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION' ACTION (Continued) I i

c) A power distribution' map is obtained from the movable incore detectors and FQ (2) and F!H are verified'to be 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% 'I of RATED THERMAL POWER. '

c. With more than one rod trippable but inoperable due to causes.other than addressed;by ACTION'a above, power operation may continue.

provided that:

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 inoperable rods are aligned to within 1-12 steps of the inoper-able rods while maintaining the rod sequence and insertion '<

limits of Specification 3.1.3.6. The. THERMAL POWER level- _b shall be restricted pursuant to Specification =3.1.3.6 during subsequent operation, and

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 ,

d. With more than one rod misaligned from its group step counter demand height by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . .

SURVEILLANCE REQUIREMENTS 4 .1. 3 .1.1 The position of each rod shall be determined to be within the group demand limit by verifying the individual rod positions 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 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 OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days. l l

l V0GTLE UNITS - 1 & 2 3/4 1-15

p- ,

{:

{(

t 6

REACTIVITY CONTROL SYSTEMS ~ '

F SHUTDOWN ROD 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). i APPLICABILITY: MODES 1* and 2* #.

ACTION: '

With a maximum of one shutdown rod inserted beyond the insertion limit

-specified in the COLR, except for survaillance 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:

J

a. Restore the rod to within the insertion limit specified in the COLR, or I 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 '

L Bank A,-B, C, or D during an approach to reactor criticality, and 1

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

l l

See _Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With Keff greater than or equal to 1.

4 V0GTLE UNITS - 1 & 2 3/4 1-20 f

4 y=

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS 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).

APPLICAB_ILITY: MODES 1* and 2* #.

ACTION:

With the control banks inserted beyond the insertion limits, except for I surveillance testing pursuant to Specification 4.1.3.1.2:

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 f raction 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 H01 STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . '

SURVEILLANCE REQUIREMENTS l

F 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 .

l l

i l

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With Keft greater than or equal to 1.

V0GTLE UNITS - 1 & 2 3/4 1-21 u..

y n

3/4.2 POWER DISTRIBU110N LIMITS t -

i F4.2.1 AXIAL FLUX OlFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated (N!-00418, N!-00428, NI-00438, NI-00448) AXIAL FLUX OlFFERENCE ( AFD) shall be maintained within the target band (flux dif ference units) about the target flux dif ference. The target band is specified in the e

CORE OPERATING LIMITS REPORT (COLR).

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

The indicated AFD may deviate outside the required target band at greater l than 15% but less than 50% of RATED 1HERMAL POWER provided the cumulative penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY: MODE 1, above 15% of RA1ED THERMAL POWER.* #

ACTION:

a. With the indicated AFD outside of the required target band and I with 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.

b. With the indicated AFD outside of the required target band for l

l more 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 hoirs or outside the Acceptable Operation Limits specified in the COLR and with THERMAL POWER less than 00% but l equal to or greater than 50% of RA1ED THERMAL POWER, reduce:

1. T51ERMAL 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 1HERMAL POWER) pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits specified 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 l the AFD outside of the above required target band during testing without penalty deviation.

V0GTLE UNITS - 1 & 2 3/42-1 ;

t q-r i _

i

[ POWER DISTRIBUTION LIMITS i

LIMITING CONDITION FOR OPERATION ACTION (Continued)

I

c. With the indicated AFD outside of the required target band for l !

i more than i hour 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 less than 50% but areater '

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 deviation has been reduced to less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in l }

the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . '

. SURVEILLANCE REOUIREMENTS '

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 per 7 days when the AFD Monitor Alarm is OPERABLE, <

and I

2) At least once per hour until the AFD Monitor Alarm is updated af ter restoration to OPERABLE status. ,

5

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 thereaf ter, when the AFD Monitor Alarm is '

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

4 4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the required target band shall l be accumulated on a time basis of:

a. One minute penalty deviation for each 1 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 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and ,

l 50% of RATED THERMAL POWER.

l 4.2.1.3 The target flux dif ference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Ef fective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

l l 4.2.1.4 The target flux dif ference shall be updated at least once per i

31 Effective Full 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

p. W+ 2 gy 95 v-wvu~pp i i

4.

t 9-t 1

i l

l n

1 I

k

.k i

i

}

(

l i

y J

.J i

ie 3, - -

4' J

1 J

1 1

1 Y

1 a

4 k a 'e

$l- b 1

4 e

4, i r

e i p

u (

i r gr r s -k

p. p y

v 7

i s

FIGURE 3.2-1 (DELETED) '

i t

b =V0G1LE UNITS - 1 & 2 3/4 2-3 r

r t

~ . - , . , - - ~~-+-.La < L)

e.

~\ L-i 8

t POWER DISTRIBUTION LIMITS .

i 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fn[Il t .

LIMITING CONDITION FOR OPERATION i >

3.2.2 Fn(Z) shall be limited by the following relationships:

FQ(Z) 5 F0 [K(Z)] for P > 0.5 P

FQ(Z) 5,g [K(Z)] for P $ 0.5 ,

Where: FQ = the FQ limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR),

P = THERMAL POWER , and RATED THERMAL POWER K(Z) = the normalized Fg(Z) as a function of core height specified in the COLR. i APPLICABILITY: MODE 1.

ACTION:

With Fg(Z) exceeding its limit:

. a. Reduce THERMAL POWER at least 1% for each 1% Fg(Z) exceeds i the limit within 15 minutes and similarly reduce *.he 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 ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1%

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

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 Fn(Z) is demonstrated through incore mapping to be l within its limit.

l l

l V0G1LE UNITS - 1 & 2 3/4 2-4

,. , - , , , , ,,e 1

0 1

4 i i 4

1

}

1 4

?

l 1

I I

i J

1 i :

i

'b

,i i

1 3

1 i

i i

9

?

I 1

L t

1 i

t j

l t

'a k

.a i

9 I

.i i

i l

t t

i'

?

a l>

b P

w C

b c

.. FIGURE 3.2-2 (DELETED) h c

' ?

V0GTLE UNITS - 1 & 2 3/4 2-5 b NbM& M94 %.um2.uA i.y._._m,_%.,,,_,m _.~.,m..__m-~,m_ m - ~ ~ "" '

>. m,

! ce .,

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS i r

i 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. '

4.2.2.2 Fxy shall be evaluated to determine if Fg(Z) is within its limit by:

L a. Using the movable incore detectors to obtain a power distribution map at any THERMAL, POWER greater than 5% of RA1ED THERMAL. POWER before exceeding 75% of RATED TilERMAL POWER following each fuel !

loading.

1

b. Increasing the measured Fx component of the power distribution map by 3% to account for manuf!cturing tolerances and further increasing -

the value by 5% to account for measurement uncertainties,

c. Comparing the Fxy computed (Fxh) obtained in Specification 4.2.2.2b.,

above to:

1) measuredcoreplanesgiveninSpecificatio[n4.2.2.2e.andf.,T below, and

2) The relationship:

Fxy=F!y [1+PFxy(1-P)],

Where Fxy is the limit for fractional THERMAL POWER operation I expressedasafunctionofF! is the Power Factor l Multiplier for Fxy specified fn PFxbOLR, the and P is the fraction of RATED THERMAL POWER at which Fxy was measured.

d. Remeasuring Fxy according to the following Scnedule:

1) WhenExhisgreaterthantheF![limitfortheappropriate ,

measured core plane but less than the Fxy relationship, additional power distribution maps shall be taken and Fxy compared to F![

l and Fxy either:

a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which Fxy was last determined, or b) At least once per 31 Effective Full Power Days (EFPD),

whichever occurs first.

YOGTLE UNI 1S - 1 & 2 3/4 2-6

c. a 1.. !

POWER DISTRIBUTION LIMITS p SURVEILLANCE REQUIREMENTS fContinued) i C RTP !

2) When the F xy is less than or equal to the Fxy _ limit for the appropriate measured core plane, additional power distribution mapsshallbetakenandFxhcomparedtoFby and Fxy at least once per 31 EFPO.

g e. The Fxy limits used in the Const4M :xial Offset Control analysis forRATEDTHERMALPOWER(FfP y ) shall be specified for all core l planes containing Bank "0" control rods and all unrodded core planes in the COLR per Specification 6.8.1.6; l

f.

The Fxy limits of Specification 4.2.2.2e., above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:

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

,- 2) Upper core region from 85 to 100%, inclusive,

3) Grid plane regions at 17.812%,32.112%,46.412%,60.6 25 and 14.9 1 2%, inclusive, and

4) Core plane regions within + 2% of core height ( 2.88 inches]

about the bank demand position of the Bank "D" icontrol rods.

g.

WithFxhexceedingFxytheeffectsofFxfs.nFn(Z)shallbeevalua to determine if F Q (Z) is within its limi o

4.2.2.3 When F determinations, an overall measured Fn(Z) shallg(Z) is measured be obtained f rom afor otherdistribution power than Fxymap and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, f

I i

I j' V0GTLE UNITS - 1 & 2 3/4 2-7

,y

t .

P9fLR_ DIS 1_RIBUT10N LIMI15 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR-Ffg LIMITING CONDITION FOR OPERATION P

3.2.3 FfH shall be limited by the following relationship:

Ffy<FaH [1 + PFAH(1-P)J i Where: F3 = The F"H limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR).

p , PFAH = The Power Factor Multiplier for F H specified in ,

the COLR, and P = THERMAL POWER RA1ED THERMAL POWER APPLICABIL[1_1: MODE 1.

ACTION:

With F H exceeding its limit: ,

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

1. Restore F"g 3 to within the above limit, or

2. Reduce THERMAL POWER to less than 50% of RATED 1HERNAL' POWER and reduce the. Power Range Neutron Flux - Hig', 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 F6 has been restored to within the above limit, or reduce THERMAL OWER 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 THERMAL POWER limit required by AC110N a.2. and/or b., above; subsequent POWER OPERATION ,

may proceed provided that F is demonstrated, through incore flux mapping to be within its lint prior to exceeding the following THERMAL POWER' levels:

1. A nominal 50% of RATED lHERMAL 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 equel to 95% of RATED THERMAL POWER.

V0GTLE UNITS - i & 2 3/4 2-8

\

s 3/4.1 REACTIVITY CONTROL SYSTEMS BASES

. i jf_42 1 BORATION CONTROL I t

3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made

subtritical 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 preclude total loss of SHUTDOWN MARGIN in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T In MODES 1 and 2, avg.

the most restrictive condition occurs at EOL, with Tavg 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 SHUTOOWN 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 1 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 f rom the initiation of the Source Range High Flux at Shutdown Alarm to total loss of SHUTDOWN MARGIN. Accordingly, the SHUTDOWN MARGIN requirement is based 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 coef ficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting

-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.

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 V0 GILE UNilS - 1 & 2 B 3/4 1-1

b 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 l

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 i boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting EOL MTC value, j 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 RTNDT temperature.

3/4.1.2 BORAT10N 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, (?) 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 boration capability of either flow path is suf ficient to provide a SHU100WN V0G1LE UN115 - 1 & 2 B 3/4 1-2

\

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Nonnal 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 CCCS 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:

Fg(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat 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!H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and Fxy(Z)- Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX OlFFERENCE The limits on AXIAL FLUX OlFFERENCE (AFO) assure that the Fg(Z) upper bound envelope of the FQ limit specified in the CORE OPERATING LIMllS REPOR1 (COLR) times K(2) is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux dif ference is determined at equilibrium xenon conditons.

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 dif ference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux dif ferenr.e at RATED THERMAL POWER for the associated core burnup conditions. Target flux dif ferences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate f ractional THERMAL POWER level. The periodic updating of the target flux dif ference value is necessary to reflect core burnup considerations.

I l

V0GTLE UNITS - 1 & 2 8 3/4 2-1 l

+

j-t POWER DISTRIBUTION LIMITS BASES AXIAL-FLUX UlFFERENCE (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 dif ference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.

This deviation Will not affeet 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 target 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'l-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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 HOTCHANNELFACTOR-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 accept 6nce 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 inidivdual rod insertion dif fering by more than i 12 steps, indicated, from the group demand position;

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

I V0GTLE UNITS - 1 & 2 8 3/4 2-2

l l i ;i i i u g Iu I IHI I 18 '

w

(

2 b

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

The Radial Peaking Factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor,nF (Z), remains within its limit. The FxylimitforRATEDTHERMALPOWER(FxhTP)asspecifiedintheCOLRper Specification 6.8.1.6 was determined f rom expected power control manuevers over the full range of burnup conditions in the core.

3/4.2.4 00ADRANT POWER TILT RATIO The QUADRANT POWER-TILT RATIO limit assures 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 Fg is reinstated by reducing 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 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal 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 avg 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 uncertainty.

V0GTLE UNITS - 1 & 2 B 3/4 2-5

s.

s ' ADMlWISTRAllVL CONIROLS SEMIANNUAL RADIOACTIVE EFFLUEN1 RELEASE REPOR1 (Continued)

The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous ef fluent 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, Of fice of Resource Management, U.S. Nuclear Regulatory Commission, Washington 0.C. 20555, with a copy to the Regional Administrator of the Regional Of fice of the NRC, no later than the 15th of each month following the calendar month covered by the report.

CORI. OPIRATING l.lMITS RFPOR1 6.8.1.6 Core operating limits shall be established and documented in the CORE OPERA 11NG LIMlls RI.POR1 (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

a. SHU100WN MARGIN LIMil FOR MODLS 1 and 2 for Specification 3/4.1.1.1,

b. SHU100WN MARGIN limits for MODES 3, 4, and 5 for Specification 3/4.1.1.2,

c. Moderator temperature coef ficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

d. Shutdown Rod Insertion I.imits for Specification 3/4.1.3.5,

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 R1P for Specification 3/4.2.2,

h. and Fxy,Enthalpy Rise Hot Channel Factor Limit and the Power Nuclear 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:

a. WCAP-9272-P-A, "WfSIINGHOUSL RiiDAD SAFEIY [ val.UAI10N ME1HODOLOGY," July 1985 (W Proprietary).

(Methodology f or Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion limit, 3.1.3.6 -

Control Bank Insertion I.imits, 3.2.1 - Axial flux Difference, 3.2.2 - lleat I lux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

i V0G1LL UN115 - 1 & 2 6-21

e-A,DMINISTRA11VE CON 1ROLS i

CORE OPERATING LIMlls REPORT (Continued)

b. WCAP-8385, ' POWER DISlRIBUTION CON 1ROL AND LOAD FOLLOWING PROCEDURES - TOPICAL REPOR1,* September 1974 (W Proprietary).

(Methodology for Specification 3.2.1 - Axial Flux Dif ference (Constant Axial Offset Control).)

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

January 31, 1980 --

Attachment:

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

(Methodology for Specification 3.2.1 - Axial Flux Dif ference

[ Constant Axial OfIset ControlJ.)

d. NUREG-0800, Standard Review Plan, U. S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981. Branch lechnical Position CPB 4.3-1, Westinghouse C9nstant Axial Of f set Control (CAOC), Rev.2, July 1981.

(Methodology for Specification 3.2.1 - Axial Flux Dif ference (Constant Axial Offset Control].)

e. WCAP-9220-P-A, Rev. 1, "WLSTING110VSE LCCS EVALUATION MODEL-1981 VERSION," February 1982 (W Proprietary).

(Methodology for Specification 3.2.2 - lleat Flux llot 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 as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERAllNG LIMils REPOR1, 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.

SPICIAL REPOR1S 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.

V0GlLL UNlls - 1 & 2 6-21a I

ML20012E013

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