Text

Amend 24 to License NPF-6,authorizing Cycle 2 Operation Subj to License Condition Temporarily Restricting Facility Operation to 70% of Licensed Full Power Level Pending NRC Review of Core Protection Calculator Sys
	ML20004G020
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/19/1981
From:	Clark R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20004G021	List: ML20004G020; ML20004G022; ML20004G023;
References
	NUDOCS 8106260496
	Download: ML20004G020 (94)
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AogaysaS POWER AND LIGHT COMPAN)

ARKaN5A5 NUCLEAR ONE, UNIT N0. 2 DOCKET NO. 50-368 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 24 License No. NPF-6 1.

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

A.

The application for amendment by Arkansas Power and Light Company (the licensee) dated February 20 and March 5, 1981, as supplemented, complies with the standards and requirements of the Atomic Energy Act of 1954,las amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.

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

There is reascnable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

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

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2. _ Accordingly, the license is. amended by deletion of Licansa Condition:

2.C.3.a. d, f, and i and cnanges to the + Technical Speci~iuativos as -

indicated in the attachment to this license amendment, and paragraphs 2.C.(1) and 2.C.(2) of Facility Operating License No. NPF-6 are hereby amended to. read as follows:

~ (1) ~ Maximum Power Level The licensee is authorized to operate the facility at steady state power levels not in excess of seventy percent of 2815 megawatts thermal.

Prior-to attaining this power level the licensee shall comply with the conditions -specified -in Paragraph 2.C.(3).

(2) ' Technical Specifications -

'The! echnical Specifications contained in Appendices A and T

B, as revised through Amendment No. 24, are hereby incorporated in the license.

The licensee shall_ operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f

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AN s.

R bert A. C ark, hief Operating Reactors Branch #3 Divisicn of Licensing

Attachment:

Changes to the Technical Specffications Date of Issuance: June 19, 1981 i

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ATTACHMENT T0. LICENSE AMENO".ENT NO. 24-FACILITY OPERATING LICENSE NO. NPF-6 GGCKET NG. 50-358 Replace the ' folic.;ing_ pages ofc the - Appendix "A" Technical Specifications with the enclosea pages. The revised pages are identified by amendment number and'contain vertical lines indicating the' area of_ change. -The corresponding overleaf pages are also provided to-maintain document completeness.

Pa ges -

3/4 1-15

' 3/4 3-9 I

3/4 1-27 3/4 3-16 II IV 3/4 2-1 3/4 3-17 IX 3/4 2-2 3/4 3 1-6.

3/4 2-3 3/4 4-1 2-1 3/4 2-4 3/4 4-2 2-4 3/4 2-5 3/4 6-7 2-5 3/4 2-6 3/4 6-18 2-6 3/4 2-7 3/4 7-3 2-7 3/4 2-8 3/4 9-7 2-8 3/4 2-9 3/4 10-2 2-9 3/4 2-10 8 3/4 1-1 B 2-1.

3/4 2 B 3/4 1-2 8 2-2 3/4 2-12 B 3/4 1-3 B 2-5 3/4 2-13 B 3/4 2-1 8 2-6 3/4 2-14 B 3/4 2-2 B 2-7 3/4 3-la B 3/4 2-3 3/4 1-3 3/4 3-2 B 3/4 2-4 3/4-1-5 3/4 3-3 8 3/4 4-1 3/4 1-8 3/4 3-Sa B 3/4 9-2 3/4 1-10 3/4 3-7 5-5 3/4.1-12 3/4 3-8 6-13 e-w - - - -

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v INDEX DEFINITIONS

-SECTION PAGE 1.0 DEFINITIONS D e fi n ed Te rms.............................................

1-1 Thermal Power..............

1 -1 Ra t ed The rma l P owe r........................................

1-1 Opera ti onal Mod e - Mod e...................................

1 -1 Action.....................................................

1 -1 Operable --Operability.....................................

1-1 Reportable Occurrence......................................

1-2 Co n ta i n me nt I n t eg ri ty......................................

1-2 Ch a n n el Ca l i b ra ti o n........................................

1-2 Channel Check..............................................

1-3 Channel Functional Test....................................

1-3

~ Core Alteration......

1-3 Shutdown Margin..........................

13 1

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' Identi fi sd L aa ka ga.........................................

14 Uni den ti fi ed L ea ka ge.....................................

1-4

- P res sure Sounda ry L ea kage..................................

1-4 Azi mutha l Power 711 t.......................................

1-4 Dose Equivalent I-131......................................

1 -4 E-Average Of sintegrati on Energy............................

1-5 S ta g g e red Tes t B a s i s.......................................

1-5 Frequ ency - Nota ti o n.........................................

1-5 Axial Shape Index...............

1-3 Reactor Trip System Response Time.........................

1-5 Engineered Safety Feature Response Time....................

'. - 6 Physics. Tests..............................................

1-6 S o f tw a r e...................................................

1-6 Pl a n a r Ra d i a l P e a k i n g Fa c to r..............................

1-6 I

ARKAiMAS - UNIT 2 I

Amendment No. ? 4 M

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lSECTION PAGE ii

-is-flj 2.1 SAFETY LIMITS R e a c t o r Co r e.......................................'.......... 2-1

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Rea ctor Cool ant Sys tem Pressure............................... 2-2 ir r+

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'!i.2 LIMITING SAFETY SYSTEM SET INGS

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Reactor Trip Setpoints........................................ 2-2 Core Protection Cal cul ator Adcressabl e Constants..............

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- SECTION PAGE

'j2.1 SAFETYLIMITS

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R e a c t o r Co r e................................................

5 2-1

!I; Reactor Coolant System Pressure..............................

3 2-2 7

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Reactor Trip Setpoints.......................................

B 2-2 s.

l ll CPC Addressable Constants....................................

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ARKANSAS - UNIT 2 II Amendment No. 2 't ""'

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 A P PL IC A B I L I TY........................................... 3/40-1 3/4.1 REACTIVITY' CONTROL SYSTEMS 3/4.1.1 BORATION_ CONTROL Shutdown Margin - T

> 200*F.......................

3/41-1 avg Shutdown Margin - T

< 200*F.......................

3/4 1-3 ayg Boron Dilution...................................... 3/41-4 Moderator Temperature Coefficien t................... 3/4 1-5 Mi nimum Temperature for Criticality.................. 3/4 1-6 3/4.1.2 BORATION SYSTEMS Fl ow P a tn s - S hu tdown................................ 3/4 1-7 Fl ow P a th s - 0ce c ati n g.............................. 3/41-8 C ha rg i n g Puma - S hu tdown...................,......... 3;'41-9 Chargi ng Pumo s - 0 cerati ng........................... 3/4 1-10 Boric Aci d Makeup Pumps - Shutdcwn...................

3/4 i-il Boric Acid Makeuo Pumps - Operati ng.................. 3/4 1-12 L

Borated Water Sources - Shutdown..................... 3/4 1-13 Borated Water Sources - Operati ng.................... 3/4 1-15 i

1 3/a.1.3 MOVABLE CONTROL ASSEMBLIES C EA P o s i ti o n......................................... 3/4 1-17 Position Indicator Channels - 0oerating.............. 3/4 1-20

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Position Indicator Channels - Shutdown.............. 3/4 1-22 l-C EA O r o p T i me........................................ 3/4 1-23 i

Shu tdown C EA In s e rtion Limi t......................... 3/4 1-24 Regulati ng C EA Inse rti on Limi ts..................... 3/4 1-25 ARKANSAS - UNIT 2 III

'l-e INDEX

, LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE I

I2 3/4.2 POWER DISTRIBUTION LIMITS

ij, 3/4.2.l' -LINEAR HEAT RATE......................................:. 3/4 2-i j

3/4.2.2 ' RAD I AL P EA KI NG FACTORS '.................................. 3/4 2-4 f

I.;

3/4.2.3 A ZIMUTHAL P OW E?, TI LT '....................................

3/4 2-5 3/4.2.4-DNBR MARGIN.............................................

3/4 2-7 l3/4.2.5 RCS FLOW RATE...........................................

3/4 2-11 1

1

!! 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE....................

3/4 2-12 ii ll-3/4.2.7 AXIAL SHAPE INDEX......................................

3/4 2-13 l

II

,,3/4.2.8 PRESSURIZER PRESSURE....................................

3/4 2-la

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t 3/4.3 INSTRUMENTATION l 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION....

3/4 3-1

'3/4.3.2 ENGINEERED SAFETY FEATURL ACTUATION SYSTEM I N S TRUt4 EN TAT I ON...................................... 3/4 3-10 II3/4.3.3 MONITORING INSTRUMENTATION li I,l Radiation Monitoring Instrumentation....................

3/4 3-24 Incore Detectors........................................

3/4 3-23

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S e i sm i c I n s t rum en ta t i o n.................................

3/4 3-30

'j Meteorological Instrumentation......................... 3/4 3-33 llt Remote Shutdown Instrumentation......................... 3/4 3-36 I l

!,,t Po s t-Acci d ent Ins trum enta ti o n........................... 3/4 3-39 l,!

Chlorine Octection Systems.............................

3/4 3-42 ij Fire Detection Instrumentation.......................

3/4 3-43 ti ll3/4.3.4 TURBINE OVERSPEED PROTECTION...........................,

3/4 3-45

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' ARKANSAS - UNIT 2 IV Anencment No. 9 4

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INDEX BASES SECTION PAGE

~3/4.0 APPLICABILITY............................................ B 3/4 0-1

'('3/4.1 REACTIVITY CONTROL SYSTEMS l

3/4.1.1 BORATION CONTROL......................................

B 3/4 1-1 3/4.1.2-BORATION SYSTEMS...................................... 8 3/4 1-2 3/4.1.3

. MOVABLE CONTROL ASSEMBLIES............................ B 3/4 1-3

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3/4.2 POWER OISTRIBUTION LIMITS l3/4.2.1 LINEAR HEAT RATE......................................

B 3/4 2-1 l{3/4.2.2 RADIAL. PEAKING FACTORS................................ B 3/4 2-2

.!!3/4.2.3 AZIMUTHAL POWER TILT..................................

B 3/4 2-2 i4.

If M 3/4.2.4 DNSR MARGIN...........................................

3 3/4 2-3 i.

_;' 3/1.2.5'.RCS FLOW RATE......................................... 3 3/4 2-4

!!3/4.2.5 REACTOR COOLANT COLD LEG TEMPERATURE.................. S 3/4 2-4

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...i! 3/4.2J AX I A L S HA P E I N D EX.................................... 3 3/4 2-4 i

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3/4.2.8 PRESSURIZER PRESSURE.................................. S 3/4 2-4

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8 3/4 3-1 l,

t,f* 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION............. B 3/4 3-l' l

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- l 3/4.3.3 MONITORING INSTRUMENTATION............................ S 3/4 3-1 i

I 3/4.3.4

' TURBINE OV ERSPEED PROTECTION.........................

3-3/4 3-1 itil I'::

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l " ARKANSAS - UNIT 2 IX Amendment No. 24 i

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.. O INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS.................................

S 3/4 4-1 3/a.4.2 and 3/4.4.3-SAFETY VALVES.............................

B 3/4 4-1 3/4. 4. 4 PRESSURIZER........................................... B 3/4 4-2 3/4.4.5; STEAM GENERATORS...................................... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE........................ B 3/4 4 3/ 4. 4. 7 ' CH EMISTRY............................................ 3 3 / 4 4-4

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3/4.4.8 SPECIFIC' ACTIVITY....................................

3 3/4 4-4 3/4.4.9' PRESSURE / TEMPERATURE LIMITS........................... S 3/4 a-5 3/4.4.13 STRUCTURAL INTEGRITY.................................. 3 3/4 4-11 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4. 5.1. SAFETY INJECTION TANKS................................ B 3/4 5-i 3.4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS........................... B 3/4 5-1 3/4.5.4 REFUELING WATER TANY (RWT)...........................

3 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.................................. B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS................

3 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES........................

3 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L.............................

3 3/4 6 4 ARKANSAS - UNIT 2 X

0EFINITIONS EI - AVERAGE DISINTEGRATION ENERGY 1.19 II shall ~ be the average (weighted in proportion to the concentration

~

of each radionuclide in the reactor coolant at the time of sampling) of the sum-of the average beta and gamma energies per disintegration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

STAGGERED' TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist 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 b.

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

FREOUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for tne performance of-Surveil-lance Requirements shall correspond to the intervais cefined in Table 1.2.

AXIAL SHAPE INDEX 1.22 The AXIAL SHAPE INDEX shall be the power generated-in the lower half of the core less the power generated in the upper half of the. core l

divided by the sum of these powers.

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REACTOR TRIP SYSTEM RESPONSE TIME 1.23 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from -when the monitored parameter exceeds its trip setpoint at the I

channel sensor untti electrical power is interrupted to the CEA drive mechanism.

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- DEFINITIONS ENGINEERED SAFETY FEATURE RESPONSE TIME 1.24 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation -

setooint at the channel sensor until the ESF equipment is capable of performing its safety' function -(i.e., the valves travel to their ^

recuired positions, pump discharge pressures reach their required -

values,etc.). Times shall include diesel generator starting and sequer.ce loading delays where applicable.

PHYSICS TESTS 1.25 PHYSICS TESTS shall be those-tests performed to measure the funda -

mental nuclear characteristics of the reactor core'and related instrumen-tation and 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.

SOFTWARE 1.26 The digital computer SOFTWARE for the reactor orotection system shall be-the program. Codes including their associated data, documentation and procedures.

PLANAR RADIAL DEAKING FACTOR --Fxy 1.27 The' PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane average power density of the individual fuel rods in a given. horizontal plane, excluding the effects of azimuthal tilt.

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM ~ SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE DNBR

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2.1.1.1. The'DNBR of the reactor core shall be maintained 3,1.24.

l APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever J he DK R of the reactor core has decreased to less than 1.24 be l

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in HOT STANDBY within 1. hour.

DEAK LINEAR HEAT RATE 2.1.1. 2 The peak linear heat rate (adjusted for fuel rod dynamics) of the fuel:shall be maintained < 21.0 kw/ft.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the peak linear heat rate (adjusted for fuel rod.ynamics) of the fuel has exceeded 21.0 kw/ft, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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ARKANSAS - UNIT 2 2-1 Amendment No. 24 t

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i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS

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REACTOR COOLANT SYSTEM PRESSURE 2.1. 2 The Reactor Coolant System pressure shall not excead 2750 isia.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANOBY with the Reactor Coolant System pressure within its limit within i hour.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure nas exceeded 2750 osia, reduce the Reactor Coolant System pressure to within i'..

limit within 5 minutes.

~

ARKANSAS - UNIT 2 22

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 -LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 - The reactor protective instrumentation setpoints shall be set consistent with the : Trip Setpoint values shown in Table'2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a reactor protective instrumentation setpoint less conserative than the value shown in the Allowable Values column of Table 2.2-1 declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setooint adjusted consistent with the Trip Setcoint

value, i

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2.2 LIMITING SAFETY SYSTEM SETTINGS -(Continueo) ~

i!COREPROTECTIONCALCULATORADDRESS'ABLECONSTANTS 11, jr2.2.2 Core Protection Calculator Addressable Constants are defined in Table

+12.2-2.

Type I Addressable Constants are expected to change frequently during

! piant operation.

Type II Addressable Constant values are determined (or confirmed) during PHYSICS TESTS following each fuel loading and are not

expected to change during plant operation.

Changes to_ Type I Addressable.

I Constants outside the Allowable Value range require Plant Safety Connittee review prior to' implementation. Changes to : Type II Addressable Constants made other than as a result of post fuel loading PHYSICS TESTS shall recuire Plant Safety Committee review prior to implementation unless the changes are t

I. reclired for Technical Specification Comoliance.

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APPLICABILITY: As shown for Core Protection Calculators in Table 3.3-1.

l ACTION: With a Core Protection Calculator Addressable Constant found to be ll non-conservative, declare the cnannel inoperable and apply the~

II applicable ACTION statement recuirement of Specification 3.3.1.1

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until the channel is restored to OPERABLE status.

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/<nendment No. 2,{

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FUNCTIONAL UNIT.

TRIP-SETPOINI ALLOWABLE VALUES

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Manual Reactor Trip Not Applicable.

Not Applicable 2.

Linear Power Level - liigh a.

four Reactor Coolant Pumps i 110% of RATED THEIMAL POWER 1 110.712% of RA1ED TilERMAL POWE Operating b.

Three Reactor Coolant Pumps 1

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Two Reactor Coolant Ptsups Operating - Same Loop ro d.

Two Reactor Coolant Pumps I

En Operating - Opposite Loop:

3 Logarillunic Power Level -

High (1) 1 0.75% of itATED TilERMAL POWER 10.819% of RATED TilERMAL POWLR i

4.

Pre! _arizer Pressure - liigh t 2362 psia 3 2~J10.887 psia l

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Pressurizer Pressure - Low L 1766 psia (2) t 1712.757 psia (2) 3

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Containment Pressure - High 1

1 11.4 psla 1 19.024 psia.

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Steam Generator Pressure - Low t 751 psia (3)

L 729.6J3 psia (3)'

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Steam Generator Level - Low t 46.7% (4) t 45.8111 (4) l

These values lef t blank pending NRC approval of safety analyses for operation with l'ess than l

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TAllLE 2.2-1_(Continued)_

REAC.T..OR_ _P_RO I E CT I V_E__I.NS. l.R..U..M..I.N.I.A.I.l.ON l.it_l P_ SE I P.O I.N T._L.I..N_I T S k

IU.NCTIONAL UNIT litlP SEIP.0 INT; ALLOWA8LE VALUES i%

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Local Power Density - liigh 20.3 kw/ft (S) 1 20.3 kw/ft (S)

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10. DNOR - Low 2,-l.24 (S)

> 1.24 (5) l

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11.. Steam Generator' Level' - liigli 193.7% (4).

1 94.589% (4) 3 TABLE NOIATION 4

l (1) Trip may be manually bypassed above! 10 % of: RATED TilEletAL POWER; bypass steall be automatically

~4 4

removed when TilERMAL POWER is 1 10 of RAIED lilEllMAL POWER.

(2) Value' may be decreased manually, to a minimum value-of 100 psia, during a planned reduction in pressurizer pressure, provided the margin between the pressurizer pressure and this value is maintained i

at 1200 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until l

the trip setpoint is reached. Trip may be manually bypassed below 400 psla; bypass shall be automatically removed whenever pressurizer pressure is, > S00 psia.

(3) Value may be decreased manually during a planned reduction in steam generator pressure provided Llie l

l margin between the steam generator pressure and this value is maintained at 1 200 psi; tiie setpoint I

i shall be increased automatically as steam generator pressure is ircreased until the trip setpoint is reached.

i b

k (4) % of the distance betwcen steam generator upper and lower level lastrument nozzles.

a.

j (S) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoir' includes measure-f 5

ment, calquiational and processor uncertainties, and dynamic allowances.

z below 10- I of RATED TilERMAL POWER; bypass shall be automatically removed when TilERMAL POWER is > 10 T

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of RATED TilERMAL POWER.

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(f> ) The ministau-allowable value of the addressable constarit JiERR1 in each OPERABLE channel is 1.174. Upon f

NRC approval of the Statistical Combination of Uncertainties metliodology as described in CEN-139(A)-P.

l the minimum allowable value of HERRI is 1.055 i

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TABLE 2.2-2 l

l CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS i

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TYPE I ADDRESSABLE CONSTANT?

POINT 10 PROGRAM ALLOWABLE NUMBEL.__

LABEL DESCRIPTION VALUE 60 FCl-Core coolant mass flow rate eslibration

<l.15

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constant i

61 FC2 Core coolant mass flow rate calibration 0.0 constant 62 CEANOP CEAC/RSPT inoperable flag 0,1, 2 or 3 63 TR Azimuthal tilt allowance

>1.02 64 TPC Thermal power calibration constant

>0.90 i

l 65 KCAL Neutron flux power calibration constant

'>0.85 66 DNBRPT DNBR pretric setpoint Unrestrictec 67 LPDPT Local power density pretrip setpoint Unrestricted i

l ARKANSAS - UNIT 2 2-7 Amendment No. 2 4 1

,w

+g.,

,, - -, - ~,,,

a,.e

,-.-o,,.y e

i TABLE 2.2-2 (Continued)

CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS

-II.

TYPE II A00RESSABLE CONSTANTS

' POINT ID

~ PROGRAM NUMBER LAB _EL DESCRIPTION 58 BERRO Thermal power uncertainty bias 69 SERR1 Power uncertainty factor used in DNBR calculation 70-

.BERR2 Power uncertainty bias used in DNBR caletlation 71 BERR3 Power uncertainty factor used in local power density calculation 72 SERR4 Power uncertainty bias used in local, power density calculation 73 EOL End of life. flag 74 ARM 1 Multiplier for planar radial peaking factor g

i 75-ARM 2 Multiplier for planar radial peaking factor j

75 ARM 3 Multiplier for planar radial peaking factor 77 ARM 4 Multiolier for planar radial peaking factor 73 ARMS ttitiplier for planar radial peaking factor 79 ARM 6 Multiplier for planar radial peaking factor 80 ARM 7 Multiplier for planar radial peaking factor 81 SC11 Shape annealing correction lactor 82 SC12 Shape annealing correction factor I

i 83 SCl3 Shape annealing correction factor 84 SC21 Shape annealing correction factor 85 SC22 Shape annealing correction factor 86 SC23 Shape annealing correction factor 87 SC31 Shape annealing cor ection factor 88 SC32 Shape annealing correction factor ARKANSAS - UNIT 2 2-8 Amendment No. 2 4

TABLE 2.2-2 (Continued)

CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS II. TYPE II ADDRESSABLE CONSTANTS (Continued) t POINT ID PROGRAM-NUMBER LABEL DESCRIPTION 89

' SC33 Shape annealing correction factor 90 PFMLTD ONBR penalty factor correction multiplier 91 PFMLTL LPD penalty factor correction multiplier 92 ASM2 Multiplier for CEA shadowing factor 93 ASM3 Multiplier for CEA shadowing factor 94 ASM4 Multiplier for CEA shadowing factor 95 ASMS Multiplier for CEA shadowing factor 96l ASMS Multiplier for CEA shadowing factor 97 ASM7 Multiplier for CEA snacowing factor 93 CORR 1 Tancerature shadowing correction factor multiclier 99 SPPCC1 30undary coint power correlation coefficient 100 BPPCC2 Soundary coint cower correlation coefficient 101 BPPCC3 Boundary point power correlation coefficient-l 102 SPPCC4 Boundary point power correlation coefficient l

l l

l' I

l f

f i

l-

. ARKANSAS - UNIT 2 2-9 Amendment No. 24 gV w y

v

.-g fm-g m--

e4-1 v-ee-S-

+ -,L---w-

=

a m

-m,m.

~- +

r f'7

o 4

2.1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation ~ which would result ~1n the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by (1) restricting fuel. operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or.less than 21 kw/ft which will not cause fuel centerline melting in any fuel rod.

First, by operating within the nucleate boiling regime of heat.

transfer, the hsat transfer coefficient is large enough so that the maximum clad surface temperatt./e is only slightly greater than the coolant saturation temoerature.

The upper boundary of the nucleate boiling regime is termed "deoarture from nucleate boiling".(DNS). At this coint, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding.temoeratures and the possibility of cladding failure.

Correlation's predict DNS and the location of DMB for axially uniform and non-uniform heat flux distributions.

The local DNB ratio (DNBR), defined as the ratio of the predicted DNB heat flux at a par-ticular core location to the actual heat flux at that location, is indicative of the margin to DNB.

The minimum value of DNBR durino normal operational occurrences is limited to 1.24 for the CE-1 correlation l

and.is established as a Safety Limit.

Second, operation witn a. peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cladding l

integri ty.

Above this peak linear heat rate level (i.e., with some me' ting in the center), fuel rod. integrity would be maintained only if i

ne design and operating conditions are appropriate throughout the life of the fuel rods. Volume changes which accompany the solid to liouid phase change are significant and require accomodation. Another con-sideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.

Because of the above factors, the steady state value of the i

peak linear heat rate which would not cause fuel centerline melting is established as.a Safety Limit. To account for fuel rod dynamics (lags ),

the directly indicated linear heat rate is dynamically adjusted.

Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High ARKANSAS - UNIT 2 3 2-1 Amendment No. p 4

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

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Linear Power Level trips, and limiting -conditions for operation on DNBR and kw/ft margin are specified such that there is a nigh degree of confidence that the specified acceptable fuel design limits are not exceeded during.

normal operatien and design basis anticipated operational occurren~ces.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction ef this Safety Limit protects the integrity of the Reactor Coolant System from overpressuritation and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

.The Reactor Coolant System caloonents are designed to.Section III of the ASME Code for Nuclear Power Plant Components.

(The reactor vessels steam generators and p:essurizar are designed to the 1968 Edition, Summer 1970 Addenda; piping tJ the 1971 Edition, original issue; and the valves to the 1968 Edition, Winter 1970 Adcenda.Section III of this Code permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 osia to oemonstrate integrity prior to initial operation.

2.2.1 RIAC70R TRIP SETp0INTS The Ret: tor Trip Setpoints specified in Table 2.2-1 are the values at which the Teactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that tne reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to. assist the Engineered Safety Features Actuation System in mitigating the conssquences of accidents. Operation with a trip set less conserva-tive than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is eoual to or less than the drift allowance assumed for each trip in the safety analyses.

The ONBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.24 and-20.3 l

kw/ft, respectively. Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable l

Values for these trips are therefore the same as the Trip Setpoints.

g.

~

ARKANSAS - UNIT 2 B 2-2 Anendment No. 2 4 l

I

(

Y v

-.a x

.y SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Level-Low The Steam Generator Level-Low trip provides protection against a

. loss of feedwater flow incident and assures that the design pressure of the Reactor Coolant System will not be' exceeded due to loss of the steam generator heat sink. This specified setpoint provides allowance that there will be sufficient water inventory in the steam generator at the time of the trip to orovide sufficient margin before emergency feedwater is required.

Local Power Density-High The Local Power Density-High trip is provided to prevent the linear heat rate (kw/ft) in the limiting fuel rod in the core from exceeding the fuel design limit in the event of any anticipated ooerational occurrence.

The local power density is calculated in the reactor protective system utilizing the following information:

a.

Nuclear flux power and axial power distribution from the excore flux monitoring system; b.

Radial peaking factors from the position measurement for the CEAs; c.

ai power from rea-tor coolant temoeratures and coolant flow measurements.

The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynsnic compensation routines. These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the peak LPD Safety Limit.

CPC uncertainties related to peak LPD are the same types used for DNBR calculation. Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.

ARKANSAS - UNIT 2 3 2-5 Amendment No. 2 4

~

o SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES DNBR-Low The DNBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of anticipated operational occurrences. The DNBR - Lcw trip incor-parates a low pressurizer pressure floor of 1750 psia. At this pressure a DNBR - Low trip will automatically occur.- The DNBR is calculated in the CPC utilizing the following information:

a.

Nuclear flux power and axial power distribution from the excore neutron flux monitoring system; b.

Reactor Coolant System pressure from pressuri:er pressure measurement; c.

Differential temoerature (aT) power from reactor coolant temperature anc coolant flow measurements; p

d.

Radial ceaking factors from the cosition measurement for tne l

CEAs; i

e.

Reactor coolant mass flow rate frem reactor coolant pump speed;

[

f.

Core inlet temperature from reactor coclant cold leg temperature measurements.

The DNBR, the trip variable, calculated by the CPC incorporates various uncertainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits. These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core DNBR is sufficiently greater than 1.24 such that the l

decrease in actual core ONBR after the trip will' not result in a viola-tion of the DNBR Safety Limit. CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties.

Dynamic comperaation is providsd in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.

The ONBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limite will result in a CPC initiated trip.

ARKANTAS - UNIT 2 8 2-6 Amendment No. 24 m3, s

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, - - - -,---eyr,-

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. 1 SAFETY LDtITS AND LIM' f dG SAFETY SYSTEM SETTINGS BASES a.

~RCS Cold Leg Temperature-Low

> 465'F b.

RCS Cold Leg Temperature-High I605'F c.

Axial Shape Index-Positive Not more positive than +0.6 d.

Axial Shape Index-Negative Not more negative than -0.6 e.

Pressurizer Pressure-Low

> 1750 psia g.

Integrated Radial Peaking

. [2400 psia f.

Pressurizer Pressure-High Factor-Low

> 1.28

~

h.

Integrated Radial Peaking

' Factor-High 5,4.28 1.

Quality Margin-Low

>0 Steam Generator Level-High The Steam Generator Level-High trip is provided to protect the turbine from excessive moisture carry over. Since the turbine is auto-matically tripoed when the reactor is tripped, this trip provides a reliable means for oroviding protection to the tureine from excessive moisture carry'over. This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for ocer-ation of this trip.

Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

2.2.2 CPC Addressable Constants The Core Protection Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate indications such as calorimetric measurenents for power level and RCS flowrate and incore detector signals for axial flux shape, radial peaking factors and CEA deviation penalties. Other CPC addressable' constants allow penalization of the calculated-ONBR and LPD values based on measurement uncertainties or inoperable equipment. Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3.1.1 and 6.8.1) ensures that inadvertent misloading is unlikely.

The methodology for detennination of CPC addressable constant values is described in AP&L letter 2CAN058113 dated May 26, 1981.

t ARKANSAS - UNIT 2 3 2-7 Anendment No. 2 4

(

+-,..,n.-.,,.n..,.rm, n-4,n,.e e

g--,e n

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REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T,yc 1 00*F 2

LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTOOWN MAF GIN shall be t 5.0%.ak/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN < 5.0% ak/k. fmmediately initiate and continue

[

baration at > 40.gpm of 1731 ppm boric acid solution or equivalent until the requitec SHUTDOWN MARGIN is restored.

SURVEILLANCE REOUIREMENTS 4.1.1. 2 The SPJTDOWN MARGIN shall be determined to be t 5.05 sk/k:

l a '.

Within one hour after detection of an inoperable CEA(s) are at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is 4-,serable.

'If the inoperable CEA is immovable or untripcable, the i.bove required SHUTDOWN MARGIN shall be increased by an amount at least ecual to the withdrawn worth of the immovable or untrio-pable CEA(s).

b.

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

1.

Reactor coolant system boron concentration, 2.

CEA position, 3.

Reactor coolant system average temperature, 4-Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

.iamarium concentration.

ARKANSAS - UNIT 2 3/4 1-3 Amendment No. 24 s'

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MM y

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REACTIVITY CONTROL SYSTEMS BORON OILUTION LIMITING' CONDITION FOR OPERATION 3.1.1. 3 The flow rate of reactor coolant through.the _ reactor' coolant system shall be 3 3000 gpm whenever a reduction 'in Reactor' Coolant

. System boron conc,entration is being made.

APPLICABILITY: ALL MODES.

ACTICN:

With the flow rate of reactor _ coolant th ough the reactor coolant system

< 3000 gpm, immediately suspend all operations involving a reduction in baron concentration of the Reactor Coolant System.

' SURVEILLANCE REOUIREMENTS '

.&.1.i.3 The flow rate of reactor coolant througn the reactor ' coolant system snall ce determined to be 3,3000 gpm within one nour prior to tne start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either:

a.

Verifying at least one reactor coolant pump is in operaSion, or b.

Verifying that at least one low pressure safety injection pump is in operation and supplying 3,3000 gpm through the reactor coolant system.

2 ARKANSAS - UNIT 2 3/41-4 i

l f'

y

. s A

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT

-l LIMITING" CONDITION FOR OPERATION L!.

l 3.1.1.4.. Tne moderator-temperature coefficient (MTC)-shall' be:

a.

Less positive than 0.5x10" tk/k/ F whenever THERMAL I

POWER is <70% of RATED THERMAL POWER, f

I ll

-b.

Less oositive than 0.0 ak/k/ F whenever THERMAL POWER U

((

.is >70% of ' RATED THERMAL POWER, and

!I c.

Less negative than -2.8x10

' Ak/k/ F at RATED THERMAL

[! '

' POWER.

j ',

l APPLICABILITY:

MODES 1 and 2*.

i

l: ACTION:

j ; With the moderator temperature coefficient outside any one of the above ll limits, be 'in at. least HOT STANDBY witnin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS

!t.l'.i.A.i The MTC snall be determined to be within its limits by confirmatory

l measurements. MTC measured values shall be extrapolated and/or compensated

to permit direct comparison with the above limits.

l!'4.1.1.4.2 The MTC shall be detemined at the following frequencies and THERMAL l[ POWER conditions during each fuel cycle:

a.

Frior to initial operation above 5%-of RATED THERMAL POWER, jj after each fuel load'ing.

il.l!

At any THERMAL POWER. within 7. EFPD after reaching a RATED o.

THERMAL POWER equilibrium boron concentration of 800 ppm.

- l c.

At any THERMAL POWER, within 7 EFPD after reaching a RATRD 1,

THERMAL POWER ecuilibrium boron concentration of 300 ppm.

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- f$+41thX,ff>1.0.

il.ll ARKANSAS - UNIT 2 3/4 1-5

/cendment No. 2 4

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t REACTIVITY CONTP.0L SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR CPERATION 3.1.1.5.The Reactor Coolant System lowest operating loop temperature (Tavg) shall be > 525'F when the reactor is critical.

APPLICABILITY: MODES 1 and 1!*.

ACTION:

With a Reactor Coolant System operating loop temperature (Tto within its limit within 15 mir 525'F, restore T f

HOTSTANDBYwithiX9the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.1.5 The Reactor Coolant System temoerature (T#V9) shall be determinec*

.t:.be 3,525'F:

a.

Within 15 minutes prior to acnfeving reactor criticality, anc b.

At least.once per 30 minutes wnen the reactor is critical and the Reactor Coolant System T is less than 535'F.

avg l

'Wi th K,ff,1.0.

3

(

"See Special Test Exception 3.10.5.

I l-ARKANSAS - UNIT 2 3/4 1-6 l

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

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. REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN--

LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:

a.

A flow path from the bcric acid makeup tank via either a boric acid makeup pump or a gravity feed connection and

~ charging pump to tne Reactor Coolant System if only the boric acid makeup tank in Specification 3.1.2.7a is OPERABLE, or b.

The flow path from the refueling water tank via either_ a cnarging pump or a high pressure safety injection pump to the Reactor Coolant system if only the refueling water tank in Specification 3.1.2.7b is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With none of tne aoove flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demon-strated OPERABLE:-

a.

At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path is above the temperature limit line shown on Figure 3.1-1 when a flow path from the l

boric acid makeup tanks is used.

b.

At least once per 31 days by verifying that each valve (manual.

(

power operated or automatic) in the flow path that is not l

. locked, sealed, or otherwise secured in positica, is in its l

correct position.

I ARKANSAS - UNIT 2 3/4 1-7 l

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:

a.

Two flow paths from the boric acid makeup tanks via either a boric acid makeup pump or a gravity feed connection, and a charging pump to the Reactor Coolant System, and b.

The flow path from the refueling water tank via a charging Dump to the Reactor Conlant System.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72

hours or ce in at least HOT STANDBY and borated to a SHUTDOWN MARGIN i equivalent to at least 5% tk/k at 200*F within the next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ; restore at least two flow catns to OPERABLE status aithin the next 7 days or ce 2

in' COLD SHUTDOWN witnin :ne next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.1. 2. 2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that the temperature of th. heat traced portion of the flow path from the boric acid makeup tanks is above the temperature limit line shown on Figure 3.1 -1.

b.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c.

At least once per 18 months during shutdown by verifying that each actuated <3ive in the flow path actuates to its correct position on a SIAS test signal.

ARKANSAS - UNIT 2 3/4 1-8 Anendment No. 24

e

}

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging. pump in the boron injection flow catn required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered Trom an OPERABLE emergency bus.

APPLICABILTTY: MODES 5 and 6.

ACTION:

With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one of the required pumps is restored to OPERABLE status.

SURVE:LLANCE REGUIREMENTS 4.1.2.3 ' No additional Surveillance Requirements otner than those required by Specificati~on 4.0.5.

ARKANSAS

' UNIT 2 3/4 1-9 u---.

o

)

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1. 2. 4 At' least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in.at least HOT STANDBY and borated to a SHUTOOWN MARGIN equivalent to at least 5% ak/k at 200*F l

witnin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to ' OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i si 0:

h SURVEILLANCE REQUIREMEffrS 4.1.2.4 No additional Surveillance Requirements other than those required by Specification 4.0.5.

ARKANSAS - UNIT 2 3/4 l-10 Amendment No. p 4

o REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 At least one boric acid makeup pump shall be OPERABLE and

apable of being pow? red from an OPERABLE emergency bus if only the flow path through the boric acid makeup pump in Specification 3.1.2.la above, is.0PERABLE.

APPLICABILITY: ~ MODES 5 and 6.

ACTION:

With no boric acid makeup pump OPERABLE as required to complete the flow path of Specification 3.1.2.la, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one boric acid makeup pump is restored to OPERABLE status.

SURVEILLANCE RE0VIREMENTS 4.1.2.5 No aeditional Surveillance Requirements other than those required by Specification 4.0.5.

i.

l l

ARKANSAS - UNIT 2 3/4 1-11 e

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0 REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS riRATING LIMITING CONDITION FOR OPERATION ~

3.1.2.6 At least the boric acid makeup pumo(s) in the baron injection flow path (s) required OPERABLE pursuant to Spe:ification 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path througn the boric acid makeup pump (s) in Specification 3.1.2.2a is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

With one boric acid makeup pump required for the baron injection flow path (s) pursuant to Specification 3.1.2.2a inoperable, restore the boric acid makeuo pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN eouivalent to at least 5% ak/k at 200'F; restore the above requi*ed beric acid pumo(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.1.2.5 No additional Surveillance Requirements other than those required by Specification 4.0.5.

ARKANSAS - UNIT 2 3/4 1-12 knendment No. 24 Y

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REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 E'ch of the following barated water sources shall be OPERABLE:

a.

At least one boric acid makeup tarik and one-associated heat tracing circuit per tank with the contents of the tank-in accordance with. Figure 3.1-1, and b.

The refueling water tank with:

1.

A contained barated water v31une of between 464,900 and 500,500 gallons (equivalent to an indicated tank level of between 91.7% and 100%, respectively),

2.

'Between 1731 and 2250 pom of boron, 3.

A minimum solution temoerature of 40*F and 4.

' A maximum solution temaerature of 100*F.

APPLICABILITY: MODES 1, 2, 3 and ?.

J ACTION:

i a.

With tne above required boric acid makeup tank inocerable, U

restore the make up tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated i

to a SHUTDOWN MARGIN equivalent to at least 5 % ak/k at 200*F; l

l restore-the above required boric acid makeup tank to OPERABLE status within ene next 7 days or be in COLD SHUTDOWN within l

the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l l

b.

With the refueling water tank inoperable, restore the tank to

- OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 nours.

SURVEILLANCE REQUIREMENTS l

l 4.1.2.8 Each of the above required borated water sources shall be demonstrated OPERABLE:

t ARKANSAS - UNIT 2 3/4 1-15 Amendment No.ph 4

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REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) a.

At least one per 7 days by:

1.

Verifying the baron concentration in each water source, 2.

Verifying the contained barated water volume in each water source, and 3.

Verifying the boric acid makeup tank solution temperature.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verify'ing the R'4T temperature.

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o ARKANSAS - UNIT 1 3/4 1-27 Amendment No. 2 4

.. -. -.. -. -. -. - -. -. -. -. ~.. -. - - -. -.. -.

3[4.2 POWER DISTRIBUTION LIMITS t

LINEAR ' HEAT RATE' LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate margin shall be maintained by operating within th.e region of acceptable operation of Fiqures 3.2-1 or 3.2-2 as acolicable.

APPLICABILITY: MODE I above 20% of RATED THERMAL POWER.

ACTION:

With the linear heat rate exceeding its limits, as indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on kw/ft; or (2) when the COLSS is not being used, any OPERABLE Local Power Density channel exceeding the linear heat rate limit, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:

a.

Restore the linear heat rate to within its limits within one hour, or b.

Be in at least HOT STANO3Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l SURVEILLANCE REQUIREMENTS I

l 4.2.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.1.2 The linear heat rate snall be determined to be w' thin its l

limits when THERMAL POWER is above 20% of RATED THERMAL POWER by con-l tinuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that the linear heat rate, as indi-cated on all CPERABLE Local Power Density channels, is within the limit E

shown on Figure 3.2-1.

I 4.2.1.3 At least once per 31 days, the COLSS Margin Alarm shall be i

verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on kw/ft.

/

l ARKANSAS - UNIT 2 3/4 2-1 Amendment No. 2 4

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.r-Figure 3.2-2 Allowalile Peak Linear lleat. Rate vs llursiup (t:01.% nial of service) i 1

1 1

l 1

f POWER DISTRIBUTION LIMITS RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION 3.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F* ) shall be less th'an or y

equal to the PLANAR ' RADIAL PEAKING FACTORS (F@C ) used in the Core Limit Supervisory System (COLSS) and in the C e Protection Calculators (CPC).

APPLICABILITY:

MODE I above 20t of RATED THERMAL-POWER.*

ACTION:

exceedingacorresponding(within6hourseither:

I With a F Adjust the CPC addressable constants to increase the multiplier a.

applied to PLANAR RADIAL PEAKING FACTOR by a factor equivalent f

/(y and restrict subsequent operation so that a margin to >F to the COLSS operating limits of at least ((F* / FC ) - 1.0] ' x 1005 ll is maintained; or

/

Y ff Adjust the affected PLANAR RADIAL PEAKING FACTORS (Fjy) used in l

b.

the COLSS and CPC to a value greater than or equal to the measured PLANAR RADIAL PEAKING FACTORS (F ); or l

c.

Se in at least HOT STANDBY.

l'

! SURVEILLANCE REOUIREMENTS l3tI!! 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

Si l

4.2.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F

), obtained by l

using the incore detection system, shall be determined to be less than C

or equal to the PLANAR RADIAL PEAXING FACTORS (F y) used in the COLSS l

and CPC at the following intervals:

a.

After each fuel loading with THERMAL POWER greater' t.ian 405 but prior to operation above 70t of RATED THERMAL POWER, and

- b. ' At least once per 31 days of accumulated operation in MODE 1.

See Special Test Exception 310 2 I

ARKANSAS - UNIT 2 3/4 2-4 Amendment No. 24

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POWER DISTRIBUTION LIMITS AZIMUTHAL POWIR T!;.T - T, LIMITINL CONDITION FOR OPERATION-3.2.3 'The AZIMUTHAL POWER TILT (T ) shall be less than or ecual to 9

Ine AZIMUTHAL POWER TILT Allowance used in the Core Protection Cal-i cuiators (CPCs).

~

APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.*

ACTION:

a.

With the measured - AZA.'4UTHAL POWER TILT detemined to. exceed the AZIMUTHAL POWER TILT Allowance used in the CPCs but

< 0.10, withir, two hours either correct the power tilt or adjust the AZIMUTHAL-POWER TIL7 Allowance used in the CPCs to greater than er &cual to the measured value, b.

With the measured AZIMUTHAL POWER TILT detemined to exceed 0.10:

'l p

1.

Due to misalignment of either a part length or full knqtn CEA, within 30 minutes verify that the Core Operating Limit Supervisory System (COLSS) (when COLSS is being used to monitor the core power oistribution oer

~

Specifications 4.2.1 and 4.2.4) is detectirJg the CEA minaiignment.

I 2.

Verify that the AZIMUTHAL POWER TILT is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the p

next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Linear Power Level - High j

trip setpoints ~o < 55% of RATED THERMAL POWER within the l

next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

[

3.

Identify and correct the cause of the out of limit con-dition prior to increasing THERMAL POWER; subsequent DOWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the AZIh0 THAL POWER TILT is verified within L

its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or graater RATED THERMAL POWER.

l.

"See Special Test Exception 3.10.2.

l ARKANSAS - UNIT Z 3/4 2-5 Amendment No.. g il i^

POWER DISTRIBUTION' LIMITS SURVEILLANCE REQUIREMENTS 4.2.3 ~.The AZIMUTHAL. POWER TILT shall be detennined to be within the limit above 20". of RATED THERMAL POWER by:

a.

Continuously monitoring the tilt with. COLSS when the:COLSS is OPERABLE.

5.

Calculating the tilt at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the COLSS

.is. inoperable.

c.

. Verifying at least once per 31 -days, that the COLSS Azimuthal Tilt Alarm is actuated at an AZIMUTHAL POWER TILT greater than the AZIMUTHAL POWER TILT Allowance used in the CPCs.

d.

'Using the incors detectors at least once oer 31 days to independently confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT.

e

' ARKANSAS - UNIT 2 3/42-6 Amendment No. 2 4 l

1

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

POWER DISTRIBUTION LIMITS DNBR MARGIN i

LIMITING CONDITION FOR OPERATICN

~

3.2.4 -The'DNBR margin shall be maintained by operating within the region of acceptable operation of Figure 3.2-3 or 3.2 as applicable.

l APPLICABILITY: MODE I above 20% of RATED THERMAL POWER.

ACTION:

With operatien outside of the region of acceptable operation, as indicated by'either (1) the COLSS~ calculated core power exceeding the COLSS calculated core power operating limit based on DNBR; cr (2) when the COLSS is not being usea, any OPERABLE Low DNBR channel exceeding the DNBR~ limit, within 15 minutes initiate corrective action to reduce the DNBR to within the limits and either:

a.

Restore the DNBR to within its limits within one hour, or b.

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

SURVEILLANCE REQUIR94ENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.

4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit l

Supervisory System (COLSS) or, with the COLSS out of service, by verify--

ing at least once per 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s-that the DNBR, as indicated on all OPERABLE DNBR channels, is ~within the limit shown on Figure 3.2-3.

t l

4.2.4.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on DNBR.

1

~

i l

, ARKANSAS - UNIT 2 3/4 2-7 Anendment No. 2 4 1

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

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.4.4 _ The following DNBR penalty factors shall be verified to.be included in the COLSS and CPC ONBR calculations at least once per 31 days:

GWD Burn'up(MTU)

DNBR Penalty (%)

0-3.1 0

3.1-5 2.0 5-10 5.9 10-15 8.8 15-20 11.4 20-25 13.5

_25-30 15.5 30-35 17.4 l

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POWER DISTRIBUTION LIMITS

'RCS-FLOW RATE LIMITING CONDITION FOR OPERATION ~

i 3.2.5' The actual Reactor Coolant System total flow rate shall be greater than or equal to 120.4 x 108 lbm/hr.

i APPLICABILITY: MODE 1.

ACTION:

'With -the actual Reactor Coolant System total flow rate determined to be

-less than the above limit, reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5 The actual Reactor Coolant System total flow rate snall be deter-mined to ce witnin its limit at least once ;:er 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i l

l f

l

[

l ARKANSAS - UNIT 2 3/4 2-11 Amendment No. 24 y

i

... -, - - - -.......,,.,, -,. - ~ ~

1 ll

POWER OISTRIBUTION LIMITS i

REACTOR COOLANT COLD LEG TEMPERATUR:

LIMITING CONDITION FOR OPERATION lt

{ 3.2.6~ The geactor Coolgnt-Cold Leg Tenperature (Tc) shall be maintained between 542 F and 554.7 F.

APPLICABILITY: MODE 1.above 30% of RATED THERMAL POWER.

ACTION:

'l With the' Reactor Coolant Cold. Leg' Tenperature exceeding its limits, restore-l the tenperature to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER f j ' to.less than 30% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

i i

l 3

.i IfSURVEILLANCEREOUIREMENTS

l I'

I-l;4.2.6 The Reactor Coolant Cold Leg Temperature shall be detennined to be

ll within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I:

t li l

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ARKANSAS - UNIT 2 3/4 2-12 Amendment No. 2 4 l

l l

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POWER DISTRIBUTION LIMITS.

- l AXIAL SHAPE INDEX LIMITING CONDITION' FOR OPERATION

'I

.2.7' The' core average AXIAL SHAPE INDEX (ASI) shall be maintained within i

3

' the following limits:

l

a. -COLSS OPERABLE

-0.28 < ASI < + 0.28 I

b.

COLSS OUT OF SERVICE (CPC) 1

-0.20 < ASI < +0.20-I APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER

i i

' ACTION:

With the core average AXIAL SHAPE INDEX (ASI) exceeding its limit, restore.

,! the ASI to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to

!!1ess than 20% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

fl l

i.

+.

!!SURVEILLANCEREOUIREMENTS t.!!-

It4.2.6 The core average-AXIAL SHAPE INDEX shall be determined to be within

!! its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months using the COLSS or any OPERABLE Core l

Protection Calculator channel, t

i l-

!l o

f I

i

! *See Special Test Exception 3.10.2.

~

L f

't

l',

! ARKANSAS - UNIT 2 3/4 2-13 Amendment No. 0 4 1

~

m

., - -~

,..w.,-.

,,.y.,

e,,

--e,v-..

m,,,,.,

p._,,y p,.n,,%,

.,,,,.y e -

e,9

,,wy-..,y,

.-g,

-p-

,,,p-----.%,

p-

POWER DISTRIBUTION LIMITS

_l PRESSURIZER PRESSURE t'

LIMITING CONDITION FOR OPERATION

-i l3.2.8 The average pressurizer pressure shall be maintained ' between 2225 psia Iand 2275 psia.

APPLICABILITY:

MODE 1 ACTION:

With-the average pressurizer pressure exceeding its limits, restore the

tenperature to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER l-l to less than 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

'l,

-1 I

,! { SURVEILLANCE REQUIREMENTS -

. ! 4.2.6 The av erage~ pressurizer' pressure shall be de,termined to be within its I flimit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

'l l

L t:\\'li-t s

c l

.?

l

'l I

i

(;

ARKANSAS - UNIT 2 3/4 2-14 Amendment No. 2 4 l

L l

I

INSTRUMENTATION SURVEILLANCE REOUIREMENTS (Continued) 2.

With 120 volts AC (60 Hz) applied for at least 30 seconds across the input, the reading on the output does not exceed 8 volts DC.

b.

For'the optical isolators: Verify that the input to output insulation resistance is greater than 10 megohms when tested using a megohmmeter.on the 500 volt DC range.

4.3.1.1.5 The Core Protection Calculator Systen shall be determined OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that less than three auto restarts have occurred on each calculator during the past 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.3.1.1.6 The Core Protection Calculator System shall be subjected to a CHANNEL-FUNCTIONAL TEST to verify OPERABILITY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of receipt of a valid High CPC Room Temperature alarm.

I ll' w

ARKANSAS - UNIT 2 3/4 3-la Amendment No. e.1 1

1 2

IAlli0 3;3-1 REACTOR PR0llCIIVE INSTRUMENTATION h-MINIMLM j._

TOTAL. NO.

CilANNELS CilANNELS.

APPLICABLE FUNCTIONAL UNIT OF CllANN[i.S 10 TRIP OPERABLE.

MODES

ACTION

-4 m

'1.

Manual Reactor Trip 2 sets of 2 1 set of 2 2 sets of 2 1, 2 and

1 2.

Linear Power Level - liigh 4

2 3

~ 1, 2 2f I

3.

Logaritimilc Power Level-liigli j

h.

Shutdown 4

.2(a)(d) 3-2 and

2#

l a.

Startup and Operating 4

0 2-3, 4, 5 3

4.

Pressurizer Pressure - liigli 4

2 3

1, 2 2#

1 u2 5.

Pressurizer Pressure - Low 4

2(b) 3 1, 2 and

2f l

f.

6 Contatsmient Pressure - liigli 4

2 3

1, 2 2d m

7.

Steam Generator Pressure - Low 4/SG 2/SG 3/SG 1, 2 and

2#.

l 8.

Steani Generator Level - Low 4/SG 2/SG 3/SG.

1, 2 2f y

l g

9.

L'ocal Power Density - liigh 4

2(c)(d) 3 1, 2 27 1

2a O

1 4

to j

u 1

4 i

i

t t

TABLE 3.3-1_(Cont.inued).

REACTOR-PROILCiIVE IhSlRUMENTAT10N' 5

5; a

MINIMUM i

=

TOTAL NO.

CilANNELS '

CilANNELS APPLICABLE U

FUNCTIONAL UNIT OF CllANNELS TO 1 HIP OPERABLE MODES ACTION t

10. DNDR - Low 4

2(c)(d) 3 1, 2 2#

11. Steam Generator Level - liigh 4/SG 2/SG

'3/SG

- 1, ' 2

'2#

12. Reactor Protection Systan Logic 4

2 4

1, 2 and

4 l

13. Reactor Trip Breakers

-4(f) 2 4

1, 2 and * '

4

14. Core Protection Calculators 4

'2(c)(d) 3 1, 2 2f and 6

{'

15. CEA Calculators 2

1 2(e).

1, 2 Si and 6 Y

w l

E" a

o i

l t,

a i

i i

?

i

4 TABLE 3.3-1 (Continued)

TABLE NOTATION With the protective system trip breakers in the closed position and -

the CEA drive system capable of CEA withdrawal.

The provisions of Specification 3.0.4 are not applicable.

-(a) Trip may be manually byoassed above 10-4% of RATED THERMAL POWER; 1

bypass shall.be automatically removed when THERMAL POWER.is < 10'4".

of RATED THERMAL POWER.

(b) Trip may!be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is > 500 psia.

(c) Trip may be manually bypassed below 10-4% of RATED THERMAL. POWER; bypass shall be automatically. removed when THERMAL POWER'is > 10-4%

of RATED T14ERMAL ' POWER. During testing pursuant to Special Test Exception 3.10.3, trip may be manually bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL -

POWER is > 1% of RATED THERMAL POWER.

(d). Trio may be bypassed during testing pursuant to Special Test Exceo-tion 3.10.3.

(c) See Special Test Excection 3.10.2.

(f) Each channel shall be comprised of two trip breakers'; actual trip logic shall be one-out-of-two taken twice.

ACTION STATEMENTS

~

With the number'of channels OPERABLE one less th'an ACTION l~

required by the Minimum Channels OPERABLE requirement.

l

- restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 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 and/or open the protective system trip breakers.

ARXANSAS-UNIT 2 3/4 3-4 f-l l

l

.~

.1

]

TABLE 3~.3-1 (Continued)

ACTION STATEMENTS b.

With both CEACs inoperable, operation may continue provided that:

1.

LWithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the margins required by Specifi-cations 3.2.1 and 3.2.4 are increased and main -

tained at a value equivalent to > 11". of RATED THERMAL-POWER.

2.

Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a)

All full length and part length CEA groups are withdrawn _ to and subsequently main-tained at the " Full Out" position, except during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2 or for control when CEA group 6 may be inserted no further than 127.5 inches withdrawn, b)

The "RSPT/CEAC Inocerable" addressable constant in the CFCs is set _to the 'inoper-able status.

c)

The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in tne "Off" mode except during CEA group 6 motion permitted by a) above, when the CEDfiCS may be operated in either the " Manual Group" or " Manual Individual" mode.

i I

3.

At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , all full length and part length CEAs are verified fully withdrawn l.

except during surveillance testing pursuant to Specification 4.1.3.1.2 or during insertion of CEA group 6 as permitted t'y 1. a) above, then verify at'least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months that the inserted CEAs are aligned within 7 inches (indicated position) of all other CEAs in its group.

ACTION 6 With three or more auto restarts of one non-bypassed 1-calculator during a 12-hour interval, demonstrate calculator OPERABILITY by perfoming a CHANNEL FUNCTIONAL TEST within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ARKANSAS - UNIT 2 3/4 3-5a Amendment No. 2 4

-l t

E

I g

1 Al!L,[ 3,,.3-2.-

'D

.g REACTOR' PROTEC.._T_I_V_E..IN. S..TRUM._EN.T_ATION RESPONSE TIMES

-se i

j E

FUNCT!0NAL UNIT

~ RESPONSE TIME d

N 1.

Manual Reactor Trip Not Applicable:

t

2. - Linear Power Level - liigh

< 0.40 seconds

3.

Logaritimilc Power Level.- liigh

.._ 0.40 seconds

4.

Pressurizer P. essure.liigh

< 0.90' seconds i

j S.

Pressurizer Pressure - Low

< 0.90 seconds 1

i 6.

Containment Pressure - liigh

< 1,59 seconds 7.

Steam Gencrator Pressure - Low

< 0.90 seconJs-i

8.. Steam Gencrator Level - Low

< 0.90 seconds l-9.

Local Power Density - liigh j

a.

Ileutron Flux Power from Excore Neutron Detectors

< 2.58 seconds

b.

CEA Positions j1.58 seconds **

l l

l i

i i

t e

I 1-l

{

JABLE 4.3-1, 2

j-

{

' REACTOR PROTECTIVE -INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

c 2

CilANNEL MODES IN WillCil C

CilANNEL CilANNEL.

FUNCIIONAL SURVEILLANCE m

FUNCTIONAL UNIT CHECK CA!IBRATION TEST REQUIRED.

l 1.

Manual Reactor Trip N.A.

N.A.

S/U(1)

N.A.

i 2.

Linear Power Level - liigh S

D(2,4),M(3,4),

M.

1, 2 ~

Q(4) 3.

Logarithailc Power Level - liigli S

R(4)

M and S/U(1) 1,2,3,4,5

{

.and *'

l

{

4.

Pressurizer Pressure - liigh S

R M

1, 2 7,

S.

Pressurizer Pressure - Low S

R M

1, 2 and

l 6.

Containment Pressure - liigli S

R M

1, 2 7.

Steam Generator Pressure - Low S

R-M 1, 2 and

l g

8.

Steam Generator Level - Low S

R H

1, 2 9.

Local' Power Density - High S

D(2,4),R(4,5) M,R(6) 1, 2 i

$o t3 j

u i

i i

i g

I TAllt.E 4.3' l (Continuedl REACTOR PROlECTIVE INSlRullfMTATION SURVEILLANCE REQUIREMENTS i

Cj CilANNEL FUNCT10NAL' SURVEILLANCE I

CilANNEL~

MOUES IN WillCH

. CilANNi'L-FilNCTIONAL UNIT

..C.il_E C.K CALIBRATION TEST REQUIRED e

I 2

q

10. DNBR - Low

.S S(7)',IX2,4), M,R(6) 1, 2 :

M(8),R[4.5) m

11.,Steaiin Gelierator 1.evel - Iligli S

R-M 1,- 2

12. Reactor Protection System Logic.

.H.A.

H.A.

JM 1,2 and

4

13. Reactor Trip Dreakers N.A.

N.A.

M 1,2 and *'

14. Core Protection Calculators S W(9)

D(2,4 ),R(4,5) M,R(6) 1, 2 l

15. CEA Calculators S

R M,R(6)'

1, 2 i.

[

E 3-e a

Ja 1

4 i

4 3

2 TABLE 4.3-1 -(Continued)

TABLE NOTATIONS

-.With: reactor trip' breakers in the closed position and the. CEA drive system capable of CEA withdrawal.

(1) '

Iff not performed in previous -7 days.

~

-(2) - Heat balance only (CHANNEL FUNCTIONAL TEST not included), above 154 of RATED THERMAL POWER; adjust the Linear Power Levtl signals and

- the CPC addressable ' constant multipliers to make the CPC aT power and CPC. nuclear power. calculations agree with the calorimetric calculation if absolute difference is > 2%.

During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed _ upon reaching each major test power plateau and prior to proceeding _to the next major test power plateau.

_(3) - Above 15% of -RATED THERMAL POWER, verify that the linear power sub-channel gains-of the excore detectors are consistent with the values used _ to establish the shape annealing matrix elements in the Core Protection Criculators. -

-(a) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

_(5) - After each fuel loading and prior to exceeding 70t of RATED THERMAL POWER, -the incore detectors shall be used to ' determine the shape annealing matrix elements and the Core Protection Calculators shall use these elements.

(6) - This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into' tne channel as close to the sensors as practicable to verify OPERABILITY including alarm and/or trip functions.

.(7) - Above 70% of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by' either using the reactor coolant pump differen-tial ' pressure instrumentation (conservatively compensated for measure-ment uncertainties) or by calorimetric calculations (conservatively compensated for measurement uncertainties) and if necessary, adjust tne CPC addressable constant flow coefficients such that each CPC

indicated flow is less than or equal to the actual flow rate. The

-flow measurement uncertainty may be included in the BERR1 term in

the CPC and is equal to or greater than 4%.

(8) - Above~ 70% of RATED THERMAL POWE2, verify that the totai RCS flow rate as: indicated by each CPC is less than or equal to the actual RCS total flow rate determined by calorimetric calculations (conservatively compensated for measurenent uncertainties).

' (9) - The' correct values of addressable constants (See Table 2.2-2) shall be verified to be installed in each OPERABLE. CPC.

ARKANSAS - UNIT 2 3/4 3-9 knendment No. pg

~

4

1 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety-Feature Actuation System (ESFAS). fnstru-mentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistant with the values shown in the Trip Setpoint column of Tab'e 3.3-4 and with RESPONSE TIMES as shown

~

in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a.

With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare 'the channel inoperable and apoly the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with.the trip set-

. point adjusted consistent with the Trip Setcoint.value.

b.

With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table -3.3-3.

SURVEILLANCE REQUIREMENTS r

4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES eed at the frequencies shown tn Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limf t at least once per 18 months. Each test shall include at least one channel per function such that all enannels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

~

ARKANSAS-UNIT 2

-3/4 3-10

-l

TABLE 3.3-3 -(Continued)-

TABLE NOTATION ACTION 10 -

- With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed conditien and the Minimum Channels OPERABLE requirement is demon-strated'within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; one additional-channel may bs bypassed or placed in-the tripped condition for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.1.

a ARKANSAS - UNIT 2 3/4 3-15

l_Aull 3.3-4 g

ENGINEEREU SAFETY FEATURE lA_C10All0N SYSTEM, INS 1RlMENTA_ TION TRIP VALUES E

g ALLOWABLE FUNCTIONAL UNIT-1 RIP,SEIP0lNT

__ VALUES _

l E

1.

SAFETY. INJECTION (SIAS).

j y

a.

Manual (Trip Buttons).

Not Applicable

Not Applicable m

.b.

Containment Pressure - liigli

. < 111.4 psia

<'19.024 psia' i

i c.

Pressurizer Pressure - Low t l/66 psia (1)

.? 1712.757 psia (1) i 2.

CONTAINMENT SPRAY (CSAS) l a.

-Manual-(Trip Buttons)

Not Applicable

.Not Applicable' y

b.

Contaltunent Pressure ---liigh-liigh

< 23.3 psia

< 23.624 psia e

3.

CONTAINMENT ISOLATION (CIAS)

{

a.

Manual (Trip Buttons)

Not Applicable-Not Applicable i

w j

S b.

Contaltunent-Pressure - liigh -

< 18.4 psia

< 19.024' psia w

k m

Si n

I 4

l i

4 i

i 3

5

-e 1

i 2

' TABLE 3.3-4 (Continue'd) i y

ENGINEERED SAFETY FEATURE AClUA_ TION SYSTEM INSTRUMENTATION TRIP VALUES 1

2 l

W ALLOWABLE f

FUNCTIONAL UNIT

. TRIP VALUE VALUES l

E 4.

MAlli STEAM AND FEEDWATER ISOLATION (MSIS)

G a.

Manual.(Trip Buttons)

Not Applicable Not Applicable m

i b.

Steam Generator Pressure - Low

> 751' psia (2)

.> 729.613 psia (2) 5.

CONTAINMENT COOLING (CCAS) a.

Manual (Trip Buttons)

Not Applicable Not Applicable

}

b.

Contalsment Pressure - liigh

< 18.4 psia-

< 19.024 psia i

I c.

Pressurizer Pressure - Low

> 1766 psia (1)

> 1712.757 psia (1) l l

R 6.

IlECIRCULATION (RAS) j

[

a.

Manual (Trip Buttons)

Not Appilcable Not Applicable 4

iG b.

Refueling Water Tank - Low 54,400 e 2,370 gallons between 51,050 and 58,600 l

(equivalent to 6.0 4 0.5%

j indicated. level)

~

gallons equivalent to between

.111% and 6.889g l-

.h" indicated level) 5 7.

LOSS OF POWER 3

I N

a.

4.16 kv Emergency Bus Undervoltage 5

(Loss of Voltage) 3120 volts (4) 3120 volts'(4)

E b.

460 volt Emergency Bus Undervoltage 423 + 2.0 volts 423 + 4.0 volts i

(Degraded Voltage) with"an 8.0 t 0.5 with an 8.0 + 0.8

~

m i

second time delay second time delay f

e i

t l

l T..A.B.L. E. 3. '. 3...4.( Con t i n ued ).

M g

ENGINEERED SAFETY FEATUlif ACIUAIION SYSTIM INSTRUMENTATION TRIP VALUES' e

e,,

ALLOWABLE FUNCTIONAL UNIT _

IRIP VALUE VALUES b

8.

EMERGENCY fEEIMATER (EFAS).

2

.-4

~

a.

Manual (trip. Button's)

Not Applicable Not Applicable I

b.

Steam Generator (A&B) Level-Low 46.7% (3)

> 45.811% (3)'

c.

Steam Generator AP-liigli (SG-A > SG-H)

~ 90 psi

<. 99.344 psi 1

d.

Steam Generator AP-liigh (SG-B > SG-A)

- 90 psi

< 99.344 psi e.

Steam Generator (A&B) Pressure - Low

- 751 psia (2)

> 729.613 psia (2) 1 1

5 t'

(1) Value niay be decreased manually, to 'a miniemai of,100. psla', during a planned reduction in' pressurizer

{

pressure, provided the snarglei betweesi tiie, pressierizer pressure and this value is maintained at < 200 psi; I

Y the setpoint shall' be increased automatically as' pressurizer pressure is increased until the trl,p set-point is reached.

Trip may be usanually bypassed below 40Q psla; bypass shall be autoniatically removed l

whenever pressurizer pressure is.1 00 psia.

5 I'

(2) Value may be decreased nianually during a plarisied reduction in steams generator pressure, provided the j

margin between tiie steasii gesierator pressure and this value is maintained at' < 200 pst; tiie setpoint shall be increased automatically as steam generator pressure is ' increased until the trip setpoint is reached.

g so R

(3)

of the distance between steam generator upper aiid lower level instrumerit nozzles.

1 5

i 3

(4)

Inverse time Yelay set value, not a trip value.

The zero voltage tri e will occur in 0.75.+ 0.075 t

2 seconds.

i o,

W i

,a i

l 3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS I

l LIMITING-CONDITION FOR OPERATION 3.4.1 ' Both reactor coolant loops and both reactor coolant pumps in each loop shall be 'in operation.

i APPLICABILITY: As noted below, but excluding MODE 6*.

l lACTIONi MODES 1 and 2:

i

FOUR PUMP OPEPATION ***

i.

i

.With less than four reactor coolant pumps in operation be in at least l HOT STANO3Y within one hour.

'l L

I

.PART LOOP OPERATION ***

Illl a.

With one reactor coolant pump not in operation, STARTUP and/or ll continued. POWER OPERATION may proceed provided THERMAL POWER is ll restricted - to < **'; of RATED THERMAu POWER and the setacint for tha Linear Power LLei - High trip has been reduced :o' tne value j

scecified.in Specification 2.2.1 for operation wi-h three reactor j

.{

coolant pumps operating.

b.

With two reactor coolant pumps in opposite loops not in operation, o

l' STARTUP and/or continued POWER OPERATION may proceed provided THERMAL i-POWER is - restricted to < ***. of RATED THERMAL POWER and the setpoint

!j for the Linear Power Level - High trip-has been reduced to the value l

specified in Specification 2.2.1 for operation with two reactor coolant pumps. operating in opposite loops.

l l

f c.

With two. reactor coolant pumps in the same loop not in operation,

'~

STARTUP and/or continued POWER OPERATION may proceed provided the

!i.

water level in both steam generators is maintained above the Steam Generator. Water Level - Low trip setpoint, the THERMAL POWER is l

restricted to < ***. of RATED THERMAL POWER, and the setpoint for l!

the Linear Power Level - High trip has been reduced to the value l

l l!

specified in Specification 2.2.1 for operation with two reactor

-l ccolant pumps operating in the same loop.

L I

lI' l

I.

l

See Special Test Exception 3.10.3.

l! **These values left blank pending NRC approval of safety analyses for

-l '

operation with less than four reactor coolant pumps operating.

! ***Part loop operation is not allowed in Modes 1 and 2 pending APL submittal i

and NRC approval of safety analyses.

I

.1 ARKANSAS - UNIT 2 3/4 4-1 Amendment No. f 4

., _. _. ~.. _... _

i' l

' l-1

.l' REACTOR COOLANT SYS. TEM

~I:

l ACTION:

. (Continued )

l!

!N0DE 3:

l!t Operation may proceed provided two reactor coolant loops 'are in -operation with

!,at least one reactor coolant pump in each loop. With less than one reactor

ccolant pumo in each loop in. operation have at least one pump in e'ach loop l j:l2 hours.in operatien within one hour or be in at least HOT SHUTDOWN within Il; NODES 4 and 5:

-ll Operation miy; proceed provided at least one reactor coolant loop is in opera-l; tion with an associated reactor coolant pump or shutdown cooling pump *, The l provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

Y p-

.I l[*All reactor coolrat pumos and shutdown cooling pumps may be de-ener*1:ed for

up to I hour, trovided no operations are permitted which could cause dilution

!: of tne reactoc coolant system boron concentration.

' SURVEILLANCE REQUIREMENTS

;4.4.1 The Reactor protective Instrumentation channels specified in tne llapplicable ACTION statement above shall be verified to have had their trio j,!setpoints changed to the values specified in Specification 2.2.1 for the

!.japplicable number of reactor coolant pumps operating either:

1 f!

a.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after switching to a different pump combination l ll' if switch is made wnile operating, or ll b.

prior to reactor criticality if switch is made while shutdown.

/>

l; Il ARKANSAS - UNIT 2 3/4 4-2 Amendment No. g 4 tt.

~-,

..,,.r...,

- + -. -

...-y.

, - - ~

i

..~

(

. e.

s 1

i 9

1 ii a

i parf6.T :h:r:-,:q.p;iea@i:a::.ma,.rsa:irng:fj: 4 Ar: j 7

e ' -t - i ! -.::.4 l'

l ' {~~~l ll l

"y:y:=.d:ad:.Eu:@s'iw ;:ns....jr r :.::::::.:nn:s..apud ce;i ~ * - e..x.

s...a:a s e j e

-t 1

r

(

e 4

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t CONTAINMENT SYSTEMS CONTAINMENT-STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.5.The structural integrity of_ the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.5.

APPLICABILITYi, MODES 1, 2, 3 and 4 ACTION:

With-the ' structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS

4. 6.1. 5.1 Containment Tendons The containment tendons' structural

' ntegrity snali ce aemonstrated at the'end of cne, three and five years i

following tne initial containment structural integrity test and at five year intervals thereafter. The tendons' structural integrity shall be demonstrated by a visual examination (to the extent practical and witt-out dismantling load bearing components of the anchorage) of a repre-

_sentativeJsample* of at least 21 tendons (e dome, 5 vertical, and 10 hoop) and ~ verifying no abnormal ~ degradation. Unless there is evidence of abnormal degradation of the containment tendons ~during the first three tests of the tendons,.the number of tendons examined during sub-sequent tests may be reduced to a representative sample of at least 9 tendons (3 dome, 3 vertical and 3 hoop).

"For each inspection, the tendons shall be selected on a random but representative basis so that the sample group will change somewhat for each inspection; however, to develop a history of tendon performance and to correlate the observed data, one tendon from each group (dome, vertteel, and hoop) may be kept unchanged after the initial selection.

ARKANSAS-UNIT 2 3/c 6-8

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CCNTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.1.2 Each isolation valve specified in Table 3.6-1 shall be-demonstrated OPERABLE during.the COLD SHUTDOWN or REFUELING MODE at least once ner 18 months by verifying that on a containment isolation test. signal, each isolation valve actuates to its isolation position.

4.6.3.1.3 The isolation time of each power ~ operated or automatic valve of Table 3 6-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.6.

t l

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ARKANSAS-UNIT 2 3/4 6-17

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.I.AUl!. 3_. 6-1 CONTAltMENI ISOLATION _ VALVES PENETRATION ISOLATION NUMBER VALVE NUMBER

[UNCIION T_IME (SEC) n A.

CONTAllMENT 1SOLATIGil 2P7 2CV-5852-2#

"A" S/G Sample Isolation (outside) 1 20 2CV-5859-2#.

"B" S/G Sample Isolation (outside).

1 20 2P8 2SV-5833-1 RCS & Pressurizer Sample Isolation (inside) -

102 2SV-5843-2 RCS & Pressurizer Sample Isolation (outside) 10 2

2P9 2CV-6207-2 ll.P. flitrogen to Si Tanks (outside) 102 i

2P14 2CV-4821-1 CVCS 1./D Isolation (inside) 1 35 2CV-4823-2 CVCS L/D isolation (outside)

< 20 2P18 20Y-4846-1 RCPSealReturnIsolation(inside) 25 s

i R

2CV-4847-2 RCP Scal' Return isolation _(outside)

I 20 l

1 2P31 2CV-2401-1 Contairmient Vent lleader (inside) 7 20 i

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2CV-2400-2 Contalamient Vent Ileader (outside) 520 m

2P37 2SV-5878-1 QuenchTankLiquidSample(inside) 1 20 i

2SV-5871-2 QuenchTankLiquidSample.(outside) 1 20 j

2SV-5876 SI Taaks Sample Isolation (outside) 1 20 1

2P39 2CV-4690-2 Quench Tank Makeup & Dem89 Water Supply i

li Isolation (9utside) 1 20 2P40 2CV-3200-2 fire Water Isolation (outside) 1 20 j

2P41 2CV-6213-2 L..P. Nitrogen Supply Isolation (outside)

' 1 20 3

2 PSI 2CV-3852-1 Chilled Water Supply isolation (outside)-

1 20 j

2PS2 2CV-5236-1.

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2CV-3851-1 Chilled Water Return Isolation outside) 1 20 g

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2P60 2CV-5254-2 CCW from RCP Coolers Isolation inside)

< 20 2CV-5255-1 CCW froa RCP Coolers Isolation inside) 520

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TABLE 3.7-2_

h HAXIMUM ALLOWABLE LINEAR POWER LEVEL-lilGil TRIP SEIPOINI WITil kN0PERABLE

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4 ARKANSAS'- tJNIT 2 3/47-4

h REFUELING OPERATIONS REFUELING MACHINE OPERADILITY LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine sha'.1 be used for movement of fuel l

assemblies and shall.be OPERABLE with:

a.

A mininium capacity of 3750 raunds, b.

An ~ overload cut off limit of < 100 pounds plus the combined weight of one fuel assembly, 3he part length CEA, and the grapple in the " fuel only" region, and c.

An overload cut off limit of < 100 pounds plus the combined weight of one fuel assembly, 5he part length CEA, the grapple, and the hoist box in the " fuel plus hoist box" region.

APPLICABILITY : During movement of CEAs or fuel assemblies within the reactor pressure vessel.

ACTION:

With tne recuirements for refueling machine OPERABILITY no: satisfisc, suspend its use from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS l:

4.9.6 The refueling machine shall be demonstrated OPERABLE within 72 l

hours prior to the start of movement of fuel assenblies within l

the reactor pressure vessel by performing a load test of at least 3750 pounds and demonstrating automatic load cut offs when the crane loads exceed 100 pcunds plus the applicable loads.

ARKANSAS - UNIT 2 3/4 9-7 Anendment No. 2 4

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REFUELING OPERATIONS

~

CRANE TRAVEL - SPENT FUEL POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Leads in excess of 2000 pounds shall be prohibited from travel over fuel assemblies in the spent fuel pool.

APPLICABILITY: With fuel assemblies in the spent fuel pool.

ACTION:

With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS J.9.7 The crane electrical power disconnect wnich prevents crane travel over the spent fuel cool shall be verified open under acministrative control at least once per 7-days, or the crane travel interlock which prevents

.c crane travel over the scent fuel pool shall be demonstrated OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to each use of the crane for lifting loads in excess of 2000 p. ands.

ARKANSAS - UNIT 2 3/4 9-8

--,,_.c.,.

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3/4.10 SPECIAL TEST EXCEPTIONS

~

SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement'of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).

APPLICABILITY: MODE 2.

ACTION:

a.

With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at > 40 gpm of 1731 ppm boric acid solution or its equivalent untiT the SHUTCOWN MARGIN required by Specification 3.1.1.1 is restored.

b.

With all full length CEAs inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue baration at > 40 gem of 1731 ppm boric acid solution or its equivalent uniil -the SHUTDOWN MARGIN re-quired by Soecification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

~

4.10.1.2 Each CEA not fully inserted shall be dimenstrated capable of full insertion when tripped from at least the 50" withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SRUTDOWN MARGIN to less than the j

limits of Specification 3.1.1.1.

i i

ARKANSAS - UNIT 2 3/4 10-1 1

i SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.1 The group height, insertion and power distribution limits of Specifications 3.1.1. 4, 3.1. 3.1, 3.1. 3. 5, 3.1. 3. 6, 3. 2.2, 3. 2. 3, 3.2.7 and the Minimum Channels OPERABLE requir2 ment of Ranctional Unit 15 of Table 3.3-1 may be suspe..ded during_ the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is restricted to the test power pIEtau which shall not exceed 85% of RATED THERMAL POWER, ana b.

.The limits of Specification 3.2.1 are maintained and determined as specified in Specif1 cation 4.10.2.2 below.

APPLICABILITY: During startup and PHYSICS TESTS.

ACTION:

Witn ~any of ne limits of Specification 3.2.1 being exceedec wnile any of the aoove requirements suspended, either:

a.

Peduce THERMAL POWER sufficiently to satisfy the reouirements of Specification 3.2.1, or b.

Se in HOT STANDBY wthin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.10.2.1.The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which any of the above requirements are suspended and shall be verified to be within the test power plateau.

4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously _with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which any of the above requirements are suspended.

ARKANSAS - UNIT 2 3/4 10-2 Amendment No. g4

'3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTOOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are coctrollable within acceptable limits, and 3) the reactor will be maintaine sufficiently subcritical to preclude inadvertent criticality in the sNtdown condition.

SHUTDOWN MARGIN requirements vary throughout cors life as a function of fuel' depletion, RCS boron concentration, and RCS T The most restrictive condition occurs at EOL, with 7,, at no 18Ed operating

- temoerature, and is associated with a postuilted steam line break accident and resulting uncontrolled RCS cooldown.

I,n tne analysis of this accident, a minimum SHUTOOWN MARGIN of 5.0t.tk/k is required to control the reactivity transient. Accordingly, the' SHUTDOWN MARGIN requirement is based uoer this limiting condition and is consistent with FSAR safety analysts c.sumptions. With T

< 200*F, the reactivity transients result-ingfromanypostulatedaccidE!areminimalanda5%:.k/kshutdown j

margin provides adequate protection.

I 3/4.1.1. 3 BORON DILUTION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant

-System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9,M5 cubic feet in approximately 25 minutes. The reactivity change rate associated with boron concen-tration reductions will therefore be within the capability of operator

- recognition and control.

3 / 4.1.1. 4 MODERATOR TEMPERATURE COE, JIENT (MTC) l The l'initations on MTC are provideu to ensure that the assumotions used in t' e accident and transient analysis remain valid through eacn j

fuel c/cie. The surveillance requirements for measurement of the MTC

-during each fue' cycle are adequate to confinn the MTC value since thi coefficient changes slowly due principall,7 to the reduction in RCS boren concentration associated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurances that the coefficient will be maintained within acceptable values throughout each fuel cycle.

ARKANSAS - UNIT 2 B 3/4-1-1 Amendment No. o 4

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

- -.. ~., -, -

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the. reactor will not be made.

critical with the Reactor Coolant System average temperature less than 525'F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed tenperature range, 2) the protective instrumentation is within its normal coerating range 3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 4) the reactor pressure vessel is above its minimum RT temperature.

NDT 3/4.1.2 'BORATION SYSTEMS The baron injection system ensures that negative reactivity control is available during each mode of facility operation..The components required to perform tais function include 1) borated water sources.- 2) charging pumps, 3) separate flow paths, 4) boric acid makeup pumps, 5) associated heat tracing-systems, and 6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temoerature above 200'F, a minimum of two secarate and redundant baron injection systems are provided to ensure

~

single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be comaleted without undue risk to overall facility safety fran injection systen failures during the repair period.

.The boration capability of either system is suffici-. ta provide a SHUTDOWN MARGIN from expected operating conditions of

.0% ak/6 after l

xenon decay.and cooldown to 200'F. The maximum expected borasion cap-ability requirement occurs at EOL from full power equilibrium xenon conditions and requires boric acid solution from the boric acid makeup tanks in the allowable concentrations and volumes of Specification 3.1.2.3 or 56,455 gallons of 1731 ppm borated water from the refueling water tank.

l With the RCS temperature below 20G'F one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restric-tions prohibiting CORE. ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

ARKANSAS - UNIT 2 B 3/4 1-2 Amendment No.

e 5

ee w

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1 REACTIVITY CONTROL SYSTEMS BASES

.The boron capability-required below 200*F is based upon providing a 55 ak/k SHUTDOWN MARGIN after xenon decay and cooldown from 200*F -to 140'F.

This condition requires either 8185 gallons of 1731 ppm barated water-fromlthe refueling water tank or boric acid solution from the boric acid makeup tanks in accordance with the requirements of Specification 3.1.2.7.

The contained water volume limits includes allowance for water notlavailable because of discharge line location and other physical characteristics.. The 35,250 gallon limit for the refueling water tank I

is based upon having an indicated -level in the tank of at least 25.

The OPERABILITY of one boron injection system during REFUELING en'sures that this system is available for reactivity control while in MODE 6.

The limits on contained water volume and boron concentration of the RWT also ensure a pH value of. between 3.9 and 11.0 for :he solution recirculated within containment after a LOCA. This pH band minimi:es the evolution of iodine and minimizes the effect of chloride and clustic stress corrosion on mechanical systems and components.

3/A.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable oower l

distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is l

maintained, and (3) the potential effects of CIA misalignments are limited to accepcable levels.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met.

l The ACTION ' statements applicable to a stuck or untrippable CEA, o two or more inoperable CEAs, and to a large misalignment (> 19 inches) of two or more CEAs, require a prompt shutdcwn of the reactor since any of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrip-l.

able CEA, the loss of SHUTDOWN MARGIN.

l For small misalignments (< 19 inches) of the CEAs, there is 1} a small effect on the time dependent long term power distributions rela-l l

tive to those used in generating LCOs and LSSS setpoints, 2) a small effect on the available SHUTDOWN MARGIN, and 3) a sna11 effect on the l:

l ejected CEA worth used in the safety analysis. Therefore, the ACTION ARKANSAS - UNIT 2 B 3/4 1-3 Amendment No. 2 4 4

l l

-,a e ev <-

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.,we,-

-w,-w,s, eve,.

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g

-y

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g9 -v g v -

or e -w y

y-+m~,w-s-w - +w wvmp

d REACTIVIT!Y CONTROL SYSTEMS BASES statement associated with small misalignments of CEAs permits a one hour time interval-during which attempts may be made to restore the CEA to within its alignment requirements. The one hour time limit is sufficient to' (1) identify causes of a misaligned CEA, (2) take aopropriate correc-tive. action to realign the CEAs and (3) minimize the effects of xenon redistribution.

The CPCs provide protection to the core in the event of a large misalignment (> 19 inches) of a CEA by applying appropriate. penalty factors to the calculation to account for the misaligned CEA. However, this misalignment'would cause distortion of'the core paw $r distribution.

This distribution may, in turn, have

significant effect on 1) the available SHUTDOWN MARGIN, 2) the time dependent long term power distri-butions relative to those used in generating LCOs and LSSS setpoints.

- and 3) the ejected CEA worth used in the safety analys.is.

There fore,-

the ACTION statement associated with.the large misalignment of a CEA

~ requires a prompt realignment of the mistligned CEA.

The ACTION statements applicable to misaligned or inocerable CEAs include requirements to align the OPERABLE CIAs in a given group with tne inoperable CIA.-

Conformance with these alignment reouirements brings tne core, within a snort period of time, to a configuration consistent with that assumed.in generating LCO and LSSS setpoints. However, extended operation with CEAs significantly inserted in-the core may lead to perturbations in 1) local burnup, 2) peaking factors and 3) available SHUTDOWN. MARGIN which are more adverse than the conditions assumed to

. exist in tne safety analyses and LCO and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

Operability of at least two CEA position indicator channels is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits. The CEA " Full In" and " Full 4

Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions. Therefore..the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the

- positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.

ARKANSAS - UNIT 2 3 3/4 1-4

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1 q3/4.2 POWER DISTRIBUTION LIMITS i

BASES 3/4.2.1 - LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200'F.

Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its its limits.

The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating limit corre-sponding to the allowable peak linear heat rate.

Reactor operation at or below this calculated power level assures that the limits of Figure 3.1-1 are not exceeded.

The COLS: calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to tne coerator.

A COLSS alarm is annunciated in the event that the core power exceeds the core power' operating limit.

This orovides adequate margin to the linear heat rate operating limit for normal steady state oceration.

Normal reactor power transients or equipment failures which do not require a reactor trio may result in this core power operating limit being exceeded.

In tne event this occurs, COL 95 alarms will be annunciated.

If the event which causes the L

COLSS limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective i

Instrumentation.

The COLSS calculation of the linear heat rate limit includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the maximum linear heat rate calculated by COLSS is greater than or equal to that existing in the core.

To ensure that the design margin to safety is maintained, the COLSS comouter engineering uncertaind[ measurement uncertainty factor of 1.053, an l

program includes an F facter of 1.03, a THERMAL POWER measurement t

t uncertainty factor of 1.02 and aporopriate uncertainty and penalty l

factors for flux peaking augmentation and rod bow.

Parameters requireo to maintain the operating limit power level based on linear heat rate, margin to DNB and total core power are also monitored by the CPCs.

Therefore, in the event that the COLSS is not l

being used, oneration within the limits of Figure 3.2-2 can be maintained by utilizing a predetermined local power density margin and a total core power limit in the CPC trip channels.

The above listed uncertainty and penalty factors are also included in the CPCs.

ARKANSAS - UNIT 2 3 3/4 2-1 Anendment No. 2 4 I-

p 4

5 l,

POWER DISTRIBUTION LIMITS

]' BASES' 3/4.2.2 RADIAL' PEAKING FACTORS C

Limiting the values of the -PLANAR RADIAL PEAKING FACTORS (Fxy) used in llthe COLSS and CPCs to values equal to or greater than the measured, PLANAR ~

RADIAL PEAKING' FACTORS (F*y) provides assurance that the limits calculated

' by COLSS'and the CPCs remain valid.

Data from the incore detectors are used

'for determining the measured. PLANAR RADIAL PEAKING FACTOR $.

The periodic surveillance requirements for determining the measured PLANAR RADIAL PEAKING :

FACTORS provides assurance that the-PLANAR RADIAL PEAKING FACTORS used in llCOLSS and the CPCs remain valid throughout the fuel cycle. Detarmining the llrressured PLANAR RADIAL PEAKING FACTORS after each fuel loading prior to i 2xceeding 70". of RATED THERMAL POWER provides additional assurance that the

- core was properly loaded.

t j 3/4.2.3 AZIMUTHAL POWER TILT - T,

!L Th'e limitations on the AZIMUTHAL POWER TILT are provided to ensure that

design safety margins are maintained. An AZIMUTHAL POWER TILT greater than

' I j 0.10 is not. expected and if it should occur operation is restricted to on'y

' those conditions recuireo to identify the cause of the tilt.

The tilt is

romally calculated by.COLSS.

The surveillance requirements specified wnen

. : COLS5 is out of service provide an -acceptacle means. of detecting tne presence

..of a steady state tilt.

It is necessary to explicitly account for cower

asymmetries because One radial peaking factors used in the core power ll distribution calculations are based on an untilted power distribution.

ll ll AZIMUTHAL POWER TILT is measured by assuming that the ratio of the power

!!at any core location in the presence of a tilt to the untilted power at the

~ ll location is of the form:

tilt 'untilt P

I 1*

g cos (e - 0,)

q ll where:

I is the-peak fractional tilt amplitude at the core periphery bl q

l1 9 is the radial normalizing factor f

l 0 is the azimuthal core location e, is the azimuthal core location of maximum tilt

'le l' ARKANSAS-UNIT 2 8 3/4 2-2 Amendment No. p

'[

e

~=,e

,e

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,w r,

e e,

y,-

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POWER DISTRIBUTION LIMITS BASES P

/P is the ratio of the power at a core location fn the tilt untilt

' presence of a tilt to the power at that location with no tilt.

i 3/4.2.4 DNBR MARGIN The limitation on DNBR as a function of AXIAL SHAPE INDEX repre-sents a conservative envelope of operating conditions consistent with the safety analysis assumptions and which nave been analytically demon-

-stratec' adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences, of which the loss of flow transient is the most limiting. Operation of the core with a ONBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.

Either of the two core power. distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the ONBR channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the DNSR does not violate its limits. The COLSS performs tnis function by con-tinuously monitoring L1e core cower distribution and calculating a core operating limit corresponding to the allowable minimum CNBR. Reactor operation at or below this calculated power level assures that the limits of Figure 3.2-3 are not violated. The COLSS calculation of core I

power operating limit based on DNBR includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the core power at which a DNBR of less than 1.24 could occur, as calculated I

by COLSS, is less than or equal to that which would actually be required in the core. To ensure that the design margin'to safety is maintained, the COLSS computer program includes an F measurement uncertainty factor I

y of 1.053, an engineering uncertainty facfor of 1.03, a THERMAL POWER i

measurement uncertainty factor of 1.02 and appropriate uncertainty and penalty factors for flux peaking augmentation and rod bow.

Parameters required to maintain the margin to ONB and total core power are clso monitored by the CPCs. Therefore, in the event that the COLSS'is not being used, operation within the limits of Figure 3.2-4 can l

be maintained by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPC.

ARKANSAS - UNIT 2 3 3/4 2-3 Anendment No. a.1 4

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a,w-e

--,-o,

POWER DISTRIBUTION LIMITS BASES

,3/4.2.5 RCS FLOW RATE iil This specification is provided to ensure that the actual RCS total flow

{ J rate is maintained at or above the minimum value usad in the LOCA ~ safety j

analyses.

3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE

' This specification is provided to ensure that the actual value of reactor

.! coolant cold leg tanperature is maintained within the range of values used in

!!thesafetyanalyses.

I' l

3/a.2.7 AXIAL SHAPE INDEX i;

I This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX s maintained within the range of values used in the safety analyses, jj3/4.2.8 PRESSURIZER PRESSURE ll This spe:1fication is provided to ensure that the actual value of pressuiizer

!' cressure is maintained witnin the range of values used in the safety analyses.

ii::

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I! ARKANSAS - UNIT 2 8 3/4 2-4 Anendment No. 21

O

b 3/4.4 REACTOR COOLANT SYSTEM BASES 1

3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.24 during all normal operations and anticipated transients.

STARTUP l

and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for the Power Level-High, Reactor Coolant Flow-Low, and Thermal Margin / Low Pressure.

trips have been reduced to their specified values. Reducin these. trio setooints ensures that the ONBR will De maintained above 1. 4 during l

three pump operation and that during two pump operation the core void fraction will be' limited to ensure carallel channel flow stability within the core and thereby prevent premature DNE.

A single reactor coolant loop with its steam generator filled above the low' level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure consi-derations' require plant cooldown if component repairs and/or corrective actions cannot be made within.the allowable out-of-service time.

t l

l 3/4.4.2 and 3/?.4.3 SAFETY VALVES

}

I

(

- The pressurizer code safety valves operate to prevent the RCS from l

being pressurized aoove its Safety Limit of 2750 psia.

Each safety valve is designed to relieve 420,000.lbs per hour of saturated steam at the valve l setpoint. The relief capacity of a single safety valve,is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating shutcown l

cooling loop, connected to the RCS, provides overpressure relief capa-l

' bility and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE l

to prevent the RCS from being pressurized above its safety limit of 2750 psia. The combined relief capacity of these valves is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of j.

2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the i

l1 loss of turbine) and also assuming no operation of the steam dump valves.

I l

i ARKANSAS - UNIT 2 3 3/4 4-1 Amendment No. 2 4 r

. - -. -. -.. ~.

r o

REACTOR COOLANT SYSTEM

B_AS ES i

Demonstratior. of the safety valves

lift settings will occur only l ;during snutoown and will be performed in accordance with the provisions

.jof Section XI of the ASME Boiler and Pressure Vessel Code.

1

! 3/4. 4'. 4 PRESSURIZER 11 A steam bubble in the pressurizer ensures that the RCS is not a lhydraulically solid system and is capable of accommodating pressure

,lsurges during operation. The steam bubble also protects the pressurizer

-l} code _ safety valves against water relief. The steam bubble functions to

(; relieve RCS pressure during all design transients.

The requirement that 150 KW of pressurizer heaters and their

! associated controls be capable of being supplied electrical power from

{lduring a loss-of-offsite power condition to maintain nattual circulation

an emergency bus provides assurance that these heaters can be energi: ed

. :at HOT STANDBY.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator uoes ensure that the structural integrity of this portien of the RCS W il be maintained. The program for inservice insoection of steam generator tubes is based on a modification of Regulatory Guice 1.83, Revision 1.

Inservice insoection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degra-

dation due to design, manufacturing errors, or ietservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also

,!! provides a msans of characteri:ing the nature and caute of any tube

> ; degradation so that corrective measures can be taken.

' !{

The plant is expected to be operated in a manner such that the

! secondary coolant will be maintained within those chemistry limits found

j to result in negligible corrosion of tne steam generator tubes.

If the l secondary coolant chemistry is not maintained within these limits, l! localized corrosion may 'ikely result in stress corrosion cracking.

The

! extent of cracking durir.; olant operation would be limited by the

!; limitation of steam generator tube leakage between the primary coolant ll system and the secondary coolant system (primary-to-secondary leakage =

0.5 GPM per steam generator). Cracks having a primary-to-secondary

!! leakage less than this limit during operation will have an adequate

margin of-safety to withstand the loads imposed during nonnal operation

!.and by postulated accidents. Operating plants have demonstrated that

! crimary-to-secondary leakage of 0.5 GPM per steam generator can readily

!:be detected by radiation monitors of steam generator blowdown.

Leakage

! in excess of this limit will require plant shutdown and an unscheduled

inspection, during which the leaking tubes will be located and plugged.

ARKANSAS - UNIT 2 3 3/4 4-2 Amendment No. 20

3/4.9 REFUELING OPERATIONS BASES 3 /4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ens'ure that:

1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limita-tions are consistent with the initial conditions assumed for the baron dilution incident in the accident analyses.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range n tutron flux monitors ensures that redundant monitoring capability is av tilable to detect changes in the reactivity condition of the core, i

3/4.9.3 DECAY TIME

[

The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that l

sufficient time has elaosed to allow the radioactive decay of the short l

livec fission products. This decay time is consistent with the assumotions used in the accident analyses.

3/4.9.4 CONTAINMENT PENETRATIONS l

The requirements on containment penetration closure and OPERASILITY of the containment purge and exhaust system HEPA filters and charcoal adsarbers ensure that a release of radioactive material within contain-l ment will be restricted from leakage to the environment or filtered through the HEPA filters and charcoal adsorbers prior to discharge to the atmo-sphere. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. Operation of the containment purge and exhaust system HEPA filters and charcoal adsorbers and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

ARKANSAS - UNIT 2 B 3/4 9-1

t REFUELING OPERATIONS I

BASES 3/4.9.5 COMMUNICATIONS The requirement for connunications capability ensures that refueling station personnel can be promptly. informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

3/4.9.6 REFUELING MACHINE OPERABILITY The OPERABILITY requirements for the refueling machine ensure that: 1) the refueling machine will be used for movement of CEAs with fuel l

assemblies and that it has sufficient load capacity to lift a fuel assembly, and 2) the core internals and pressure vessel are pro-tected from excessive lifting for:e in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads-in excess of the nominal weight of a fuel assembly, CIA and associated handling tool over other fuel assemblies in the storage pool ensures that in the event :nis load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

2/4.9.8 COOLANT CIRCULATION The requirenent that at least one shutdown cooling loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

ARKANSAS - UNIT 2 3 3/4 9-2 Anendment No. 2 4 I

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a-s DESIGN FEATURES VOLUME 5.4.2. The total water and steam volume of the reactor coolant system is 10,295 + 400 cubic fewt at nominal T,yg of 545'F.

5.5 METEOROLOGICAL TOWER LOCATION 3.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1.1 The spent fuel. storage racks are designed and shall be maintained l

with a r.cminal 12.8 inch center-to-center distance between fuel assemblies having a maximum enrichment of 4.3 weight percent U-235 placed in the storage racks to ensure a k equivalent to < 0.95 when flooded with unborated water. The k.

N,'s0.95includesaconservativeallowance of 1.7', ik/k for uncertif, ties as described in Section 9.1.2.3 of the n

-:9AR.. In addition, fuel in the storage pool shall have a U-235 loading o# 1 27.8 grams of U-235 per axial centimeter of fuel assemoly.

CRITICALITY - NEW FUEL 5.6.1.2 The new fuel storage racks are designed and shall be maintained i

with a nominal 25.0 inch center-to-center distance between new fuel assemblies such that-X enrichmentof3.7weigIl(fwillnotexceed0.98whenfuelhavingamaximum percent U-235 is in place and aqueous foam moderation is assumed and K will not exceed 0.95 when the storage area is flooded with unborated wIff The calculated K,ff includes a conserva-er.

tive allowance of 1.0% ak/k for uncertainties.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 399' 10 1/2".

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 486 fuei assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

ARKANSAS-UNIT 2 5-5 Amendment No. o4

g.

9 IABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS di h

CYCLIC OR DESIGN CYCLE q

COMPONENT TRANSIENI LIMil 0R TRANSIENT

^

N Reactor Coolant. Systent 500 system.heatup aied couldown Heatup cycle - T from < 200*F cycles at rates L 100*F/hr.

to t 545*F; coo 1880ncycle-T,,9 from t 545*F to 1 200*F.

500 pressurizer heatup and lleatup cycle - Pressurizer temperature-cooldown cycles at rates from 1 200*f to 1 653*F; cooldown i 200*F/hr.

cycle - Pressurizer temperature from t 653*F to 1 00*F.

2 10 hydrostatic testing cycles.

RCS pressurized to 3110 psig with RCS temperature 1 60*F above the most.llaiting components' MOTT value.

200 leak testing cycles.

RCS pressured to 2250 psia witle RCS temperature greater than minimum for hydrostatic testing, but less than minimum RCS temperature for criticality.

400 reactor trip cycles.

Trip from 100% of RATED TilERMAL h

j POWER.

i f

,I 40 turbine trip cycles with-Turbine trip (total load rejectlon) delayed reactor trip.

from 100% of RATED TilERMAL POWER followed by resulting reactor trip.

200 seismic stress cycles.

Subjection to a sei nic event equal to one half th4 desthu basis j

earthquake (DBE).

I I

l s

t i

u

. ADMINISTRATIVE CONTROLS f 6.7 SAFETY LIMIT VIOLATION 6.7.1, The following actions shall be taken in the event a Safety Limit is violated:

l a.

The unit shall be placed in at least HOT STANDBY within one hour, b.

The Safety Limit violation shall be reported to the Commission, the Director, Nuclear Operations and to the SRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

A Safety Limit Violation Report shall be prepared.

The report snall be reviewed by the PSC.

This report shall describe (1)

l applicable circumstances preceding the violation (2) effects of the violation 'upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

!!l' d.

The Safety Limit Violation Report shall be submitted to the Comission, the SRC and the Director, Nuclear Operations within

{

14 days of the violation.

i ! 6.8 PROCEDURES

! ! 6.8.1 Written procedures chall be establisned, implemented and maintair.ed

cover',; the activities referenced below:

Y:

The applicable procedures recommended in Appendix "A" of a.

,{

of Regulatory Guide 1.33, Revision 2. February 1973.

b.

Refueling operations.

c.

Surveillance and test activities of safety related equignent.

ti d.

Security Plan impleentation.

I

{

e.

Emergency Plan impleentation.

.jj f.

Fire Protection Program implementation.

ll g.

Modification of Core Protection Calculator (CPC) Addressable Constants NOTE: Modification to the CP C addressable constants based on information obtained through the Plant Computer -

g CPC data link shall not be made without prior approval of the Plant Safety Committee.

i ll6.8.2

[

Each procedure of 6.8.1 above, and changes thereto, shall be reviewed l' by the PSC and approved by the General Manager prior to implementation and i f reviewed periodically as set forth in administrative procedures.

l II I

l!ARXANSAS-UNIT 2 6-13 Amendment No. 3, 77, 2 1 l

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ADMINISTRATIVE' CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made pro-vided:

a.

The intent of the original procedure is not altered.

b.

.The change is approved by two members of the plant management staff, at least one of wnom holds a Senior Reactor Operator's License on the unit affected, The change is documented, reviewed by the PSC and approved by c.

the General Manager within 14 days of implementation.

l 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless

,otherwise noted.

STARTUp rep 0RT 6.9.1.1 A summary recort of plant startup and power escalation testin snall be submitted following (1) receipt of an operating license, (2) g -

amendment to the license involving a planned increase in power level, I

(3) installation of fuel that has a different design or has been manu-

?

i factured by a different fuel supplier, and (4) modifications that may I

have significantly altered the nuclear, thermal, or hydraulic perfor-mance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and i

specificationsi Any corrective actions that were required to obtain satisfactory operation shall also be de ribed. Any additional specific details required in license conditions based on other commitments shal~.

be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or conmencement of commercial power operation, or (3) 9 montns following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least L

I every three months until all three events have been completed.

l

[

ARKANSAS - UNIT 2 6-14 Amendment No. 5 i

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ML20004G020

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