ML20008E686

Cycle 2 Reload Analysis Rept, Remaining sections.Marked-up Tech Specs Encl

ML20008E686
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/05/1981
From:	ARKANSAS POWER & LIGHT CO.
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ML20008E684	List: ML20008E683 ML20008E686
References
NUDOCS 8103090440
Download: ML20008E686 (175)

Text

Arkansas Nuclear One Unit 2 Cycle 2 Reload Analysis Report l

1 l

8103090440

TABLE OF CONTENTS Section Page

1. Introduction and Summary 1-1

2. Operating History of the Reference Cycle 2-1

3. General Description 3-1 4 Fuel System Design 4-1

5. Nuclear Design 5-1

6. Thermal-Hydraulic Design 6-1

7. Transient Analysis 7-1

8. ECCS Analysis 8-1

9. Reactor Protection and Monitoring System 9-1

10. 00E Test Assemblies 10-1 l

l 11. Technical Specifications 11-1

12. Startup Testing 12-1

13. References 13-1 i

t

List of Figures 3-1 Enrichment Zoning Pattern for D and 0* Fuel Assemblies 3-2 Cycle 2 Core Map 3-3 Assembly Average Burnup and Initial Enrichment Distribution 5-1 CEA Insertion Limits vs. Thennal Power 5-2 Assembly Relative Power Density, HFP at 80C, Equilibrium Xenon 5-3 Assembly Relative Power Density, HFP at 5 GWD/T, Equilibrium Xenon 5-4 Assembly Relative Power Density, HFP at EOC, Equilibrium Xenon 5-5 Assembly Relative Power Density wi m PL'EAS Inserted, HFP, B0C 5-6 Assembly Relative Power Density with Bank 6, HFP, 80C 5-7 Assembly Relative Power Density with Bank 6 and PLCEA's, HFP, BCC 5-0 Assembly Relative Power Density with PLCEAS, HFP, E0C 5-9 Assembly Relative Power Density with Bank 6, HPF, E0C 5-10 Assembly Relative Power Density with Bank 6 and PLCEAS, HFP, E0C 7.1.4-1 Loss of Load / Loss of Condenser Vacuun, Core Power vs. Time 7.1.4-2 Loss of Load / Loss of Condenser Vacuum, Core-Average Heat Flux vs. Time 7.1.4-3 Loss of Load / Loss of Condenser Vacuum, Reactor Coolant System Pressure vs. Time 7.1.4-4 Loss of Load / Loss of Condenser Vacuum, Reactor Coolant System Tencerature vs. Time 7.1.4-5 Lo 25 25 Power of Bank 6 Maximum Initial Linear Heat Rate K!J/ f t 14.5 14.5 Steady State Linear Heat Rate to KW/ft 21.0 21.0 Fuel Centerline fielt CEA Drop Time from Removal of Power sec 3.0 3.0 to Holding Coils to 90" Insertion

• Technical Specification 3.2.2 requires the measured planar radial peaking factor to be less than or equal to the planar radial peaking factor used by COLSS and CPC.

l The values used in any given analysis which takes credit for the margin preserved by LCO's were thora which produced a limiting case.

• • To bound future cycles, some of the transient analyses assured the ltTC to be within the range -3.5+ + .5 x 10-4do/ F. The MTC used for each analysis is given in the I

appropriate tables for the section where the analysis of that event is presented.

7-4

TABLE 7-3 ALLOWABLE INITIAL C0tlDITIONS FOR SAFETY AtlALYSIS Parameter Value Cold Leg Temperature 540 1 Tin 2 556.7 Pressurizer Pressure 2200 < P s 2300 6

Reactor Coolant flow Rate > 120.4x10 lbm/hr Axial Shape Index (ASI) .3 < ASI < +.3 l

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7-5

7.1.1 Baron Dilution Event '

The Boron Dilution e/ent was reanalyzed for Cycle 2 to demonstrate that sufficient time is available for an operator to identify the cause of &nd

, to terminate an approach to criticality for all subcritical modes of operation and to demonstrate that sufficient scram worth is available ,

in all operating modes. The results of the analyses establish corresponding shutdown margin requireinents for iiodes 3 tnrougn 5. t 1

During operation at power (i.e., Modes 1 and 2), an inadvertent boron dilution adds positive reactivity and can cause an approach to the DNBR *

and CTM limits. The core protection calculator (CPC) trip system monitors

the transient behavior of pertinent safety parameters and will generate [

a reactor trip if necessary to prevent the DNBR and CTM limits from being >

exceeded. The high pressurizer pressure trip will prevent reaching the I RCS pressure upset limit. The trip which is actuated depends on the rate '

of reactivity addit ion. For a boron dilution initiated from the hot zero power :

portion of Mode 2,tre power transient resulting from the reactivity -

insertion would be terminated by the high logarithmic power level trip l prior to approaching these limits. For the subcritical modes (i.e., Modes 3 {

thrnuoh 6), the time required to achieve criticality due to boron dilution l is dependent on the initial and critical boron concentrations, the boron  ;

reactivity worth, and the rate of dilution, i Table 7.1.1-1 compares the values of the key transient parameters assumed in each mode of operation for Cycle 2 and the reference cycle. The con-  !

servative input data chosen consists of high critical boron concentrations '

and low inverse boron worths. These choices produce the most adverse effects by reducing the calculated time to criticality in initially sub-critical modes. As in the FSAR, the time to criticality was determined by using the following expression-et crit B0 in Initial C

crit where at = Time interval to dilute to critical  ;

I BD

= Time constant C

crit = Critical boron concentration (ppn)

C Initial = Initial boron concentration (ppm) 7-6

Table 7.1.1-2 compares the results of the analysis for Cycle 2 with those for the FSAR. The key results are the minimun times required to lose the prescribed negative reactivity in each operational mode. As seen from Table 7.1.1-2, sufficient time exists for the operator to take appropriate action to mitigate the consequences of this event.

If the RCS is drained significantly in Cold Shutdown (Mode 5) to allow repairs (such as reactor coolant pemp seal replacement) the time to lose shutdown margin may be less. The analysis for this plant condition conservatively assumed water volume in the reactor vessel only up to the lower lip of the outlet nozzle. With this assumption, a time period of approximately 35.0 minutes is calculated to elapse before the available shutdown margin of 5.0 percent would be lost.

7-7

TABLE 7.1.1 '

KEY PARR4ETERS ASSU4E0 IN THE BORON DILUTI0fl AtlALYSIS Parameter FSAR Cycle 2 Critical Boron Concentration, PR4 (All Rods Out, Zero Xenon)

Power Operation (Mode 1) 1000 Startup (Mode 2)

}*

J

~*

1300 Hot Standy (Mode 3) 1300

+

987 Hot Shutdown (Mode 4) .

1300 Cold Shutdown (Mode 5) 998 1300 Refueling (Mode 6) 1068 1250 Inverse Baron Worth, PN / %v Power Operation 82 82 Startup 82 82 Hot Standy 82 62 Hot Shutdown 82 62 Cold Shutdown 82 62 Refueling 82 62 Shutdown Marain Assumed Sio Power Operation ~* 5.4 Startup 5.0

~

Hot Standy 5.0

( 2.0 Hot Shutdown J 5.0 Cold Shutdown 2.0 5.0 Refueling >5.0 >5.0

• The STS mode definitions were pronulgated subsequent to the FSAR. The i FSAR analysis treats these two modes as a sinqle node.

7-ia

TABLE 7.1.1-2 RESi:LTS OF THE BORON DILUTI0ft EVEilT Time to Lose r,escribed Shutdown Acceptance Criterion Mode Marqin (Minutes) (Minutes )

FSAR Cycle 2 Startup (Mode 2) 63 72* 15 Hot Standby (Mode 3) 50* 15 L 81 50* 15 Hot Shutdown (Mode 4}

Cold Shutdown (Mode 5) 66 48** 15 Refueling (Mode 6) 66 40 30

• To faci'itate comparisons with the FSAR results, the time to criticality for Moa- s 2 through 5 is evaluated as the time to lose a 2Nas initial shutdown reactivity. Continuing di'ution after loss of the 25 4 will result in a high logarithmic power level pre-trip alarm when the indicated neutron flux level reaches .01 percent power and a high logarithmic power level trip at 2 percent power. The reactor trip results in the insertion of any of the withdrawn CEAs. For Modes 2 through 5, the negative reactivity insertion by the tripped CEAs provides an additional period in excess of 75 minutes.

• • Tine shown assumes RCS is not drained (see text).

7-9

7.1.4 Loss of Load / Loss of Condenser Vacuum The Loss of Load (LOL), Loss of Condenser Vacuum (LOC V), and Turbine Trip (TT) events were presented in the FSAR as separate events (see Sections 15.1.7,15.1.28,15.1.29). As stated in the FSAR these events are analyzed to demonstrate that the RCS and main stean system pressures do not exeed 110", af design values (i.e., 2750 psia and 1210 psia, respectively). For Cycle 2 an analysis was made of a single event which bounds all three FSAR events. The purpose of the analysis was to denonstrate that the RCS and secondary pressure upset limits would not be violated for Cycle 2 operation.

The bounding event considered is a Loss of Load event initiated by a turbine trip without a simultaneous reactor trip, and assuming the Steam Dunp and Bypass system is inoperable. If the turbine trip were caused by a Loss of Condenser Vacuum, the main feedwater pump steam turbines would trip at the same time. Therefore, a LOL concurrent with loss of feed was analyzed to cover these events. The loss of load causes steam qenerator pressure to increase to the opening pressure of the main steam safety valves.

The reduced secondary heat sink leads to a heatup of the RCS and, in the presence of the assumed positive f1TC, an increase in core power. The transient is terminated by a reactor trip on high pressurizer pressure.

The LOL/LOCV event was initiated at the conditions shown in Table 7.1.4.1.

The combination of parameters shown in Table 7.1.4-1 maximizes the calculated peak RCS pressure. The key parameters for this event are the initial primary and secondary pressures and the moderator and fuel temperature coefficients of reactivity. The methods used to analyze this event are consistent with those described in the FSAR.

The initial core average axial power distribution for this analysis was assened to be a bottom peaked shape. This distribution is assumed because it ninimizes the negative reactivity inserted during the initial portion of the scram following a reactor trip and maximizes the time required to mitigate the pressure and heat flux increases. A Moderator Temperature Coefficient (r1TC) of +.5X10-4ap/ F was assumed in this analysis. This t1TC in conjunction with the increasing coolant temperatures, enhances the rate of change of heat flux and the pressure at the time of reactor trip. A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions was used in the analysis. This FTC causes the least anount of negative reactivity feedback to mitigate the transient increases in both the core heat flux and the pressure. The uncertainty on the FTC used in the analysis is shown in Table 7.1.4-1 The lower initial RCS pressure is used to maximize the rate of change of pressure and thus peak pressure following trip. The lower coolant inlet temperature and lower steam generator pressure combination made the secondary transient more severe by delaying opening of the main stean safety valves.

7-10

b The Loss of Load event, initiated from the conditions given in Table 7.1.4-1, results in a high pressurizer pressure trip condition at 5.6 seconds. At r l

9.0 seconds, the primary pressure reaches its maximum value of 2671 psia. I The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 8 3 seconds. The secondary pressure reaches its maximum value of 1144 psia at U J seconds after initiation l t

of the event. i Table 7.1.4-2 presents the sequence of events for this event. Figures 7.1.4-1 to 7.1.4-5 show the transient behavior of power, heat flux, the  ;

c RCS pressure, RCS coolant temperatures, and the steam generator pressure.

The results of this analysis demonstrate that the Loss of Load type event  !

will not result in peak RCS pressure or peak main stean pressure in excess '

of their respective upset pressure limits.

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TABLE 7.1.4-1 KEY PAR #1ETERS ASSlNED Ifl THE LOSS O F LOAD / LOSS 0 F CONDENSER VACUlti ANALYSIS Parameter Units, FSAR Cycle 2 Initial Core Power Level MWT 2900 2900 Initial Core Inlet Coolant F 556.7 540 Temperature Initial Reactor Coolant System psia 2200 2200 Pressure Initial Steam Generator psia 923 810 Pressure Moderator Temperature Coefficient X10-4ap/ F +5. +.5 8e1 Tenperature Coefficient Multiplier . 85 .85 CEA Worth at Trip iao -5.4 -5.4 Time to 90% Insertion of sec 3.0 3.0 Scram Rods Reactor Regulating System Operating flode flanual Manual (Control Roas)

Stean Dunp and Bypass System Operating Mode Inoperative Inoperative 1

7-12

TABLE 7.1.4-2 SEQlilCE Of EVErlTS FOR THE LOSS OF LOAD / LOSS O F C0flDErlSER VACUlJ1 EVEtlT Time (sec) Event Setpoint or Value 0.0 Loss of Secondary Load -----

5.6 High Pressurizer Pressure Trio 2422 ,sia Analysis Setpoint Reached 6.6 Pressurizer Safety Valyes Ooen 2500 psia 6.8 CEAs Begin to Drop Into Core -----

6. 9 Peak Power 111?' (of 2815 MWt) 8.3 Steam Generator Safety Valves 1093 psia Open 9.0 Maximum RCS Pressure 2671 psia 13.3 Maximum Steam Generator Pressure 1144 psia l

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Arkansas CORE POVER vs TIME 7.1.4-1 Nuclear One - Unit 2 7-14

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POW & LIG iT CO LOSS OF LOAD / LOSS OF CONDENSER VACUUM n,nansas CORE AVERAGE HEAT PLUX vs TIME 7.1. 4 -2 Nuclear One Unit 2 7-15

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Arkansas REACTOR COOLANT SYSTEM PRESSURE vs TIME 7.1.4-3 Nuclear One - Unit 2 7-16

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7-18 __ _ _

7.1.6 CEA Withdrawal Event The CEA Withdrawal Event was reanalyzed for both subcritical and critical initial conditions for Cycle 2 to demonstrate that the initial margin maintained by the LC0's in conjunction with the RPS trips prevent violation of the OrlBR, CTM. and RCS pressure design limits. In addition, the core power and heat flux responses during this event are input to the calculation of CPC dynamic constants. The dynamic calculation performed by algorithms within the CPC are required to provide conservative estimates of minimum transient OflBR and maximum transient Local Power Density (LPD). This assures that low DtlBR and/or high LPD trip will be initiated when needed to prevent violation of DNBR and CTri limits during a CEA Withdrawal event from critical conditions.

For the CEA withdrawal fron subcritical conditions, the high logarithmic power level reacLar trip is initiated when the power exceeds two percent of full power. The key r"ameters for the transient analysis are the reactivity addition rate cue to rod motion, moderator temperature feed-back effects, and initial axial power distribution. The input values selected maxinize the power increase and thus the nargin degradation.

The values of key parameters used in the analysis of this event are presented in Table 7.1.6-1.

The sequence of events for tr.e subcritical CEA Withdrawal case is presented ii' Table 7.1.6-2. Figures ?.1 A-1 to 7.1.6-4 present the corresponding transient behavior of core power, core average heat flux, RCS pressure, and the RCS coolant temperatures. This transient results in a minimum CE-1 DtlBR of 1,2G. Also, the analysis shows that the fuel centerline temperatures are well below those corresponding to the fuel centerline melt limit.

The key parameters usec for the analysis of the CEA Withdrawal from one percent power are given in Table 7.1.6-3. The sequence of events and NSSS response are shown in Table 7.1.6-4 and Figures 7.1.C-5 through 7.1.6-8. For this limiting case, the transient is ternMated by d high pressurizer pressure trip and results in a maximun RCS pressure of 2662 psia. Since the high pressurizer pressure trip occurs before the tine at which a high LPD or low DflBR trip would be required, DflBR and Cm limits are not exceeded.

Protection against exceeding the DNB and CTM limits for a CEA Withdrawal at full power is provided by the CPCs which provide an automatic reactor trip on low DNB or high LPD. When initiated from the extremes of the LCOs, the CPC low DtlBR trip will be initiated prior to 14.6 seconds for the most severe full power case . Key parameters for the full power case are detailed in Table 7.1.6-5. The sequence of events is given in Table 7.1.6-6 and NSSS parameter responses are shown in Figures 7.1.6-9 through 7.1.6-13.

As shown by the analyses, the CEA Withdrawal events from critical and sub-critical conditions do not lead to the violation of DNB, CTM, and RCS pressure design limits. For the limiting cases analyzed, the high pressurizer pressure trip, low DNBR CPC trip, high LPD CPC trip, and high logarithmic power trip function to assure the design limits are not exceeded. .

7 - l 's

TABLE 7.1.5-1  :

KEY PARAMETERS ASSUf1ED If1 THE AtlALYSIS OF UNCONTROLLED CEA WITHDRAWAL FROM A SUBCRITICAL C0fiDITI0ft i Parameter Units FSAR Cycle 2 Initial Core Power (MWt) 2900 x 10-10 2900 x 10-10 ,

Level Initial Inlet Coolant (F) 544.6 544.6 Teraperature Core fiass Flow (1061bm/hr) 55.54* 116.2 Rate Reactor Coolant System (psia) 2250 2200 Pressure Stean Generator Pressure (psia) 1000 990.

Total fluclear Heat 4.68 4.2 Flux Factor fioderator Temperature (10-4ac/F) +0.5 +0.5 Coefficient Fuel Tenperature 0.85 0.85 '

Coefficient fluitiplier CEA flaxinun Reactivity (10-4ao/sec) 3.0 2.5 Addition Rate CEA Worth on Trip (10-2ao) -?.4 -5.0 Stean Bypass Systen Manual Manual Feedwater Regulating Manual fianual System

• 0ne-loop operation was allowed by the Tech Spec during approach to criticality at the time FSAR analysis was performed.

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7- 20

TABLE 7.1.6-2 SEQUENCE OF EVENTS FOR THE UNCONTROLLED CEA WITHDRAWAL FROM A SUBCRITICAL CONDITION Set Point Time (sec) ,

Event Or value 0.0 Initiation of withdrawal ---

40.0 Reactor reaches criticality ---

58.8 High Logarithmic power level +2 percent of trip condition full power 59.2 Trip breakers open ---

59.5 Shutdown CEAs begin to drop ---

into core 59.6 Peak transient neutron flux 129 percent of 2815 t1Wt 61.4 Peak transient core average 75 percent of full heat flux power value 61.4 Minimum transient DNBR 1.28 62.0 Peak fuel centerline 3800 F temperature 63.7 fiaximum pressurizer pressure 2498 psia 1

L 7 21

TABLE 7.1.6-3 KEY PARAMETERS ASSUMED If1 THE AtlALYSIS OF UtlC0f1 TROLLED CEA UITHDRAWAL FR0f1 ONE PERCENT POWER Parameter Units FSAR Cycle 2 Initial Core Power (f1Wt) 29.0 29.0 Level Core Inlet Coolant (F) 54a.6 544.6 Temperature Core flass Flow Rate 6 (10 1bm/hr) 116.2 116.2 Reactor Coolant System (psia) 2200 2200 Pressure 5 team Generator Pressure (psia) 997 978 Total fluclear Heat 4.68 4.2 Flux Factor fioderator Temperature (10-4ac/F) +0.5 to -3.5 +.5 Coefficient Fuel Temperature 0.85 0.85 Coefficientfiultiplier CEA Gorth on Trip (10-260) -2.4 -5.0 Steam Bypass System Automa tic Manual Feedwater Regulating flanual flanual System Reactivity flaximum (10 4ac/sec) 3.0 1.5 Addition Rate Automatic Withdrawal Inoperative Inonerative Prohibit

TABLE 7.1.6-4 SEQUEf1CES OF EVENTS FOR THE UNC0flTROLLED CEA WITHDRAWAL FR0H Ot1E PERCENT POWER Set Point Time (sec) Event Or Value 0.0 Initiation of Uncontrolled ---

Sequential CEA Withdrawal 25.1 High pressurizer pressure 2422 trip condition 26.0 Trip Breakers Open ___

26.2 Pressurizer Safety Valves 2500 psia Begin to Open 26.3 Shutdown CEAs begin to ---

drop into core 27.6 Peak Core Power Occurs 91 % (of 2815 MWt) 28.2 Peak Core Average Heat Flux 76L (of full power occurs heat flux) 23.2 Minimum DNBR Occurs 11 24 28.5 Peck RCS Pressure Occurs 2662 psia I

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TABLE 7.1.6 5 KEY PARAMETERS ASSUf1ED IN THE ANALYSIS OF UNCONTROLLED CEA WITHDRAWAL FROM FULL POWER Parameter Units FSAR Cycle 2 Initial Core Power (f twt) 2900 2900 Level Core Inlet Coolant (F) 556.7 556.7 Temperature Core fiass Flow Rate (106 1bm /hr) 116.2 116.2 Reactor Coolant System (psia) 2250 2200 Pressure Steam Generator Pressure (psia) 923 939

?!oderator Temperature (10-4ac/F) +0.5 to -3.5 +.5 Coefficient Fuel Temperature -

0.85 .85 Coefficient Multiolier Minimum Available CEA (10-260) -5.4 -5.4 Worth on Trip Steam Bypass System Automatic lianual Feedwater Regulating Automatic Automa tic System flaximun Reactivity (10-4ao/sec) .7 .5 Addition Rate 1

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7-24

TABLE 7.1.6-6 SEQUENCES OF EVENTS FOR THE UNCCflTROLLED CEA WITHDRAWAL FROM FULL POWER Set Point Time (sec) Event Or Value 0.0 Initiation of Uncontrolled Sequential CEA Llithdrawal 14.6 Low DNBR 1.24 (CPC Trip Condition projected) 14.75 Trip Breakers Open 15.05 Shutdown CEAs begin to drop into core 15.15 Peak Core Power Occurs 124%(of'2815ttWt) 15.65 Peak Core Average Heat Flux 120% (of full power) :

occurs 15.65 Minimum DNBR Occurs 21.24 17.6 Peak RCS Pressure Occurs 2425 psia l

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l Arkansas REACTOR COOLANT SYSTEM TEMPERATURES vs TIME 7.1.6-12 l Nuclear One - Unit 2 l

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7- 38

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l-7.1.7 CEA Misoperation

' The CEA Misoperation Events were reanalyzed for Cycle 2 to determine the initial margin that must be maintained by the Limiting Conditions for

-Operation (LCOs) and the CEA position-related penalty factors which are supplied to the CPCs such that low DNBR or high LPD trip will be initiated, when needed, to prevent violation of the DNB and CTM design limits.

Since a CEA drop is the most rapid transient caused by CEA misoperation, the full length CEA drop and the part length CEA drop .were analyzed.

Protection against exceeding the appropriate DNBR or CTM limits is provided by the initial steady state thermal margin which is maintained by adhering to the Technical Specification LCOs and by the CPC low DNBR and high LPD trips.

Each of the four CPCs monitor the position of every CEA group and calculates a penalty factor which accounts for any CEA group misalignment. Each of the two CEAC's monitors the position of every CEA within each CEA group and i

provides a penalty factor to the CPC's which account for any single CEA misalignment up to and including a drop. These penalty factors ensure that the CPC's will provide a conservative estimate of the minimum transient DNBR and maximum LPD during CEA misoperation events and initiate a reactor

trip if needed to prevent violation of the DNBR or LPD design limit.

Single Full Length CEA Drop This event is initiated by dropping a single full length CEA over a ceriod of I second. The nost limiting dropped CEA can cause the radial peaking factor to increase by 17% over the pre-drop value if the core were to achieve stabilized conditions without a trip. This increase may be wfficiently severe to cause the initiation of an immediate CPC low DNBR trip. The CPC low DNBR and high LPD trips monitor the position of every 4

CEA group and receive information from the CEA calculators on penalty ,

j factors appropriate for the single CEA dropped. Consequently, the CPC's provide conservative estimates of transient minimum DNBR and transient t maxinum transient LPD and will initiate a trip if necessary. For less limiting dropped CEAs, an immediate trip is not likely to be generated

, by the CPCs. In the absence of a trip the post drop radial redistribution of xenon will further increase the maximum radial peaking factor.

The CPCs account for this effect dynamically and a low DNBR or high LPD trip signal will be initiated at some later time if required. Tor these cases, tinely reduction of core power can eliminate the need for any ,

j- RPS trips.

The on-line determination of the low DNBR and high LPD trip conditions as performed by the CPCs makes one CEA drop case no more limiting than any other. The CEA drop case selected for presentation is conservatively assumed-to cause the maximum instantaneous increase in radial peaking factor of 170'and to insert the minimum amount of negative reactivity of .101 a.

i t

Table 7.1.7-3 presents the sequence of events for the full length CEA j drop initiated at the conditions listed in Table 7.1.7-1. The transient  !

E behavior of key flSSS parameters are shown in Figures 7.1.7-1 to 7.1.7-4.

The transient initiated at conditions given in Table 7.1.7-1 results in {

-a low DflBR trip within 1.6 seconds which prevents the Of1BR limit of 1.24

, fron being exceeded. The fuel centerline temperature remains well below the CTM limit during the event.

l Part Length CEA (PLCEA) Drop PLCEA drops can produce a rapid approach to the Df1BR limit due to  !

insertion of positive reactivity and distortion of radial and axial power distributions. Protection against exceeding the DflBR limit is F

, provided by the initial steady state thermal margin which is maintained I i

by adhering to the-Technical Specifications LCOs on DilBR margin and by '

the response of the CPCs which provide an automatic reactor trip on low {

D!lBR. '

The nethods used in the analysis are consistent with those used in the '

FSAR. STRIKIN is used to model power and heat flux during the event t and CESEC.is used to model the NSSS response.  :

j The nethods used in selecting limiting cases for analysis are consistent l with those discussed in the FSAR. The PLCEA subgroup drop is more limiting than the single PLCEA drop and thus is the case presented. {

The PLCEA subgroup drop initiated from the nost negative axial shape index i LC0 of .3 and NSSS conditions listed in Table 7.1.7-2 is nost limiting l in that the maxinum positive reactivity is inserted and the subsequent i

power transient leads to the largest and most rapid OflBR margin degradation. l The event is initiated by dropping a part length CEA sabgroup from an initial insertion of 50", and axial shape index of .3 in 1.5 seconds. This results in a positive reactivity insertion of .068p and a more distorted top .

peaked axial power distribution (i.e., a shape index shift of .11 ASI units). Table 7.1.7-4 presents the sequence of parameters for the PLCEA ,

subgroup drop. The transient behavior of key flSSS parameters are shown '

in Figures 7.1.7-5 to 7.1.7-8. ,

The PLCEA subgroup drop initiated from conditions given in Table 7.1.7-2 results in a'CPC low DflBR trip within 4.0 seconds which prevents the Of1BR linit of 1.24 from being violated. The fuel centerline temperature remains well below the CTf1 limit during the event.

The results demonstrate that the limiting PLCEA drop event, initiated frna I within the Tech Spec LCOs, with the intervention of the CPC low DNBR  !

and/or high LPD trips, will not-exceed the design DNBR and CTf1 limits.

i ,

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(_ 7-40 1

TABLE 7.1.7-1 l ASSuf1PTIONS FOR THE FULL LENGTH CEA DROP l

Parameter Units FSAR Cycle 2 Initial Cnre Power (MWt) 2900 2900

Level Core Inlet Coolant (F) 556.7 556.7 Temperature Core Mass Flow Rate 6 116.2 (10 1bm/hr) 116.2 Reactor Coolant System (psia) 2250 2200 Pressure Steam Generator (psia) 923 939

'cessure Dropped CEA Reactivity (10-2ap) (1) .1

'Jorth Ratio of After-Drop ___

(1) 1.17 to Before-Drop Radial Peak Time for Dropped CEA (sec) 1.0 1.0 to fully insert flodera tti iemperature (10-4aa/F) -3.5 -3.5 Coefficient Doppler Coefficient ---

1.15 1.15 Mul tiplier CEA Worth on Trip (10-2ac) -5.3 -5.4 Reactor Regulatin9 ---

Automatic Automatic Systen (2)

(1) See Table 15.1.3-4 of the FSAR for values at two initial axial shape indices (2) Since a reactor trip is initiated shortly after the CEA Drop (.6 seconds for for the limiting case) action of the Reactor Regulating System is not significant.

7-41

TABLE 7.1.7-2 KEY PARAMETERS ASSUMED IN THE ANALYSIS OF PART LENGTH CEA SUBGROUP DROP Parameter Units FSAR Cycle 2 Initial Core Power (Mut) 2900 2900 Core Inlet Coolant (F) 556.7 556.7 Temperature 6

Core Mass Flow Rate (10 1bm/hr) 116.2 116.2 Reactor Coolant System (psia) 2250 2200 Pressure Steam Generator (psia) 923 939 Pressure Axial Shape Index (1) .3 Dropped CEA Reactivity (10-2o0) (1) .068 Time for Dropped CEA (sec) 1.5 1.5 to be fully inserted tioderator Temperature (10~4ao/F) +.5 to -3.5 +.5 to -3.5 Coefficient Doppler Coefficient ---

.85 .85 tiultiplier CEA Worth on Trip (10-2ap) -5.4 -5.4 (II See Table 15.1.3-6 for a family of values at four initial axial shape indices.

The single Cycle 2 case is presented because it is limiting.

I

( 7-42

TABLE 7.1.7-3 SEQUENCE OF EVENTS FOR FULL LENGTH CEA DROP Time (sec) Event Setpoint or Value 0.0 Full Length CEA Drop initiated --

1.0 Full Length CEA Reaches Botto: Of Core --

1.6 Low DNBR Trip Condition 1.24 (CPC Projected) 1.75 Trip Breakers Open --

2.05 Shutdown CEAs Begin to Drop Into Core --

2.5 Core Power Returns to its fiaximum Value 98.3% of 2815 MWL After the CEA Drop 2.75 flinimum DNBR for Transient is Reached [1.24

TABLE 7.1.7 4 SEQUENCE OF EVENTS FOR PART L NGTH CEA SUBGROUP DROP

( .3 ASI Time (sec) Event Setooint or Value 0.0 Part Length CEA Subgroup Drop Initiated --

1.0 Peak Core Power Occurs 128,54 (of 2815 MWt) 1.5 Peak Core Heat Flux Occurs 110.7% (of full power heat flux) 1.5 Part Length CEA Subgroup Redches Botton --

of Core 4.0 Low DNBR Trip Generated 1.24 (CPC Projected) 4.15 Trip Breakers Open --

4.45 Shutdown CEAs Begin to Drop Into Core --

4.8 tiinimum DNBR for Transient is Reached >1.24 i

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POWER & LIGHT CO. FULL LENGTH CEA DROP

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7.1.8 Loss of Coolant Flow Event The Loss of Coolant flow event was reanalyzed for Cycle 2 to determine the mininum initial margin that must be maintained by the Limiting Conditions for Operations (LCOs) and the margin degradation rate which must be projected by the CPCs such that a low DNBR trip will be initiated before the DNBR limit is exceeded.

The methods used to analyze this event are consistent with those discussed in the FSAR with the exception that the design thermal marqin nodel CETOP (discussed in Reference 7-12) was used for all DNBR calculations. ~ The event is presented because of this change in analytical nodels.

The 4-Pump Loss of Coolant Flow produces an approach to the DNBR limit due to the decrease in the core coolant flow. Protection against exceeding the DNBR limit for this transient is provided by the initial steady state thermal margin which is maintained by adhering to  ;

the Technical Specifications' LCOs on DNBR margin and by the response of  :

the RPS which orovides an autonatic reactor trip on low DNBR as calculated by the CPCs.

The Loss of Coolant flow transient is characterized by the flow coastdown curve given in Figure 7.1.8-1 Table 7.1.8-1 lists the key transient para- I neters used in the present analysis. The integrated radial peakina Factor (Fr) was chosen so that the transient is initiated from conditions" which would correspond to a POL.

Table 7.1.8-2 presents tne NSSS and RPS responses during a four pump loss  !

or' flow initiatad froin an axial shape with a negative shape index of .18. i Tne COLS$ and CPC data bases are established parametrically as a function  :

of axial shape index. The representative case shown is one of the limiting cases. The CPCs initiate the low DNBR trip at .6 seconds and the scram r rods start dropping into the core .45 seconds later. A minimum CE-1 DNBR of 1.24 is reached at 2.90 seconds. Figures 7.1.8-2 to 7.1.8-5 present the core power, heat flux, RCS pressure, and core coolant temoeratures as a function of time. Figure 7.1.8-6 presents a trace of hot cnannel DNBR vs. time for the limiting case presented.

The CPC low DNBR trip guarantees that Loss of Ccolant flow events initiated from within the Technical Specification LCO's will not result in a violation of the DNBR design limit.

f l

7-55

TABLE 7.1.8-1 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW AllALYSIS Parameter Units FSAR Cycle 2 Initial Core Power ftWt 2900 2900 Level Initial Core Inlet *F 556.7 556.7 Coolant Temperature Iritial Core Mass 106 lbm/hr 116.2 116.2 Flow Rate Reactor Coolant System psia 2250 2200 Pressure Moderator Temperature 4ao/F +.5 +.5 Coefficient Fuel Temperature --

.85 .85 Coefficient Itultiplier CPC Trip Response Time sec .75 .75 CEA Holding Coil Delay sec .3 .3 CEA tine to 90% sec 3.0 3.0 Insertion (Including Holding Coil Delay)

CEA Wortn at Trip 10-2ap -5.41 -8.00 (all rods out) 4-Pump RCS Flow Coastdown Figure 15.1.5-1 Figure 7.1.8-1 l

l 7-54

i I

TABLE 7.1.8-2 SEQUENCE OF EVENTS FOR l LOSS OF FLOW l

Time (sec) Event Setpoint or Value 0.0 Loss of Power to all Four Reactor --

, Coolant Pumps 0.60 CPC Low DNBR Trip Signal 1.24 (CPC Generated Projected)

.75 Trip Breakers Open --

1.05 Shutdown, CEAs Begin to Drop --

into Core 2.90 Minimum CE-1 DNBR 11.24 i

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Arkansas Nuclear One - Unit 2 MINIMUM HOT CHANNEL CE-1 DNBR vs TIME 7.1.8-6 i

7-61

.- . . ._. .. . _ _ . _ _ - - _ _ - .= _ .- -- . _ _ _ _

l 7.1.10 Transients Resulting From the Instantaneous Closure of a Sinale MSIV

The transients resulting from the instantaneous closure of a single Main Steam Isolation Valve (MSIV) were analyzed for Cycle 2 to determine the  !

CPC Asymmetric Steam Generator Transient Protection (ASGTP) trip setpoint such that in conjunction with the initial margins maintained by the LCOs the  ;

OfiBR and fuel centerline melt (CTM) design limits are not exceeded. .

This event was analyzed for Cycle 1 operation prior to the installation .

of the Asymmetric Stean Generator Transient protection related CPC trip l l

in All0-2, however, the results were not presented in any previous formal  !

writeup. This analysis establishes the reference analysis for future p cycles. ,

The event is initiated by the inadvertent closure of a single main steam E isolation valve causing a loss of load to one steam generator. Upon loss of load, pressure (and temperature) in the affected steam generator increase to the opening [

pressure of the main steam safety values. The intact steam generator " picks up"  ;

the lost load, which causes its temperature and pressure to decrease. The cold  :

leg temperature asyneetry leads to a reactor inlet temperature tilt which pro-duces an azimuthal core power tilt. The most negative moderator temperature  !

coefficient is assumed since this maximizes the power tilt and hot channel  !

radial peaking factor increase. With this assumed sequence of events, the transient results in the greatest asymmetry in core inlet temperature distri- 1 bution, the greatest increase in hot channel radial peaking factors and  !

the most limiting DilBR.

The transient was initiated at the initial conditions given in Table 7.1.10-1. i i

j Table 7.1.10-2 presents the sequence of events for the Loss of Load to l l One Steam Generator. The transient behavior of key NSSS paraneters are i presented in Figures 7.1.10-1 to 7.1.10-5. A reactor trip is generated  !

by the CPCs low DNBR Trip at 2.5 seconds based on high differential l temperature between the cold legs associated with the two steam generators, t The ASGTP trip setpoint within the CPCs ensures that acceptable DNRR limit will not be exceeded during the event, j A maxinum allowable initial linear heat generation rate of 16.5 KU/ft could exist as an initial condition without exceeding the acceptable fuel to center- i

line melt of 21.0 KW/ft during this transient. This amount of margin is i assured by setting the Linear Heat Rate LC0 based on the more limiting of the .

allowable linear heat ra;.e for LOCA and other transients. l The event initiated from the extremes of the LC0 in conjunction with the CPC i l

(ASGT protective) trip will prevent DNBR or centerline fuel temperatures >

t from exceeding the design limits.

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_ __ _ _ 7-62._ _ ___ . _ _ _ , _ _ _ _ _.

TABLE 7.1.10-1 KEY PARAMETERS ASSUtiED IN TliE ANALYSIS OF LOSS OF LOAD TO ONE STEAM GENERATOR Parameters Units Cycle 2 Initial Core Power MWt 2900 Initial Core Inlet F 556.7 Temperature Initial Reactor Coolant psia 2200 System Pressure Moderator Temperature ac/ F -3.5X10-4 Coefficient Fuel Temperature 1.15 Coefficient Multiplier Axial Shape Index asiu .30 l

l l 7-63 i

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TABLE 7.1.10-2 SEQUENCE OF EVENTS FOR LOSS OF LOAD TO ONE STEAM GENERATOR Time (sec) Event Setpoint or Value 0.0 Spurious closure of a single main ---

steam isolation valve 0.0 Steam flow from unaffected steam ---

generator increases to maintain turbine power 3.4 Safety valves open on isolated steam 1093 psia generator 3.6 ASGTP* Analysis setpoint reached 14 F (differential temperature) 3.75 Trip Breakers open ,

4.05 CEAs begin to drop into core ---

6.10 Minimum DNBR occurs >1.24 II. 2 Maximum steam generator pressure 1137 psia t

• ASGTP - Asynnetric Steam Generator Transient Protection i

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7-69

7.2 Postulated Accidents The events in this category were analyzed for operation of Af10-2, Cycle 2 to ensure acceptable consequences. For these transients some amount of fuel failure is acceptable provided the predicted site boundary dose rates meet 10CFR100 guidelines.

The following sections present the results of the analyses, i

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l 7-70

7.2.1 CEA Ejection Event The CEA Ejection event was reanalyzed for Cycle 2 to determine the fraction of fuel pins that exceed the criteria for clad damage.

The analytical method employed in the reanalysis of this event is the NRC approved Combustion Engineering CEA Ejection method which is described in CENPD-190-A, (Reference 7-2).

The procedure outlined in Figure 2.1 of Reference 7-2 is used to determine the radial average and centerline enthalpies in the hottest axial region of the rod. The calculated enthaloy values are compared to threshold enthalov values to determine the amount of fuel exceeding these thresholds. These thresWold enthalpy values are the sane as presented in the FSAR (References 7-3, 7-4, and 7-5).

Clad Damage Threshold:

Total Average Enthalpy = 200 cal /gm Incipient Centerline Melting Threshold:

Total Centerline Enthalpy = 250 cal /gm Fully Molten Centerline Threshold:

Total Centerline Enthalpy = 310 cal /gm Table 7.2.1-1 lists all the key parameters used in the zero power and full power analyses.

To bound the most adverse conditions during the cycle, the most limiting of either the Beginning of Cycle (BOC) or End of Cycle (EOC) parameter values were used in the analysis. A B0C Doppler defect was used since it produces the least amount of negative reactivity feedback to mitigate the transient.

A BOC moderator temperature coefficient of +0.5X10-4ao/ F was used because a positive MTC results in positive reactivity feedback and thus increases both coolant temperatures and stored energy. An E0C delayed neutron fraction was used in the analysis to produce the highest power rise during the event.

The power transient produced by a CEA ejection initiated at the maximun allowed power is shown in Fiaure 7.2.1-1. Similar results for the zero power case are shown in Figure 7.2.1-2. Both full power and zero power cases l are terminated by the high linear power level trip.

The results of the two CEA ejection cases analyzed (Table 7.2.1-2) show that the maximum total energy deposited during the event is less than the l criterion for clad damage (i.e., 200 cal /gm). And, only a small l fraction (1 005) of the fuel reaches the incipient centerline melt threshold.

l 7-71

TABLE 7.2.1-1 XEY PARAMETERS ASSUf1ED IN THE CEA EJECTION ANALYSES Parameter Full Power Units FSAR Cycle 2 Core Power Level f1Wt 2900 2900 Core Average Linear Heat KW/ft 5.53 5.54 Generation Rate at 2815 MWt Moderator Temperature 10-4ap/* F +.5 +.5 Coefficient Ejected CEA Uorth %ao .24 .30 Delayed Neutron Fraction, s ,00497 .00482 flaximum Post-Ejected Radial Power Peak 2.40 2.94 Axial Power Peak 1.61 1.75 CEA Bank Worth at Trip ";ao -5.4 -5.4 Fuel Temperature 0.85 0.85 Coefficient CEA Ejection Time sec .05 .05 Zero Power Core Power Level MWt 1.0 1.0 Ejected CEA Worth %ao 1.10 .82 Post-Ejected Radial Power Peak 9.14 8.23 Axial Power Peak 1.61 2.50 CEA Bank Worth at Trip %Ao -2.4 -2.4

Fuel Temperature 0.85 Coefficient Multiplier 0.85 l

i

• CCA Ejection Time sec .05 .05 1

7-72

TABLE 7.2.1-2 CEA EJECTI0ft EVEtiT RESULTS Full Power FSAR Cycle 2 Total Average Enthalpy of Hottest Fuel Pellet 177.9 156 (cr.1/gm)

Total Centerline Enthalpy of Hottest Fuel 242.9 267 Pellet (cal /gm)

Fraction of Rods that Suffer Clad Damage 0 0 (average Enthalpy 1 200 cal /gm)

Fraction of Fuel Having at least Incipient 0 1 005 Centerline fielting (Centerline Enthalpy 1250 cal /gn)

Fracticn of Fuel Having a Fully fiolten Center- 0 0 line Condition (Centerline Enthalpy 3 10 3 cal /gm)

Zero Power FSAR Cycle 2 Total Average Enthalpy of Hottest Fuel Pellet 275 164 (cal /gm) i Total Centerline Enthalpy of Hottest Fuel 280 296 Pellet (cal /gm)

Fraction of Rods that Suffer Clad Damage .04123 0 ,

(Average Enthalpy ; 200 cal /gm)

Fraction of Fuel Having at least Incipient .01805 1 005 ,

Centerline Melting (Centerline Enthalpy 1 250 cal /ga)

Fraction of Fuel Having a Fully Molten Centerline 0 0 Condition (Centerline Enthalpy 1310 cal /gm) i l

7-73

3 , , , ,

FULL POWER a

"2 -

c5 z

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U E .

W o ,

a 1 -

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0 1 2 3 4 5 TIME, SECONDS 4

f POW & LI HT CO, CEA EJECTION EVENT Figure Arkanso$ CORE POWER vs TIME l Nuclear One - Unit 2 7.2.1-1 l

7-74 '

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7.2.2 Steam Line Break-Incident The full power Steam Line Break (SLB) events were re-evaluated for Cycle 2 to demonstrate that results presented in the FSAR are still bounding even when the following changes in Cycle 2 parameters are accounted for:

1. Reduced Shutdown flargin

2. Increased Doppler Feedback

3. Decreased Reactivity Insertion During Moderator Cooldown The steam generator blowdown and associated Reactor Conlant System (RCS) cooldown were not recalculated for Cycle 2 since the net effect of changes in the above parameters on the blowdown will be small.

The major impact of the changes in these parameters is on the time dependent reactivity variation. Thus, only these reactivity variations were evaluated to assess whether there is an increased potential for return to power. The reactivity assessment was made by combining the transient plant response generated for the FSAR analysis with the Cycle 2 shut-down margin and reactivity feedbacks. The minimum available Cycle 2 full power shutdown worth is -7.9 N as compared to -8.6 %; for Cycle 1.

The .7%w decrease in shutdown worth and increased positive doppler reactivity insertion is offset by the decreased positive reactivity insertion due to moderator cooldown associated with the Cycle 2 cool-down curve. To evaluate these tradeoffs the following SLB cases were re-evaluated.

1. 2 loop - full power (2900 Milt), without Loss of AC (LOAC)

2. 2 loop - full power (2900 Milt), with Loss of AC (LOAC)

The results of the re-evaluation are presented in Figures 7.2.2-1 and 7.2.2-2 for the no-LOAC and LOAC cases, respectively. On each figure the corresponding Cycle 1 and Cycle 2 results are presented. The Cycle 1 cases are based on cooldown curve associated with an initial allowable MTC of -3.5x10-4L/*F, while the Cycle 2 cases are based on the cool-down curve associated with an initial allowable MTC of -2.8x10-44-/ F.

Comparison of the Cycle 1 and Cycle 2 results from these curves shows that the positive reactivity insertion due to cooldown of the moderator is less for Cycle 2 by 1.1%p at the time of minimum negative reactivity (i.e., 60 seconds for the no-LOAC case and 74 seconds for the LOAC case).

This improvement in the moderator cooldown behavior is sufficient to completely offset the .7%M decrease in available shutdown worth and

.2%s increase in positive reactivity insertion due to Doppler feedback.

l from the re-evaluation of the SLB events presented above it has been demonstrated that the peak reactivity experienced during the transient for Cycle 2 is bounded by the FSAR results. On this basis it can be conciuded that the FSAR results are conservative and that the conclusions presented in the FSAR remain valid for Cycle 2.

7-76

10.0 , i i i i i i i i

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,., CYCLE 2

, 6. 0 -

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-6.0 -

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-10.0 ' ' ' '

0 20 40 60 80 100 120 140 160 180 200 '

TIME, SECONDS I

l ARKANSAS STEAM LINE RUPTURE INCIDENT WITHOUT LOSS OF Figure l POWER & LIGHT CO. AC PWR TWO LOOP FULL PWR INITIAL CONDITION l

Arkansa$

l Nuclear One 'Jnit 2 N0ZZLE BREAK WITHOUT MOISTURE CARRYOVER 7.2.2-1 l

7-77

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CYCLE 2

-6.0 - -

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-10.0 0 20 40 60 80 100 120 140 160 180 200 220 TIME, SECONDS l

ARKANSAS Figure STEAM LINE RUPTURE INCIDENT WITH LOSS OF AC POWER & LIGHT CO.

^' k "5 5 POWER TWO LOOP FULL POWER INITIAL CONDITION

7. 2. 2 -2 Nuclear One " Unit 2 N0ZZLE BREAK WITHOUTMOISTURE CARRYOVER 7-78

i

. 7.2.3 Feedwater Line Break Accident  ;

The feedwater line break accident was reanalyzed for Cycle 2 to determine j that the RCS pressure upset limit of 2750 psia is not exceeded during the i' transient. ,

t 1

The rupture of a feedline will cause capid reduction of the liquid inventory '

in the affected steam generator and therefore partial loss of the secondary  ;

heat sink. This leads to the heatup of the RCS and an increase in primary i i pressure. Depending on initial conditions, break size, break locations and steam generator inventory, any of the several Plant Protective System (PPS) sctions may occur. A decrease in the steam generator water level will ,

initiate a reactor trip on low steam generator water level. The decrease .

in the steam generator pressure may result in a low steam generator pressure l trip signal and cause the main steam isolation valves and the main feed-water isolation valves to close. The partial loss of the secondary heat i sink causes the RCS to heat up. This may result in a high RCS pressurizer  ;

l trip. Additional protection against complete loss of secondary heat sink is provided by automatic initiation of emergency feedwater to the intact j stean generator. .

The feedwater line break analyzed was assumed to occur during fu.1 power  ;

operation and with concurrent loss of non-emergency A-C power at time of i trip. This is limiting from the standpoint of potential RCS pressure i 4 increase, since this results in the maximum initial stored energy and mininum steam generator inventory. In addition, in response to loss i of non-emergency AC power upon trip, the following will occur to maximize RCS pressure increases

1. Turbine trip valves close immediately;

, 2. Reactor coolant pumps begin to coastdown;  ;

3. Pressurizer control systems are lost; and

! 4 112.4 sec rather than 97.4 sec are required for the automatic initiation of emergency feedwater to the unaffected steam i

generator.

t The limiting breaksize was determined by parametric study performed with  !

the methodology reported in tne FSAR.  !

The feedwater line break event was initiated at the conditions shown in Table 7.2.3-1. This combination of parameters maximizes the calculated RCS peak pressure. Table 7.2.3-2 and Figures 7.2.3-1 through 7.2.3-5 show the NSSS and RPS responses for this event.

l The results indicate that the reduction of the secondary heat sink due to the discharging of saturated water through the feedwater line break and the i subsequent emptying of the affected steam generator cause the RCS pressure to increase to 2705 psia. Following reactor trip on either high pressurizer  :

l pressure or low steam generator water level, the decay in core power and the t

! action of the primary and secondary safety valves result in a reduction of the RCS pressure. Subsequently, the competing effects of system flow coastdown due to loss-of-AC upon trip, continued blowdown of steam from the intact stean generator through the break and the entering of emergency feedwater to the intact steam generator 7-79

cause the RCS first to go through a mild pressure increase and then a steady decrease. The decrease is reversed when the low steam generator pressure initiates the closure of Main Steam Isolation Valves (MSIV).

The f1SIV closure terminates the blowdown of steam through the break thus causing the RCS to heat up once more. Eventually, the heatup is terminated by the opening of secondary safety valves.

The results of this analysis demonstrate that the Feedwater Line Break '

Event will not result in a peak RCS pressure which exceeds the upset pressure limit, i

7-80

TABLE 7.2.3-1 ASSUMPTIONS FOR FEEDWATER LINE BREAK ANALYSIS Parameter FSAR Cycle 2 Initial Core Power Level,11Wt 2900 2900 Initial Inlet Coolant 556.7 556.7 Temperature, F Core Mass Flow Rate, 116.2 116.2 106 lbm/hr Steam Generator Pressure, psia 923 939 RCS Pressure, psia 2200 2300 Moderator Temperature Coefficient 0.5 0.0 (10-4 /4/ F)

Fuel Temperature Coefficient 0.85 0.85 Multiplier flinimum CEA Worth at Trip, -5.4 -8.0 10-2u.

l l 7-81

TABLE 7.2.3-2 SEQUENCE OF EVEflTS FOR THE FEEDWATER LIllE BREAK ACCIDEflT WITH LCSS OF A-C Time (sec) Event Setpoint or Value 0.0 Initiation of rupture ---

31.4 Affected steam generator empties ---

31.4 High pressurizer pressure trip condition 2422 psia occurs 31.4 Low steam generator water level trip 5% of condition occurs in intact steam generator instrument range 31.4 A-C power is lost ---

32.3 Trip breakers open ---

32.5 Pressurizer safety valves open 2500 psia 32.6 Shutdown CEAs begin to drop into core ---

35.5 Peak RCS pressure occurs 2705 psia 38.2 Steam Generator Safety valves open 1093 psia 42.2 Pressurizer safety valves close 2400 psia 52.2 Steam Generator Safety Valves close 1093 psia 143.8 Emergency feedwater enters intact steam ---

generator 165.0 Steam generator low pressure trip 678 psia condition and f1 SIS initiated 170.9 Complete closure of flain Steam Isolation Valves- ---

terminating blowdown from the intact steam genera tor 275.0 Intact Steam Generator Safety Valves open 1093 psia 7-82

l s

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$ 60 -

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8 20 -

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O 100 200 300 400 500 TIME, SECONDS POW & Li HT CO. FEEDWATER LINE RUPTURE WITH LOSS OF AC POWER Figure

^'k o"5 5 Nuclear One - Unit 2 CORE POWER vs TIME 7.2.3-1 7-83

120 i i i i 100 3 -

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3000 i i i i 2800 - _

2600 - -

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5l h

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2000 - _

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1800 0 100 200 300 400 500 TIME, SECONDS ARKANSAS Figure POWER & LIGHT CO. FEEDWATER LINE RUPTURE WITH LOSS OF AC POWER

^ *ansas REACTOR COOLANT SYSTEM PRESSURE vs TIME Nuclear One - Unit 2 7.2.3-3 7-85

i 650 630 -

4 TOUTLET

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g590 4 -

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550 - -

530 I I I i 0 100 200 300 400 500 TIME, SECONDS PowNYt$. oN co.FEEDWATER LINE RUPTURE WITH LOSS OF AC POWER Figure Arkansas Nuclear one - Unit 2 REACTOR COOLANT SYSTEM TEMPERATURE vs TIME 7.2.3-4 7-86

1200 i i i i 1000 -

E N 800 - -

5?

EE g 600 _ _

!E e5 5

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0 100 200 300 400 500 TIME, SECONDS l

i l

POW & LIGHT CO. FEEDWATER LINE RUPTURE WITH LOSS OF AC POWER Arkansas STEAM GENERATOR PRESSURE vs TIME  !

Nuclear One - Unit 2 7.2.3-5 7-87

7.2.5 Seized Shaft fvent The Seized Shaft event was reanalyzed for Cycle 2 to demonstrate that the RCS pressure limit of 2750 psia will not be exceeded and the amount of fuel pins credicted to fail during this event will not cause the 10CFR100 site boundary dose guidelines to be exceeded. The event is presented because of changes in the input parameters and because the methods used in the Loss of Flow analysis to determine the initial thermal margin have changed.

The single reactor coolant pump shaft seizure is postulated to occur as a consequence of a mechanical failure. The single reactor coolant pump shaft seizure results in a rapid reduction in the reactor coolant flow from the four-pump to the three-pump value. The rapid reduction in coolant flow results in a rapid reduction in the margin to DNB, so that a CPC low DflBR trip occurs to terminate the transient.

The initial conditions for the Seized Shaft event are listed in Table 7.2.5-1.

These conditions are consistent with the initial conditions assumed for the

~

LOF event (see Section 7.1.8). Other assumptions on key parameters are also listed in this table.

Consistent with the FSAR methodolnay, the analysis was performed in the following steos:

A. Upon initiation of this transient, core flow rate rapidly reduces to the asymptotic three pump core flow value of 74.6% of four pump flow.

The flow coastdown is given in Figure 7.2.5-1.

B. The resultant flow is used as input to CESEC, which simulates the (1555 to denonstrate that the reactor coolant system (RCS) pressure will remain below the upset limit of 2750 psia (110% of design).

C. CETOP/CE-1 as described in Reference 7-12 was used for calculation of DriBR during the transient. The seized shaft transient is initiated at the Limiting Conditions for Operation to determine the minimum DriBR.

D. The DNBR calculation is repeated with the above steps using CETOP for fuel pins of various radial peaks. An integral fuel damage calculation is then carried out by combining results from the CETOP code with the number of fuel pins having a given radial peaking factor. The number of fuel pins versus radial peaking factor is taken from a distribution of the fraction of fuel pins with nuclear radial peaking factors in a given range.

The fiSSS and RPS responses are shown in Table 7.2.5-2 for the seized shaf t event initiated from an axial shape ir.dex value of .18. This case is selected to be consistent with the Loss of Flow case presented in Section 7.1.8.

l The pressurizer pressure reached a maximun value of 2294 psia at 3.95 seconds.

j Figures 7.2.5-1 through 7.2.5-5 show core flow, core po'. er, core average heat l flux, RCS pressure, and coolant tenperatures during the transient.

7-88

i l

A conservatively " flat" pin census distribution (a histogram of the number of pins with radial peaks in intervals of 0.01 in radial peak normalized to the maximum peak) is used to determine the number of pins that experience DNB. The results show that the number of fuel pins predicted to experience DNB is less than 2.0%. This is the same value reported in the FSAR.

For the case of the loss of coolant flow resulting from a seizure of a reactor coolant pump shaft, a trip on low DNBR is initiated to limit the minimum value of DNBR and the total number of pins with DNBR's below 1.24 for a short period of time. As shown in the FSAR, this assures that the radiological releases are much less than the 10CFR100 guidelines. In addition, the maximum RCS pressure experienced during the event will be well under the upset pressure lim:. of 2750 psia.

7-89

TABLE 7.2.5-1 KEY PARAMETERS ASSUMED Iti SEIZED SHAFT Af4ALYSIS Parameter Units FSAR Cycle 2 Initial Core Power Level MWt 2900 2900 Core Inlet Coolant Temperature F 556.7 556.7 4 Pump Core Mass Flow Rate 1061bm/hr 116.2 116.2 3 Pump Core Mass Flow Rate 106 1bm/hr 89.8 86.6 Reactor Coolant System Pressure psia 2250 2200 Moderator Temperature Coefficient X10-4ap/ F +.5 +.5 Fuel Temperature Coefficient ---

.85 .85 Multiplier CEA Tlorth on Trip %ao -5.4 -8.0 i

l i

7-90

TABLE 7.2.5-2 SEQUENCE OF EVENTS FOR THE SEIZED SHAFT ANALYSIS Time (sec) Event Setpoint or Value 0.0 Shaft seizure on one reactor ---

coolant pump

.77 Low DNBR trip signal generated 1.24 (CPC Projected)

.92 Trip breakers open ---

1.22 Shutdown CEAs begin to drop ---

into core '

2.00 Minimum DNBR for the transient 1.014 occurs 3.00 Core flow reaches asympt' otic 6 86.6x10 1bm/hr three-pump 3.90 Maximum RCS pressure 2294 psia 7-91

1.00 . , , , ,

0.95 -

@0.90 -

cd w

e O

o p 0. 85 -

z S

s o ,

E 0. 80 -

s 0.75 - -

0.70 ' '

O 1 2 3 4 5 TIME, SECONDS ARKANSAS POWER & LIGHT CO. SEIZED SHAFT Figure Akonsm Nuclear One - Unit 2 CORE FLOW vs TIME 7.2.5-1 7-92

=-. . - . . . .

f 1

120 i i i i i 100 -

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l 0 ' ' i i i l 0 1 2 3 4 5 6 TIME, SECONDS POW t.lGHT CO. SEIZED SHAFT Figure

^'ko"5 5 Nuclear One - Unit 2 CORE POWER vs TIME 7.2.5-2 7-93 _. ..-. - - --

110 , , , i i x

3

'- 100 85 s:

O 90 -

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50 ' ' ' i i 0 1 2 3 4 5 6 TIME, SECONOS ARKANSAS POWER & LIGHT CO. SEIZED SHAFT Figure Arkonsos Nuclear One - Unit 2 CORE HEAT FLUX vs TIME 7. 2. 5-3 7-94

DT i i i i i 5

E m 2280 5

u o- 2260 _

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C w

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5 8

o  :

x S 2220 -

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6 x

2200 ' i i 0 1 2 3 4 5 6 TIME, SECONDS l

l ARKANSAS POWER & LIGHT CO. LOSS OF COOLANT FLOW Figure SEIZED SHAFT Nucleo ne - Unit 2 REACTOR COOLANT SYSTEM PRESSURE vs TIME 7.2.5-4 7-95

650 , , , , ,

cy- 630 -

N CORE OUTLET COOLANT TEMP.

?

c.

610

E W

s b590 -

8 CORE AVERAGE COOLANT TEMP.

8

!3 g 570 -

e:

CORE INLET COOLANT TEMP.

550 ' ' '

0 1 2 3 4 5 6 TIME, SECONDS POW LIGHT CO- LOSS OF COOLANT FLOW Figure Arkansas SEIZED SHAFT Nuclear One - Unit 2 REACTOR COOLANT TEMPERATURE vs TIME 7.2.5-5

11.0 Technical Specifications This section provides all recommenced changes that should be made to the Technical Specifications in order to update the Technical Specifications for Cycle 2 operation of ANO-2.

A list of all recommended Technical Specifications changes anc expianations are provided in Table 11-1.

Each page from the Technical Specifications which must be modified is shown with the modification noted.

)

11-1

TABLE 11-1 Arkansas fluclear One - Unit 2 Cycle 2 Reconnended Technical Specification Changes CHANGE # TECH. SPEC. # PAGE # ACTI0fl EXPLAtlATI0ft I IflDEX I Add af ter "Sof tware" " Planar Radial In accordance with Tech. Spec. change #6 Peaking Factor . . 1-6" 2 INDEX IV Change the title of 3/4.2.6 to In accordance with Tech. Spec. change #32 read " Reactor Coolant Cold Leg Temperature" Change page numbers for Tech. Specs. In accordance with Tech. Spec. change

$ 3/4.2.2, 3/4.2.3 and 3/4.2.4 to #'s 26, 27, and 28.

rl, 3/4 2-4, 3/4 2-5 and 3/4 2-7 respect-ively.

3 INDEX IV After 3/4.2.6 add: In accordance with Tech. Spec. change #32 "3/4.2.7 AXIAL SHAPE IflDEX.. 3/4 2-12" "3/4.2.8 PRESSURIZER PRESSURE.3/4 2-13" 4 INDEX IX Change the title of Bases 3/4.2.6 In accordance with Tech. Spec. change #32 to read " Reactor Coolant Cold Leg Temperature" 5 INDEX IX After Bases 3/4.2.6 add: In accordance with Tech. Spec. change #32 "3/4.2.7 AXIAL SHAPE INDEX..B3/4 2-4" "3/4.2.8 PRESSURIZER PRESSURE. 83/4 F -4"

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 6 1.27 1-6 Added in order to standardize Add PLANAR definition of Fxb FACTOR - Fxy.Technical RADIAL PEAKIN Specification Definition 1.27 The PLANAR RADIAL PEAKING and symbol with other C-E plants.

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.

7 2.1.1.1 2-1 Change all references to DNBR limit Minimum DNBR changed to reflect the from "1.30" to "1.24" change in methodology from COSM0/W-3 to TORC /C-E-1.

2 0

8 2.2.1 2-4 Change "ECCS" to " Safety" Footnote applies to non-LOCA analyses Table 2.2-1 as well as ECCS Footnote 9 2.2.1 2-4 Change Linear Power Level-High trip Setpoint lowered to lowest possible value Table 2.2-1 setpoint from "<123%to "< 110%" and without impacting plant operation.

Functional allosable limi ts from"<123.712%" to 2

gnig#,4, 7 nd '1110.712%" (Functional Unit 2) and Change trip setpoints and allowable 8

Changes to trip setpoints and allowable limi ts for functional Units 4, 5, 6, limits were necessitated by the Fisher-7 and 8. Porter /Rosemount transmitter changeout.

10 2.2.1 2-5 Change DNBR- Low trip setpoints from Minimum DNBR trip setpoint and allowable Table 2.2-1 > l limits changed to reflect the change in Functional "ITmits l.3" tofrom "z'>.24" and 1.3" allowable to ">l.24" and methodology from COSM0/W-3 to TORC /CE-1 Unit #10 &

11 Change trip setpoint and allowable Changes to trip setpoint and allowable limits for Functional Unit 11 limits were necessitated by the Fisher-Porter /Rosemount transmitter changeout.

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 11 2.2.1 2-5 In flote 2, change "as pressurizer Clarification of RPS trip setpoint Table 2.2-1 pressure is reduced" to "during a reduction during Modes 2 and 3 operation, Table flotation planned reduction in pressurizer in accordance with Safety Analysis pressure", and assumptions.

In flote 3, change "as steam generator pressure is reduced" to "during a planned reduction in steam generator pressure".

12 B.2.1.1 82-1 Change U-3 DilBR correlation and 1.3 Minimum DilBR Limit chang' ed to reflect limi t, to CE-1 correlation and 1.24 changes in methodology from COSM0/W-3 respectively. to TORC /CE-1.

r 1.

13 B.2.2.1 B2 Change Limiting Safety System Setting Minimum DNBR Limit changed to reflect for low OflBR from "1.30" to "1.24" change in methodology from COSM0/W-3 to TORC /CE-1 14 B.2.2.1 B2-5 Change "a margin of at least 10 The 10 minute nargin is not required for minutes" to " sufficient margin" ANO-2 due to automatic initiation of emergency feedwater and is not supported by Safety Analysis.

15 B.2.2.1 82-6 Change low DflBR limi t of 1.30 to 1.24 Minimum DflBR has been changed to reflect changes in methodology from COSMOS /W-3

, to TORC /CE-1 16 3/4.1.1.2 3/4 1-3 Change minimum required shutdown The shutdown margin has been increased to margin from 1.0%Ak/k to 5.0%Ak/k lengthen the time available for operator action for the' Boron Dilution analysis.

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 17 3.1.1.4 3/4 1-5 Replace 3.1.1.4a and 3.1.1.4b with a. No change

a. Less positive then 0.5x10-4Ak/k/ F b. Changed to yield acceptable results whenever TilERMAL POWER is <70% of for the Feedwater Line Break analysis RATED THERMAL POWER. c. Changed to yield acceptable results

b. Less positive then 0.0Ak/k/ F for the Full Power Steam Line Break whenever TilERMAL POWER is >70% of analysis.

RATED THERMAL POWER, and

c. Less negative then -2.8 x 10-4A k/k/

F at RATED THERMAL POWER.

18 3.1.2.2 3/4 1-8 Change minimum required shutdown Changed in accordance with Tech. Spec.

margin from 1%Ak/k to 5%Ak/k 3.1.1.2 change #16

=.

19 3.1. 2. 4 3/4 1-10 Change minimum required shutdown Changed in accordance with Tech. Spec.

margin from 1%Ak/k to 5%tk/L 3.1.1.2 change #16 20 3.1.2.6 3/4 1-12 Change minimum required shutdown Changed in accordance with Tech. Spec.

margin from 1%tA/k to 5%ak/k 3.1.1.2 change # 16 21 DELETED 22 3.1.2.8 3/4 1-15 Change minimum required shutdown Changed in accordance with Tech. Spec.

Action a margin from 1%Ak/k to 5%Ak/k 3.1.1.2 change # 16

TABLE 11-1 (Continued)

CHANGE # . TECH SPEC. # PAGE # ACTION EXPLANATION 23 3.1.3.1 3/4 1-17 Remove Action c and insert as a note Previous ACTION C requires no specific thru under LCO. Reletter Action items ~

action be taken nor does it impose any

.3/4 1-19 from d thru g to c thru f operating restriction. It should become a NOTE to clarify the LCO.

24 3.1.3.6 3/4 1-27 Replace Figure 3.1-2 with enclosed The PDIL has been changed to yield Figure 3.1-2 Figure 3.1-2 adequate shutdown margin in MODES 1 & 2 25 3.2.1 3/4 2-1 Change Tech. Spec. 3.2.1 to read: Previous wording has been found to be "The linear heat rate margin shall be difficult to interpret. Cycle 1 wording maintained by operating within the implies that COLSS outputs linear heat region of acceptable operation of rate values in kw/ft. However, COLSS only

_. Figure 3.2-1 or 3.3-2 as applicable." displays a power operating limit value in 7

m percent power. New wording is compatible with COLSS and is similar to the related DNBR Tech. Spec. 3.2.4 26 3.2.1 3/4 2-2 Change Figure '3.2-1 to 3.2-2, page Also replaces one figure with two Figures 3.2-1 3/4 2-2 to 3/4 2-3 and add the Figure 3.2-1 is for COLSS operable and 3.2-2 following note under the title Figure 3.2-2 is for COLSS out of service

("COLSS OUT OF SERVICE")

Insert new Figure 3.2-1 as page 3/4 2-2 27 3.2.2 and 3/4 2-3 Change all references to Fr" and FrP Changed in accordance with the new Tech.

4.2.2.2 to Fx/1 and Fxyc respectively, change Spec. Definition 1.2.7 (Change #6) and due all references to planar radial to additional figure in Change #26 peaking factors to capitalized type and change page number from 3/4 2-3 to 3/4 2-4 28 3/4.2.3 3/4 2-4 Change page numbers from 3/4'2-4 thru Page and figure numbe'r changes due to and to Tech. Spec. change #26 3/4 2-6 to 3/4 2-5 thru 3/4 2-7 3/4.2.4 3/4 2-6 a, wrwt .-w- w - w .- - , .=- ,-__

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 3/4.2.3 3/4 2-4 (continued) and to Change figure numbers in Tech. Spec.

3/4.2.4 ,/4 3 2-6 3.2.4 from 3.2-2 and 3.2-3 to 3.2-3 and 3.2-4 Correct spelling of AZIMUTHAL in 3.2.3 Correction of typographical error 29 4.2.4.4 3/4 2-7 Delete page 3/4 2-7 As described in CEN-139(A)-P, rod bow penalty factor is implicitly accounted for in the minimum DNBR 1imit of 1.24, thus eliminating the need for Tech. Spec.

4.2.4.4 T

NJ 30 3.2.4 3/4 2-8 Change figure number from 3.2-2 to Figure number changed due to Figure 3.2-2 3.2-3 Tech. Spec. change #26 31 3.2.4 3/4 2-9 Change figure number to 3.2-4 and New figure number due to figure number Figure 3.2-3 replace with enclosed Figure 3.2-4 changes in Tech. Spec. change #26 and reanalysis of COLSS out of service DNBR margin requirements for Cycle 2.

32 3.2.6 3/4 2-11 Replace Tech. Spec. 3.2.6 (page 3/4 To incorporate into the Tech. Spec's the 2-ll)with enclosed Tech. Specs. 3.2.6, limits on DNB related parameters assumed

" REACTOR COOLANT COLD LEG TEMPERATURE", in the Safety Analys.s.

3.2.7 " AXIAL SHAPE INDEX" and 3.2.8

" PRESSURIZER PRESSURE (pages 3/4 2-11 thru 3/4 2-13) 33 3.3.1.' 3/4 3-2 for Functional Unit 3.a remove Mode 1 Table 3.3-1 as an Applicable Mode In accordance with Table 2.2-1 of Tech.

Functional Spec 2.2.1 trip setpoint is +0.7S% rated Unit 3.a thennal power and can be bypassed at

/ >10-4 rated thermal power.

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 34 3.3.1.1 3/4 3-2 Und' Applicable Modes change "1,2" The additional modes of applicability for Table 3.3-1 to "1,2 and *"' for both Funct al these trip functions is required to yiele Functional Units 5 and 7 acceptable results for Steam Line Break Units 5. and 7 -

analysis. (In conjunction with Tech.

Spec. 2.2.1, Table 2.2-1 Notation clarification, Change # ll) 35 3.3.1.1 3/4 3-3 Under Action change "5 and 6" to Changed to be made consistent with ACTION Table 3.3-1 "S# and 6" for Functional Unit 15 requirements of CPC's (Functional Unit 14' Functional allowing mode changes while under ACTION Unit 15 requirements 36 3. 3.1.1 3/43-2, The Table is noted,where appropriate The PPS is designed and built as a three

Table 3.3-1 3-? 3-4, with new footnote (g) and Action channel system with an installed spare.

& Nota tion 3-5 statement 2 is revised consistent with This change treats the PPS as a three Action 2 three channel system operation. channel system rather than a four channel system.

37 3.3.1.1 3/4 3-5a Change "28% of RATED THERMAL POWER" Reanalysis' of CEAC inoperable margin Table 3.3-1 to "> 11 % of RATED THERMAL POWER. requirements.

Action 5-b.1 38 4.3.1.1.1 3/4 3-7 In Table 4.3-1 under surveillance Clarification of surveillance requirements:

Table 4.3-1 modes, change "1,2" to "1,2 and *" in accordance with Tech. Spec. 3.3.1.1 Functional for Functional Units 5 and 7. change # 34 Units 5 and 7 O

it i'

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 39 4.3.1.1.1 3/4 3-11, The Table is noted where appropriate Clarification of surveillance requirement Table 3.3-3 3-12,3-13 with new footnote (d) and Action in accordance with Tech. Spec. 3.3.1.1 Nota tion 3-14 statement 9 is revised consistent with change #36 Action 9 three channel system operation 40 3.3.2.1 3/4 3-16 Change trip values and allowable Changes to trip values and allowable Table 3.3-4 thru values for Functional Units 1, 4, 5, values necessitated by Fisher-Perter /  ;

Functional 6 and 8.

Rosemount transmitter changeout Units 1,4,5,6 3/4 3-18 and 8 3.3.2.1 3/4 3-18 In Note 1 channe "as pressurizer Clarification of ESFAS function setpoint .

Table 3.3-4 pressure is reduced" to "during a reduction during Modes 2 and 3 operation, Notation planned reduction in pressurizer in accordance with Safety Analysis

nressure". and assumptions. (In conjunction with Tech.

j, In Note 2 change "as steam generator Spec. 2.2.1 Table 2.2-1 Notation i pressure is reduced" to "during a clarification, Change #11) {'

planned reduction in steam generator pressure" .

i i -

M d

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

TABLE 11-1 (Continued)

CHANGE # TECli SPEC. # PAGE # ACTION EXPLANATION 41 3.4.1 3/4 4-1 Insert after Modes 1 and 2 Part loop operation has not been approved Action by the NRC. An Action statement has been Four Pump Ope ra ti on* *

• added to require shutdown in the event of With less than four reactor coolant a loss of four reactor coolant pump pumps in operation be in at least conditions.

HOT STANDBY within one hour.

Pa rt Loop Operation ***

42 3.4.1 3/4 4-1 Add footnote "***" In accordance with Tech. Spec change #41 Footnote "*** Part loop operation is not allowed in Modes 1 and 2 pending NRC approval of sa fety analyses."

d k 43 ,

3.4.1 3/4 4-1 Change "ECCS" to " Safety" in Action Footnote applies to non-LOCA analyses Action No ta tion " **" as well as ECCS, Notation 44 3.4.1 3/4 4-2 Add the following action item before Wording added to delineate minimum RC. pump Ac tion Action " MODES 3, 4 and 5:" " MODE 3: requirements for operation in Mode 3 as Operation may proceed provided two compared to Modes 4 & 5 so as to yield reactor coolant loops are in operation acceptable results in Steam Line Break with at least one reactor coolant analysis (only 2 loop operation has been pump in each loop With less than one analyzeo).

reactor coolant pump in each loop -

in operation , have at least one pump in each loop in operation within one hour or be in at least 110T SitUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Change

" MODE 3, 4 and 5:" to " MODES 4 and 5:

45 4.6.3.1.2 3/4 6-18 Change valve number under penetration Correctior: of typographical error.

Table 3.6-1 #2P18 from cCV-3847-2 to 2CV-4847-2.

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 46 3.7.1.1 3/4 7-3 Change "ECCS" to " Safety" in Table Footnote applies to non-LOCA analyses Table 3.7-2 3.7-2 notation as well as ECCS Footnote The safety analysis regarding EFW inspectio 47 4.7.1.2 3/4 7-5 Delete from surveillance requirement Iis only concerned with flow rate and dis-4.7.1.2.a.1 "when the secondary steam j charge pressure. Inlet steam pressure and supply pressure is greater than 865 turbine speed are not critical parameters.

osig and the pump speed is 13600 rpm" Pump degradation is monitored for by ASME Section XI testing which AP&L has conmitted to for this pump.

48 3/4.7.1.4 3/4 7-8 Change "10 10pCi/ gram DOSE EQUIVALENT Changes dealing with secondary coolant and Action I-131" to "10 046pCi/ gram DOSE activity required to yield acceptable EQUIVALENT I-131" (in two places) and results for Safety Analyses involving add a second surveillance requirement secondary atmospheric releases.

_, (4.7.1.4.2)

?

C The ANO-2 refueling machine does not 49 3/4.9.6 3/4 9-7 Change "CEAs or fuel assemblies" to include a CEA mast. CEAs are moved with

" fuel assemblies in 3/4.9.6 a cunual tool or with a CEA change mech-anism located over the fuel transfer 50 B3/4.1.1.1 anc B3/4 1-1 Change shutdown margin from "l%Ak/k" B3/4.1.1.2 to "5%Ak/k" Changed in accordance with Tech. Spec.

3.1.1.2 change #16 51 B3/4.1.2 B3/4 1-2 Change shutdown margin from "1.0fak/k" Changed in accordance with Tech. Spec.

to "5.0%Ak/k," 3.1.1.2 change # 16 Change "40,200 gallons of 1731 ppm The minimum required horated water volume borated water" to "56,455 gallons of has been changed to meet requirements of 1731 ppm borated water." Cycle 2 due to the higher core average enrichment.

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 52 B3/4.1.2 83/4 1-3 Change "4,700 gallons of 1731 ppm The volume of borated water has been borated water" to "8185 gallons of changed due to the increased shutdown 1731 ppm borated water" margin for Mode 5. (Tech. Spec. 3.1.1.2)

Cor t of typgraphical errors, 5 0 all r a 1 IlllS to a e utdown margin from "l%Ak/k to ^cco ance with Tech. Spec.

n 5%Ak/k.

53 B3/4.2.1 B3/4 2-1 Change "Fr measurement uncertainty Changed in accordance with the new Tech.

factor of 1.080" to "Fxy measurement Spec definition 1.2.7 (Change #61 New uncertainty factor of 1.053" measurement uncertainty factor calculated and as per approved topical CENPD-153-P Rev l-P-A, May 1980.

Change figure number f r^ n 3.2-3 to Changed in accordanc,e with Tech. Spec.

3.2-4 Change #27 5 B3/4.2.2 B3/4 2-2 c" N

54 Change"Fr

" " to " Fx , "Fr"" to Changed in accordance with the new Tech.

"F , and all ref rences to planar Spec definition 1.2.7 change #6 rabfal peaking factors to capitalized type.

55 83/4.2.4 B3/4 2-3 Change "Fr measurement uncertainty Changed in accordance with the new Tech.

factor of 1.080" to "Fxy measurement Spec definition 1.2.7. (Change #6). New uncertainty factor of 1.053" Measurement uncertainty factor calculated as per approved topical CENPD-153-P Rev l-P-A, May, 1980.

Change DNBR value from "1.30" to "1.24 ' Minimum DNBR changed to reflect the change and in methodology from COSM0/W-3 to TORC /CE-1 Change figure numbers from 3.2-2 to Changed in accordance with Tech. Spec 3.2-3 and 3.2-3 to 3.2-4 change #26

TABLE 11-1 (Continued)

CHANGE # TECH SPEC. # PAGE # ACTION EXPLANATION 56 B3/4.2.5 and B3/4 2-4 Replace bases 3/4.2.6 Basis has been added for new Tech. Spec.

B3/4.2.6 with enclosed bases 3/4.2.6 " REACTOR 3.2.6 change #32 COOLANT COLD LEG TEMPERATURE". 3/4.2.7 "AXI AL SHAPE INDEX" and 3/4.2.'l

" PRESSURIZER PRESSURE 57 B3/4.4.1 83/4 4-1 Change DNBR value from "1.30" to Minimum DNBR changed to reflect the change "1.24" in methodology from COSM0/W-3 to TORC /CE-1 58 B3/4.4.2 and B3/4 4-1 Change "to relieve 395,000 lbs per Update of preliminary design value, B3/4.4.3 hour of saturated steam" to "to 395,000 lbs. per hour, to the actual relieve 420,000 lbs. per hour of rated value of 420,000 lbs. per hours.

saturated steam"-

L Sg B3/4.4.6.2 83/4 4-4 Change "The total steam generator tube The basis for Tech. Spec. 3/4.4.6.2 is leakage limit of 1GPM for all steam steam generator integrity during a main generators ensures that the dosage steam line rupture or LOCA events.

contribution from the tube leakage will Compliance with dosage assumptions used be limited to a small fraction of Part for safety analysis is the basis for 100 limits in the event of either a Tech. Spec. 3.7.1.4 change #'s 48, 60 steam generator tube rupture or and 62.

steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.

The 0.5 GPM leakage limit per steam generator . ... " to -

"fhe steam generator tube leakage limit of 0.50 GPM per steam generator. ..

l TABLE 11-1 (Continued)

CHANGE # TECH 5PEC. # PAGE # ACTION EXPLANATION 60 63/4.4.8 B3/4 4-5 Change total primary-to-secondary leakage rate from "1.0 GPM" to "100 100 GPD is the primary-to-secondary GPD" leak rate which is consistent with d secondary activity limit of 0.046nCi/

gram in accordance with Tech Spec change #8s 18, 59 and 62.

61 B3/4.7.1.2 B3/4 7-2 Change "Each emergency feedwater pump 485 gpm and 1100 psig are only two of ,

is capable of delivering a total several input parameters to the safety feedwater flow of 485 opm at a analysis. Recent problems with ANO-2 pressure of 1100 psig to the entrance electric motor driven EFW pump and assoc-of the steam generators. This capacity iated analyses have highlighted the need to is sufficient to ensure . .. " to revise this basis to allow evaluation of "Each emergency feedwater pump is EFP adequacy based on various safety capable of delivering sufficient analysis assumptions.

C feedwater flow to ensure. . .

1 100 GPD is the primary-to-secondary

• 62 B3/4.7.1.4 B3/4 7-3 Change primary-to-secondary SG tube leak rate which is consistent with leakage rate from "1.0 GPM" to "100 a secondary activity limit of 0.046nCi/

GPD" gram in accordance with Tech Spec change #'s 48, 59 and 60 63 B/4.9.6 B3/4 9-2 Change "CEA's and fuel assemblies" to "CEA's with fuel assemblies" and Ir, accordance with Tech. Spec. Change #49 "CEA or fuel assembly" to " fuel assembly" 64 S.6 5-5 Change Tech. Spec. number for Spent Correct error in Tech. Spec. number fuel Cri ticality. from "5.6.1" to ,

"5.6.1.1" and Tech Spec number for New fuel Criticality from "5.6.2" to "S.6.1.2"

i 1

1 1

e INDEX i

DEFINITIONS SECTION PAGE

~

1.0 0EFINITIONS i.

D e f i n e d T e rm s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 Thermal Power.............................................. 1-1 Rated Thermal Power........................................  :-l O p e ra t i o n a l Mod e - Mo d e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 Action.... ................................................ 1 -1 Operable - Operability.................. . . . . . . . . . . .. . .. . . . 1-1 Repo rta bl e Oc cu rrenc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2

! Co n ta i nme n t I n t e g r i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 Channel Calibration........................................ 1-2 Channel Check.............................................. 1-3 O Channel Functional Test....................................

Core Alteration............................................

1-3 1-3 4

Shutdown Margin............................................ 1-3 I d e n t i f i e d L ea k a g a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 Uni d en ti fi ed L ea ka ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 Pressure Boundary Leakage.................................. 1-4 l Azimuthal Power Tilt....................................... 1-4 1

Do s e E qu i va l en t I - 131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4

'E'-Average Disintegration Energy............................ 1-5 l

c S ta g ge re d T e s t Ba s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1 1 1 Frequency Notation......................................... 1-5 Axial Shape Index.......................................... 1-5 Reactor Trip System Response Time.......................... 1-5

Engineered Safety Feature Response Time.................... 1-6

, Physics Tests.............................................. 1-6 1

e Software................................................... 1-6 Pe a k, ngF . s e . . . . . . . . . . . . . . . . . . . . .

, Plana,- %h l gs

.a O

7 ARKANSAS - UNIT 2 I I

V--

---~.m....y. -

v- ,v-

/ INDEX I

LIMITING CON,9ITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 4

j SECTION PAG [

i 3/4.2 POWER DISTRIBUTION LIMITS

,l 3/4.2.1 LINEAR HEAT RATE..................................... 3/4 2-1 3

3/4.2.2 RAD I AL P E AKI NG FACT 0 RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2 23s 4

! 3/4.2.3 AZIMUTHAL POWER TILT................................. 3/4 224s 5

~

3/4.2.4 DNBR MARGIN.......................................... 3/4 2}&s7 3/4.2.5 RCS FLOW RATE........................................ 3/4 2-10 ,

%EAcTcR "* '

3/4.2.6 _ CE ??.': RAGE COOL?.NT^ TEMPERATURE . . . . . . . . . . . . . . . . . . .3/4 . . 2-11 3/y.1.7 M / AL .Th'MC MM Y - .. . -.. . . . . 3lq g -jg, s/4.29

} PR Es cu R s ttR TREssuRG . .. . .... . .

. . , ,3)q g _ , 3 l j 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION . . . . . . . . . . . . . . . . . . . 3/43-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM

. INSTRUMENTATION.................................... 3/4 3-10

'N 3/4.3.3 MONITORING INSTRUMENTATION Radiati on Moni toring Instrumen ta ti on . . . . . . . . . . . . . . . . . 3/4 3-24 Incore Detectors..................................... 3/4 3-23 I Seismic Instrumentation.............................. 3/4 3-30 i

j Meteorologi cal I ns trumen tati on. . . . . . . . . . . . . . . . . . . . . . . 3/4 3-33 l Remote Shutdown Instrumentation...................... 3/4 3-36 ,

I l Post-Accident Ins trumen tati on . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-39

• I Chlorine Detection Systems........................... 3/4 3-42 l

l Fire Detection Instrumentation....................... 3/4 3-43 l

l 3/4.3.4 TURBINE OVERSPEED PR0TECTION......................... 3/4 3-45 l

1 t

ARXANSAS - UNIT 2 IV e

i O INDEX BASES PAGE

'. SECTION

', , ( 3]4.0 APPLICABILITY.......................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTR0L..................................... B 3/41-1 3/4.1.2 B0 RATION SYSTEMS..................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTRCL ASSEM5 LIES........................... B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE..................................... B 3/4 2-1 3/4.2.2 RAb i AL P EA KI NG F ACT0RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-2 M

i 3/4.2.3 AZIMUTHAL POWER TILT................................. B 3/4 2-2 i$

3/4.2.4 onBR MAaGIn.......................................... B 3/4 2-3 q 3/4.2.5 RCS FLOW RATE........................................ B 3/4 2-4 i REACTOR W '#0 3/4.2.6 N "Ji ME C OOL ANT T EM P ERATURE . . . . . . . . . . . . . . . . . . . . . B 3/4 2-4 3h .1.*7 AYlet M PE /ND2X..... ... .. . . . . . .

, 93 gjy g _ q 5/4.1.9 PRE.SSU R 12 ER FREstaQG . _ _ _ ___, , _

_ _ ,g _

3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION........................... B 3/4 3-1 i

1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION . . . . . . . . . . . B 3/4 3-1 j 3/4.3.3 MONITORING INSTRUMENTATION........................... B 3/4 3-1 1

3/4.3.4 TURBINE OVERSPEED PROTECTION......................... B 3/4 3-3 4

1

.k ARKANSAS - UNIT 2 IX

s-DEFINITIONS .

i 1

l ENGINEERED SAFETY FEATURE RESPONSE TIME 1.24 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that . time

, j interval from when the monitored oarameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equi;: ment is capable of 1 performing its safety function (i.e. , the valves travel to their .

required positions, pump discharge pressures reach their recuired values,etc.). Times shall include diesel generator starting and

. 'I sequence 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 j 1.26 The digital computer SOFTWARE for the reactor protection system shall be the program codes including their associated data, documentation and procedures. ,

PLANAR RADIAL PEAKING FACTOR - Fxy I- ,

l 1.27 The PLANAR RADIAL PEAKING FACTOR is the ratio of the ceak to plane average power density of the individual fuel rods in a given horizontal .

plane, excluding the effects of azimuthal tilt. '

. Y 1

ADD .  :

t, .

t 9

i l*

s ARKANSAS - UNIT 2 1-6

-~ .. .. . -. .

h __ _

l l

l

'o i 2.0 SAFETY LIMITS AND LIf1ITING SAFETY SYSTEM SETTINGS l

! 2.1 SAFETY LIMITS 6

, , 2.1.1 REACTOR CORE DNBR i 1.1V

'l 2.1.1.1 The DNBR of the reactor core shall be maintain:d .aE

, s APPLICABILITY: MODES 1 and 2.

1 ACTION:

,l I.2//

Whenever the DNBR of the reactor core has decreased to less thanM be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

1 PEAK LINEAR HEAT RATE

, [.

h 2.1.1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of

. the fuel shall be maintained < 21.0 kw/ft, d APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21.0.kw/ft, be in HOT STANDBY within I hour.

i i,

e

..s ARVANSAS - UNIT 2 2-1

M TABLE 2.2-1 5

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS g FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

] 1. Manual Reactor Trip Not Applicable Not Applicable

2. Linear Power Level - liigh 3 ))o% 4110.712 %

a. Four Reactor Coolant Pumps W of RATED TilERMAL POWER t--H%742.L of RATED TilERMAL POWER Operating

b. Three Reactor Coolant Pumps *

• Ope ra ting

c. Two Reactor Coolant Pumps *

• Operating - Same Loop m d. Two Reactor Coolant Pumps *

• b Operating - Opposite Loops 3 Logarithmic Power Level -

liigh (1) < 0.75% of RATED THERMAL POWER < 0.819% of RATED TilERMAL POWER 13 c2. 2370.197

4. Pressurizer Pressure - liigh < M psia ~

< 235hB3L psia 174& 1712.757

[ 5. Pressurizer Pressure - Low > 1740. psia (2) >N%J5 psia (2) f

6. Containment Pressure - liigh < 18.4 psia 3 < 19.024 psia 75/ 719 4/3

[ 7. Steam Generator Pressure - Low > 72tt psia (3) ~

o > 70fdpsia (3)

'f 'IS. 911 h 8. Steam Generator Level - Low >_'46.5%-(4)

G.7. >15413 (4)

T These values lef t blank pending NRC approval ofganalyses for operation with less than four reactor , coolant pumps operating. S4 My

__ . c R_..m _ _

_ ._m m- _

mm___.__maia h W'm h m. L

-.a O O O l TABLE 2.2-1 (Continued) 1 g $" REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS

5 a g

! $; FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES b 9. Local Power Density - liigh 1 20.3 kw/ft (5) 1 20.3 kw/ft (5)

. 5 f.2V J.:W

10. DNBR - Low > L ( 5 ) > L-3"(5)

]"" _

33.7 9e/.589 l 11. Steam Generator Level - !!igh S M (4) 174Aa9% (4) d TABLE NOTATION i

-4 (1) Trip may be manually bypassed above 10 ~4

% of RATED TilERMAL POWER; bypass shall be automatically a removed when THERMAL POWER is < 10 % of RATED TilERMAL POWER. ~

~ dariqaylened ceJuchan on prenueiner

'n (2) Value may be decreased manually, to a minimum value of 100 psia, n y.uurizer prc33urc is pressucc reduccd, provided the margin between the pressurizer pressure and this value is maintained at

.] 1 200 psi; the setpoint shall be increased automatically as pressurizer pressure is increased

, until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall i

be automatically removed whenever pressurizer pressure is 1 500 psia.

Aur; y a p%,,ul reheWn in sleam p0.4.< pressure (3) Value may be decreased manually-as ste generatc?prc=.rc is rci.ced, provided the margin between the steam generator pressure and this value is maintained at 1 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

(4) % of the distance between steam generator upper and lower level instrument nozzles.

(5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement,calculationalgndprocessoruncertainties,anddynamicallowances. Trip may be I 10- % of RATED TilERMAL POWER; bypass shall be automatically removed when manually bypassed belog% of RATED THERMAL POWER.

TilERMAL POWER is 1 10-I

2.1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

/

I BASES l

l i 1

2.1.1 REACTOR CORE  !

f The restrictions of these safety limits prevent overheating of the

. fuel cladding and possible cladding perforation which would result in I l the release of fission products to the reactor coolant. Overheating of E

the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient i i is large and the cladding surface temperature is slightly above the  !

i coolant saturation temperature, and (2) maintaining the dynamically t adjusted peak linear heat rate of the fuel at or less than 21 kw/ft l l which will not cause fuel centerline melting in any fuel rod. l First, by operating within the nucleate boiling regime of heat

)

transfer, the heat transfer coefficient is large enough so that the '

maximum clad surface temperature is only slightly greater than the  !

coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB). At

this point, there is a sharp reduction of the heat transfer coefficient,

which would result in higher cladding temperatures and the possibility i of cladding failure.

Correlations predict DNS and the location of DNB for axially i i uniform and non-uniform heat flux distributions. The local DNB ratio 4

(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 l indicative of the margin to DNB. The minimum value of DNBR during r normal operational occurrences is limited to L-Pfor the A correlation l and is established as a Safety Limit. l.28 CE- / l Second, operation with t peak linear heat rate below that which j would cause fuel centerline melting maintains fuel rod and cladding [

g integri ty. Above this peak linear heat rate level (i.e. , with some  !

I melting in the center), fuel rod integrity would be maintained only if  !

i the design and operating conditions are appropriate throughout the life  !

of the fuel rods. Volume changes wnich accompany the so. lid to liquid

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

• l peak linear heat rate which would not cause fuel centerline melting is

, established as a Safety Limit. To account for fuel rod dynamics (lag:),  :

l the directly indicated linear heat rate is dynamically adjusted.

Limiting safety system settings for the Low DNBR, High Local Power

( [. Density, High Lcgarithmic Power Level, Low Pressurizer Pressure and High v

ARKANSAS - UNIT 2 B 2-1

. _ - - . _ - - . , - - .,, _, ._-._..,.._,.y_.. _ . _ - . _ . _ ,m... .

SAFETY LIMITS AND LIMITING SAFETY SY! TEM SETTINGS

! BASES

' Linear Power Level trips, and limiting conditions for operation on DNBR and j;

kw/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during 7 normal operation and design basis anticipated operational occurrences.

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

i The Reactor Coolan; System canponents are designed to Section III of the ASME Code for Nuclear Power Plant Components. (The reactor j vessel, steam generators and oressurizer are designed to the 1968 Edition, Summer 1970 Addenda; piping to the 1971 Edition, original issue; and the valves to the 1968 Edition, Winter 1970 Addenda. 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

[l the design criteria and associated code requirements. .

.+

ihe entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the 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 Featuras Actuation System in mitigating the consequences 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 equal to or less than the drift alTowance assumed for each trip in the safety analyses.

1.2Y TheDNBR-LowandLocalPowerDensity-Higharedigitaliz/ generated I

trip setpoints based on Limiting Safety System Settings of le'9 and 20.3

, kw/ft, respectively. Since these trips are digitally generated by the j Core Protection Calculators, the trip values are not subject to drifts 1 common to trips generated by analog type equipment. The Allowable a Values for these trips are therefore the same as the Trip Setpoints.

1

! ARKANSAS - UNIT 2 B 2-2 i

y ..--..m -

7 .,- .

i

, SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 1

Steam Generator Level-Low I The Steam Generator Level-Low trip provides protection against a h") loss of feedwater flow incident and assures that the design p essure of i' 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 4 time of the trip to provide 2 ' ic:%1: mi.4es before c SN f 'IIc l* " f ##$ l"

~

] emergency feedwater is required. -W

. k j Local Power Densitv-Hich 1

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 j fuel design limit in the event of any anticipated ocerational occurrence.

-Q The Incal power density is calculated in the reactor protective system j utilizing the following information:

(

a. Nuclear flux power and axial power distribution from the 1 excore flux monitoring system; 4

q b. Radial peaking factors from the position measurement for the CEAs; i c. aT power from reactor coolant temperatures and coolant flow measurements.

The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines. These I uncertainties and dynamic compensation routines ensure that a reactor j 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 DNSR j calculation. Dynamic compensation for peak LPD is provided for the 1 effects of core fuel centerline temperature delays (relative to changes

! in power density), sensor time delays, and protection system equipment time delays.

1 3

,1

O 1 ARKANSAS - UNIT 2 B 2-5 d

i v .- . _ , . _ , . - , . _ - ,

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES i

I

. DNBR-Low The DNBR - Low trip is provided to prevent the DNBR in the limiting i coolant channel in the core from exceeding the fuel design liinit in the I event of anticipated operational occurrences. The DNBR - Low trip incor-porates 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 Ccolant System cressure from pressurizer pressura measurement; I c. Differential temperature (AT) power from reactor coolaat i temperature and coolant flow measurements;

- d. Radial peaking factors from the position measurement for the )

l7; CEAs; v j e. Reactor coolant mass flow rate from reactor coolant pump speed;

!d f. Core inlet temperature from reactor coolant cold leg temperature measurements.

The DNBR, the trip variable, calculated ' y 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 trio occurs when g,1q the actual core DNBR is sufficiently greater than Je$'Such that the decrease in actual core DNBR after the trip will not result in a viola- ,

tion of the DNBR Safety Limit. CPC uncertainties related to DNBR cover i CPC input measurement uncertainties, algorithm modelling uncertainties,

( . and computer equipment processing uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport

! delays, core heat flux delays (relative to changes in core power), sensor ,

i time delays, and protection system equipment time delays.

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

i r

ARKANSAS - UNIT 2 B 2-6 4

! '\

. ,, /

REACTIVITY CONTROL SYSTEMS SHUTOOWN MARGIN - T avg < 200*F i

LIMITING CONDITION FOR OPERATION 50%

3.1.1.2 The SHUTDOWN MARGIN shall be > M ak/k.

I APPLICABILITY: MODE 5.

ACTION:

5.0%

With the SHUTDOWN MARGIN < 'bWQi ak/k. immediately initiate and continue boration at > 40 gpm of 1731 ppm toric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.

l SURVEILLANCE REOUIREMENTS l

,L-s.0%

4 .1.1. 2 The SHUTDOWN MARGIN shall be determined to be >'P?S4 ak/k:

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

If the inoperable CEA is immovable or untripcable, the above required SHUTDOWN MARGIN shall be increased by an amcunt at least equal to the witndrawn worth of the immovable or untrip-

• pable CEA(s).

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

i 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. Samarium concentration.

i l ,

I

,-~

l ARKANSAS - UNIT 2 3/4 1-3

~ , -

r /

l REACTIVITY CONTROL SYSTEMS

MODERATOR TEMPERATURE COEFFICIENT .

I LIMIT!?:G CONDITION FOR OPERATION x r 3.1.1.4 The moderator temperature coefficient (MTC) shall be:

a. Less positive than 0.5 x 10-4 ak/k/*F, and

, b. Less negative than -3.5 x 10-4 ak/k/'FothtATEDTHERMAL i POWER.

REPLnc! WITH APPLICABILITY: MODES 1 and 2*. enc 4esED /NSERT ACTION:

With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

i l SURVEILLANCE REOUIREMENTS 8

f 4.1.1.4.1 The MTC shal'1 be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated

and/or compensated to permit direct ccmparison with the above limits.

4.1.1.4.2 The MTC shall be determined at the following frequencies and THEPJiAL POWER conditions during each fuel cycle: '

a. Prior to initial operation above 5% of RATED THERMAL POWER, l af ter each fuel loading.

2 q l I b. At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium baron concentration of 800 ppm.

+

c. At any THERMAL POWER, within 7 EFPD af ter reach.ing a RATED .

THERMA'. POWER equilibrium boron concentration of 300 ppm. "

• With Ke f f.> 1.0.

i O  % .-

• ARKANSAS - UNIT 2 3/4 1-5

( I l

l REPLACEMENT IflSERT

a. Less positive then 0.5x10 ak/k/ F whenever  ;

THERMAL POWER IS <70% of RATED THERMAL POWER,

b. Less positive then 0.0a k/k/*F whenever i THERMAL POWER is >70% of RATED THERMAL POWER, and ;

-4

c. Less negative then -2.8x10 ak/k/ F at RATED THERMAL POWER  ;

(

i 1

REACTIVITY CONTROL SYSTEMS O

i FLOW PATHS - OPERATING

]

1 LIMITING CONDITION FOR OPERATION 4

4 3 .1. 2 .2 At least two of the following three boron injection flow paths I 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 l . pump to the Reactor Coolant System.

I i APPLICABILITY: MODES 1, 2, 3 and 4.

, ACTION:

With only one of the above required boron injection flow paths to the

^

A Reactor Coolant System OPERABLE, restore at least two boron injection r- flow paths to the Reactor Coolar. System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTCOWN MARGIN equivalent to at least X .k/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths [to OPERABLE status within the next 7 days or be in COLD SHUTDOWN withinithe next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

5%

q SURVEILLANCE REQUIREMENTS 1

4 .1. 2 . 2 At least two of the above required flow ;:aths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the 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 verif fing 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.

l c. At least once per 18 months during shutdown by verifying that

< each actuated valve in the flow path actuates to its correct position on a SIAS test signal. -

/

ij 3

ARKANSAS - UNIT 2 3/4 1-8 l

' - ~ ~ ~

= - - m-

i /

. REACTIVITY CONTROL SYSTEMS

CHARGING PUMPS - OPERATING d

LIMITING CONDITION FOR OPERATION

3.1. 2 . 4 At least two charging pumps shall be OPERABLE.

1 4

APPLICABILITY: MODES 1, 2, 3 and 4.

l ACTION:

i With only one charging pump OPERABLE, restore at least two charging

' pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY sy %

and borated to a SHUTDOWN MARGIN equivalent to at least % k/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.  :

1

I 4

4 l

1 4

SURVEILLANCE REQUIREMENTS i.

4.1.2.4 No additional Surveillance Requirements other than those required {

by Speci fication 4.0.5.

i 1 .

i (

I I i I

.% I i

I ARKANSAS - UNIT 2 3/4 1-10  ;

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

I e

)

REACTIVITY CONTROL SYSTEMS ...

BORIC ACID MAKEUP PUMPS - OPERATING l

4 LIMITING CONDITION FOR OPERATION i 3.1.2.6 At least the boric acid makeup pump (s) in the boron injection flow path (s) required OPERABLE pursuant to Specification 3.1.2.2a shall  !

I be OPERABLE and capable of being powered from an OPERABLE emergency bus i if the flow path through the boric acid makeup pump (s) in Specification '

3.1.2.2a is OPERABLE.

i

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

ACTION:

With one boric acid makeup pump required for the boron injection flow path (s) pursuant to Scecification 3.1.2.2a inoperable, restore the boric '

acid makeup pump to OPERABLE status witnin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least  !

HOT STANDBY within the next 6 t.ours and horated to a SHUTDOWN MARGIN I equiv& lent to at least't( ak/k at 200*F; restore the above required boric acid pump (s) to OPERABLE status within the next 7 days or be in m i l- COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. ,

SURVEILLANCE RECUIREMENTS

i

4.1.2.6 No additional Surveillance Requirements other than those required  !

by Specification 4.0.5. j 4  !

l I l  !  ;

i  !

1 l i

i 7

ARKANSAS - UNIT 2 3/4 1-12 4

,, -_p ,,-,n.-

lO" REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION

.; 3.1.2.8 Each of the following borated water sources shall be OPERABLE:

i

a. At least one boric acid makeup tank 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:

I 1. A contained borated water volume 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 ppm of boron, 1

3. A minimum solution temperature of 40 F, and I '

4. A maximum solution temperature of 100 F.

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

0 .;

ACTION:

1 1

a. With the above required boric acid makeup tank inoperable, j restore the make up tank to OPEP.ABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at least K ak/k at 200'F; restore the above required boric acid makeup tank to CPERABLE

! status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. 5%

1

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN wit'hin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

] SURVEILLANCE REQUIREMENTS

}

4.1.2.8 Each of the above required borated water sources shall be

.1 demonstrated OPERABLE:

ARKANSAS - UNIT 2 3/4 1-15 i

1 - . .- n , , ; - --w _ --

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4

3 .

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA POSITION

[ LIMITING CONDITION FOR OPERATION N

"- 3.1.3.1 All full length (shutdcwn and regulating) CEAs, and all part length CEAs which are inserted in the core, shall be OPEPABLE with each CEA of a given group positioned within 7 inches (indicated position) of all other CEAs in its group.

MAKE j i

Men ,,C , '

APPLICABILITY: MODES 1* and 2*.

r r> n n s MDE R ACTION:

Lc0

a. With one or more full length CEAs inoperable due to being im-s movable as a result of excessive friction or mechanical inter-ference or known to be untrippable, determine that the SHUTDOWN i

\ MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

t

b. With one full length CEA inoperable due to causes other than addressed by ACTION a, abcVe, and inserted beyond the Lcng

'? Term Steady State Insertion Limits but within its above specified alignment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification i 3.1.3.6.

t NOTE:

\. WithonefulllengthCEAinoogableduetocausesotherthan addressed by ACTION a, cif, M within its atws.soecified ,

alignment requirements and either fully withdrawn or within  !

the Long Term Steady State Insertion Limits if in full length !

CEA group 6, operation in MODES 1 and 2 may continue.

C% With one or more full length or part length CEAs misaligned from any other CEAs in its group by more than 7 inches but less than or equal to 19 inches, operation in-MODES 1 and 2 may continue, provided that within one hour the misaligned CEA(s) is either:

1. Restored to OPERABLE status within its above specified alignment requirements, or

! l l J *

?, See Special Test Exceptions 3.10.2 and 3.10.4.

1 ARKANSAS - UNIT 2 3/4 1-17 I

y -

i REACTIVITY CONTROL SYSTEMS ACTION: (Continued) t I

2. Declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. After declaring

the CEA inoperable, operation in MODES 1 and 2 may continue

( pursuant to the requirements of Specification 3.1.3.6 l provided:

i a) Within one hour the remainder of the CEAs in the i group with the inoperable CEA shall be aligned to I within 7 inches of the inoperable CEA while main-taining the allowable CEA sequence and insertion i limits shown on Figure 3.1-2; the THETJ4AL POWER level shall be restricted pursuant to Specification

3.1.3.6 during subsequent operation.

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

Otherwise, be in at least HOT STAN0BY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ,,

With one full length or part length CEA misaligned from any s>

d% other CEA in its group by more than 19 inches, operation in MODES 1 and 2 may continue, provided that within one hour the misaligned CEA is either:

1

1. Restored to OPERABLE status within its above specified alignment requirements, or

2. Declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. After declaring the CEA inoperable, operation in MODES 1 and 2 may 1 continue pursuant to the requirements of Specification 3.1.3.6 provided:

i a) Within one hour the remainder of the CEAs in the group with the inoperable CEA shall be aligned to I within 7 inches of the inoperable CEA while main-taining the allowable CEA sequence and insertion limits shown on Figure 3.1-2; the THEPfiAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least one per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Othentise, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ARKANSAS - UNIT 2 3/4 1-18 l t

---

w, . , - . , - - , - -

y I REACTIVITY ;0NTROL SYSTEMS ACTION: (Continued) a i

e %. With one part length CEA inoperable and inserted in the core,

, operation may continue provided the alignment of the inoperable J PLCEA is maintained within 7 inches (indicated position) of all 1

other PLCEAs in its group.

f1 With more than one full length or part length CEA inoperable or misaligned from any other CEA in its group by more than 19 I

inches (indicated position), be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS t ,

4.1.3.1.1 The position of each full length and part length CEA shall be determined to be within 7 inches (indicated position) of all other CEAs in its group at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when one CEAC is inoperable or when both CEACs are inoperable, then j verify the individual CEA posit Mns at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.1.2 Each full length CEA not fully inserted and each part length a CEA hhich is inserted in the core shall be determined to be OPERABLE by movement of at least 5 inches in any one direction at least once per 31 days.

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0 3/4.2 POWER DISTRIBUTION LI? TITS LINEAR HEAT RATE

.-. LIMITING CONDITION FOR OPERATION t

In be ma1nIalned by p cP'#*419 winin 3.2.l' The linear heat rate shall "' evceed th

imits sh a.. .; g ,_

W+he reqipn of recepta ble opera &ien of Ryaees 3.2-1 or 3.1-2 as affh ca ble.

APPLICABILIiY: MODE 1 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 en kw/ft; or (2) when the COLSS is not being used, any 0PEPASLE Local Power Density channel exceeding tne linear heat rate limit, within 15 minutes initiate corrective action to recuce the linear heat rate to within the limits and either:

a. Restore the linear heat rate to within its limits within one

( hour, or j b. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

1

! SURVEILLANCE RE0UIREMENTS i

H 4.2.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.1.2 The linear heat rate shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by con-tinuously monitoring the core power distribution with the Core Operating Limit Supervisory System (CCLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the linear heat rate, as indi-cated on all OPERABLE Local Power Density channels, is within the limit shown on Figure 3.2-1.

4.2.1.3 At least once per 31 days, the COLSS Margin Alar.n shall be

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

1

?O I Y I ARKANSAS - UNIT 2 3/4 2-1

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POWER DISTRIBUTION LIMITS t

RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION PLANAR khDIRL ifAklNG FAC.TCRS F**1

', 3.2.2 The measured _rlaa'" "2 dbl p= Un; fa a ,2 (r* shall be less

. PL =n'- "' . PEANINGpc&ngu than or equal to tne n9NAR RRD,lk._

RAcTORS ws (, ) used in the Core H OperatingLimitSupervisorySystem(COLSS)andinthe{CoreProtection i

Calculators (CPC). pc YY APPLICABILITY: MODE 1 above 20*I, of RATED THERMAL POWER.*

ACTI0ft : 7' With a exceeding a corresponding , within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

a. Adjust the CPC addressable constants to increase the multiplier applied to al'a'- -?d'al exi.x bh'cRa factor equivalent to m PLANAR RADIAL PEhWING Spc Fy /F,Gy tM anc restrict subsequent operation so that a margin to the COLSS operating limits of at least [(

. is maintained; or hyy-J.0],)x1001, l_  : b.

PLANAR RnDIAL PEAK)NG FACTCRS Adjust the affected nr'- 2d'21 p x king fe;tm 2

,4 c xyus{[c d in a

the COLSS and CPC to a value greater than or equal t she measured nian'- 2+ 21 p xmin; fe a a %; or q

1 PLANAR RADIAL PEAklNG FACTORS

, c. Be in at least HOT STANDBY. (

i SURVEILANCE RECUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

PLMAR R4 DIAL PEAKWG Fac7cRS 4.2.2.2 The measured _n1'n'- -adial pc i n; fcc m 2  %,f"' #Ubtained by using the incore detection system, shall,be determined to be less than or equal to the n1'a'- -edial p= kin- fa a 3 c) used in the COLS$ and Pl.AHAR RRDIAL. PEAVIIV4 FACTDRS c CPC at the following intervals: F XY

a. After each fuel loading with THEPJ4AL POWER greater than 40S

but prior to operation above 70% of RATED THERMAL POWER, and

b. At least once per 3'i days of accumulated operation in MODE 1.

• 5ee special Test Exception 3.10.2.

.j ARKANSAS - UNIT 2 3/4 2 l l 1

l A- - ~ v- , .r w - . - , - - _ _ _ ,- _ ,

I

1 POWER DfSTRIBUTION LlMITS j

s AZIMUTHAL POWER TILT - Tq I,

i LIMITING CONDITION FOR OPERATION l

N 1 j 3.2.3!TheAZIMUTHALPOWERTILT(T)shallbelessthanorequalto W 9 1 the A%IMUTHAL POWER TILT Allowance used in the Core Protection Cal-  !

culators (CPCs).

1  ;

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

ACTION:

a. With the measured AZIMUTHAL POWER TILT determined to exceed the AZIMUTHAL POWER TILT Allowance used in the CPCs but 1 0.10, within two hours either correct the power tilt or  !

adjust the AZIMUTHAL POWER TILT Allowance used in the CPCs  :

to greater than or equal to the measured value. l

b. With the measured AZIMUTHAL POWER TILT detennined to exceed i f 0.10: '

l. Due to misalignment of either a part length or full 1

f length CEA, within 30 minutes verify that the Core s '

Operating Limit Supervisory System (COLSS) (when COLSS is being used to monitor the core pcwer distribution per Specifications 4.2.1 and 4.2.4) is detecting the CEA -

~J misalignment. ,

2. Verify that the AZIMUTHAL POWER TILT is within its limit '

within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the

' next 2 hcurs and reduce the Linear Power Level - High l Orip setpoints to 5,55% of RATED THEPJ1AL POWER within the J

$ next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

t Identify and correct the cause of the out of limit con-

3. i dition prior to increasing THEPMAL POWER; subsequent POWER

(

OPERATION above 505 of RATED THERMAL POWER may proceed provided that the AZIMUTHAL POWER TILT is verified within -

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

• l See Special Test Exception 3.10.2.

f' ARKANSAS - UNIT 2 3/42-\ I i

,w .

,w-9

,- - - - - - - , , . - , . r, . ,,__, -,~y,w, , _ , . --r, - -,- . -,,-.- -- - ,,- , -, - ... , - , - . , -

,_--r-.

i i

POWER DISTRIBUTION LIMITS

{ f SURVEILLANCE REQUIREMENTS 4.2.3 The AZIMUTHAL POWER TILT shall be determined to be within the

.' limit above 20% of RATED THERMAL POWER by:

1 g a. Continuously monitoring the tilt witn COLSS when the COLSS is OPERABLE.

b. Calculating the tilt at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the COLSS is inoperable.

A

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 CFCs.

1

d. Using the incore detectors at least once per 31 days ;o independently confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT.

s

!, -o 1

'b I

r-l 4

4 1

l _

l l 1 4

1 L

M (d?h b/ 6 4 ARKANSAS - UNIT 2 3/42\g a

j y r. - . _ - .

. ~ . . . - , -

. = .. .- , ..

P

i \

,j POWER DISTRIBUTION LIrilTS DNBR MARGIN i

LIMITING CONDITION FOR OPERATION 3.2.4 The DNf5R mar ('n shall be maintained b operatin within the region of acceptable operation of Figure 3.2- or 3.2 , as applicable.

! 3 9 APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.

l ACTION:

With operation outside of the region of acceptable operation, as 3

indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on DNBR; or (2) when the COLSS is not being used, 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 STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

x./

. SURVEILLANCE REQUIREMENTS 4

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 Supervisory System (COLSS) or, with the COLSS out of service, by verify-ing at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the DNBR, as indicated on all OPERABLE DNBR channeis, is within the limit shown on Figure 3.2-3.

i 4.2.4.3 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 DNBR.

't 7 .-

ARKANSAS - UNIT 2 3/42-%

l

_- .- __:.__ x 3.. _ - - . .. - -- . ._ . _ .-

i J

P R DISTRIBUTION LItiITS SURVEILL 'lCE REQUIREMENTS (Continued) -

x -

4.2.4.4 The fo owing DNBR penalty factors shall be rified to be

/

included in the C SS and CPC DNBR calculations a east once per 31

,j, days: ,

BurnuoiGWD) U NBR Penalty (%)

0-3.1 0 i 3.1-5 , 2.0 l 5-10 5.9 10-1 8.8

-20 11.4 20-25 13.6 25-30 1 6 7 30-35 17.4 j

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FOR CORE AVEf t AGE UtlitNUP ll q

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(2.36,0 4)

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OF >6000 MWD /MTU 4 A th w 2 '-# i s s 2.21,0.6)

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g AXI AL Sif APE INDEX y Figure 3.2-3 DNBR Margin Operating Limit Based on Core Protection Calculators (COLSS Out of Service) i

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APJ@SAS - UNIT 2 3/42-9

-i

)

J

l Ox Q OWER DISTRIBUTION LIMITS C AVERAGE COOLANT TEMPERATURE 9

'l LIMITING NDITION FOR OPERATION x _

e

'$t If 3.2.6 The core a- age coolant temperature avg) shall be < 588.2 F.

APPLICABILITY: MODE 1 d ACTION:

S With the core average coolant t erature exceeding its limit, restore y the temperature to within its im.* within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL i POWER to less than 5% of RA7 D THERc. L POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

i

-J .

j i

P- .

SURVEILLANCE EOUIREMENTS e x b

] i 4.2. The core average coolant temperature shall be deter..'ned to be wi+ in its limit at least once per 12 hourr.

4 1

/

/

REPLAG WITH ENCLOSE D TECH. SPEC.

3 . 2 . (,

[fg3/9A-//)

l , 3.2.7 [pg 3/4 2- 12 ) -

l 3.R.S g 3/4 2-/3 )

\

l l u l1 15 l1 a O 1 '

ARKANSAS - UNIT 2 3/4 2-11 3

7e., -- - - . . . . - _

._,m-- - . _ . --

T_

REPLMdMENT "h. "3 X f, I

POWER DISTRIBUTION LIMITS REACT". COOLANT COLD LEG TEMPERATURE LIMITING CONDITION FOR OPERATION 3.2.6 The Reactor Coolant Cold Leg Temperature (Tc) shall be maintained between 542 F and 554.7"F.

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

With the Reactor coolant cold leg temperature exceeding its limit, restore the temperature to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 30% of RATED THEPJ4AL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

i SURVEILLANCE REQUIREPENTS 4.2.6 The Reactor coolant cold leg temperature shall be determined to be

.within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

d i

ARKANSAS - UNIT 2 3/4 2-11

NtM%ME/V7 TEs. 3.2.p POWER DISTRIBUTION LIMITS AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The core average AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:

a) COLSS OPERABLE

-0.28 < ASI < + 0.28 b) COLSS OUT OF SERVICE (CPC)

-0.20 < ASI < +0.20 APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 20% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.6 The core average AXIAL SHAPE INDEX shall be determined to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> using the COLSS or any OPERABLE Core Protection Calculator channel .

ARKANSAS - UNIT 2 3/4 2-12

MPMkfhtfMT TS. 3. 2. 4 l

i POWER DISTRIBUTION LIMITS PRESSURIZER PRESSURE LIMITING CONDITION FOR OPERATION 3.2.8 The average pressurizer pressure shall be maintained between 2225 psia and 2275 psia.

APPLICABILITY: MODE 1 ACTION:

With the average pressurizer pressure exceeding its limits, restore the temperature to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERFAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.6 The average pressurizer pressure shall be detennined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ARKANSAS - UNIT 2 3/4 2-13

, ,--,-.,:--,-e..- ,m- '-~ ,. - - , - - - - - - - - - - -

TABLE 3.3-1 E?

g REACTOR PROTECTIVE INSTRUMENTATION MINIMUM i h TOTAL NO. CHANNELS CHANNELS APPLICABLE

a. TO TRIP OPERABLE MODES ACTION z FUNCTIONAL UNIT OF CHANNELS N 1. Manual Reactor Trip 2 sets of 2 1 set of 2 2 sets of 2 1, 2 and

• 1

2. Linear Power Level - High 4 2 3(g) 1, 2 2#

3. Logarithmic Power Level-High {pgccrd

a. Startup and Operating 4 2(a)(d) 3 (p ) Y,2and* 2#

4 0 2 3, 4, 5 3

b. Shutdown Pressurizer Pressure - High 1, 2 2#

4. 4 2 3 (3) 4 2(b) 3 (9) 1, 2 and A' 2#

2 5. Pressurizer Pressure - Low N

6. Containment Pressure - High 4 2 3 (S) 1, 2 [ ADD] 2#

[ /

2/SG 3/SG (3) 1, 2 aml

• 2#

7. Steam Generator Pressure - Low 4/SG 1, 2 2=

8. Steam Generator Level - Low 4/SG 2/SG 3/SG(3)

9. Local Power Density - High 4 2(c)(d) 3(g) 1, 2 2#

1_;

i e e

! \

l 1

j -

TABLE 3.3-1 (Continued)

! E j g REACTOR PR0fECTIVE INSTRUMENTATION i Si i

i a e

I E MINIMUM TOTAL N0. CHANflELS CHANT 1ELS APPLICABLE

! Z FUNCTI0tJAL UiiIT OF CHATJt1ELS TO TRIP

~ OPERABLE MODES ACTION

{ 10. Of1SR - Lovi 4

2(c)(d) 3(9) 1, 2 2#

11. Steam Generator Level - High 4/SG 2/SG 3/SG(g) 1, 2 2#

12. Reactor Protection System Logic 4 2 4 1, 2 and

• 4

13. Reactor Trip Breakers 4(f) 2 4 1, 2 and

• 4 5

14. Core Protection Calculators 4 2(c)(d) 3 (3) 1, 2 2# and 6 m

2 15. CEA Calculators 2 1 2(e) 1, 2

\and6 Y

m 5#

l 1

l i

i

r

. hJ TABLE 3.3-1 (Continued)

TABLE NOTATION j

~

With the protective system trip breakers in the closed position and y the CEA drive system capable of CEA withdrawal. j l

The provisions of Specification 3.0.4 are not applicable. l

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

bypass shall be automatical;y removed when THERMAL POWER is < 10"% _

1 I

of RATED THERMAL POWER.

i) ' (b) Trip may be manually bypassed below 400 psia; bypass shall be

' automatically removed whenever pressurizer pressure is > 500 psia.

-4 (c) Trip may be manually bypassed below 10 % of RATED THERMAL POWER;4 bypass shcIl be automatically removad when THERMAL POWER is > 10- %

3 of RATED THERMAL POWER. During testing pursuant to Special Test li 7 Exception 3.10.3, trip may be manually bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL

!9 j e POWER is > 1% of RATED THERMAL POWER.

i (d) Trip may be bypassed during testing pursuant to Special Test Excep-9 tion 3.10.3. f,

~k (e) See Special Test Exception 3.10.2. g

- (f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice. 1

~

ACTION STATEMENTS

(

I ACTION 1 -

With the number of channels OPERABLE one less than t

required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within

' 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the protective system trip breakers.

I taill be main fained

• (9) doms .m///GPERABLE l car Cb2snel.5 conoWiert, Howerer, oe ist 1st inoperable channel may be bypassed for up to 90 days from the time of initial loss of OPEPABILITY if the inoperable component is located outside of t containmente nu mmen,The inoperable channel shall then be either restroed to OPEPABLE status or placed in the tripped condition. 7/jg ,

, s M inoperable channel may be bypassed until the next refueling if the inoperable component is '

located inside of containment. - '*; .'

3/43-9  % '-

h ^.~"i M ''.'?W  !

r

(-

I 4

TABLE 3.3-1 (Continued) l ACTION STATEfiEtlTS l

/ l ACTIOil 2 -

With the number of OPERABLE channels one less than the Ab/u Cfwe/3 "' ' "-

~ ^b d STARTUP and/or POWER OPERATION OPEt/BLE may procepd provided the following conditions are satisfied: l Oae of d e wo

a. -lhe inoperable channel

• 3 is placed in Jiipasr the d:'p!@2: Essa tripped condition within I hour A BiluJ

+

^ 7" . :n M m -..% m v44 * "=, -t* c + 4se-- -

C& =.eeks.4&p;; =:n- f c T:! TM ^ A ^M

• % f:kM kM==:2. :;2C-2% +_w- -&>

swer.ents': ==:d :2642& 2MLer =.tg

C2.r~. 'mNQ h &2ue.ye;4 caennWPeg 8 Withinonehour,allfunctgallogicunitsreceiving an input from -

_ _ . _3 d'hannel are also placed g' ,

t in the SeMC"iac444wrc,ceauP

=^^^2 ==:  % & s = M +2 m r: tripped 1m4:cou//!/o MadeMnHeewermsw I ~

~

~

m, a 08Sf*4Blafrwewww.t _i; s ._ _ ,

F- - ,_

f_ u m 7- -

- s ,a - o s -_ _

t  ;

__ . =. -_ _ _

-r - m--. -- - ; _ _ _

. , _ , . , . . __ 3 n 4-xxy

$2 M_Y :;r K ' 2 % -L,-q d W & r ' __

}  %, 'i ^ ' _ % m QJ rd R _

ACTION 3 -

With the number of channels OPERABLE one less than required

by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 I

hour and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereaf ter.

l l

ACTION 4 -

With the number of channels OPERABLE one less than required e

by the Minimum Channels OPERABLE requirement, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to I hour for surveillance testing per Specification 4.3.1.1.1.

ACTION 5 -

a. With one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, i

each CEA is verified to be within 7 inches (indicated position) of all other CEAs in its group.

l i t l 1 l

1 ARKANSAS - UNIT 2 3/4 3-5 h

l (t 1

i

. . - - - - . , , - - - - - - - - - - - - - - - - - ~ ~ ~ ~ ~ ~ ~ ~ - - ~ ~ ' ' ~ ' " '

j, e TABLE 3.3-1 (Continued)

ACTION STATEMENTS

b. With both CEACs inoperable, operation may continue provided that:

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the margins required by Specifi-2-

cations 3.2.1 and 3.2.4 are increased and main-

,j tained at a value equivalent to % of RATED 1;j THERMAL POWER. 2117, b .

2. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a) All full length and part length CEA groups are withdrawn to and subsequently main-tained at the " Full Out" position, ucept 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 Inoperable" addressable constant in the CPCs is set to the :.ioper-

_ able status, i

j c) The Control Element Drive Mechanism Control jj System (CEDMCS) is placed in and subsequently j maintained in the "Off" mode except during CEA group 6 motion permitted by a) above, 4 when the CEDitCS may be operated in either j the " Manual Group" or " Manual Individual" a mode.

3. At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, all full length and

! i part length CEAs are verified fully withdrawn except during surveillance testing pursuant to l Specification 4.1.3.1.2 or during insertion of l j CEA group 6 as permitted by 1. a) above, then l verify at least once per a hours that the l

inserted CEAs are aligned within 7 inches (indicated position) of all other CEAs in its 1 group.

l l i ACTION 6 - With three or more auto restarts of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL FUNCTIONAL 1

=4 TEST within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

j O

ARKANSAS - UNIT 2 3/4 3-Sa i  ;

3

- ' = ' ~~-~

't , . - ~ , -:. ,- ~=m.,--

n . , - .- , - - - -

l

} .. . _ . - -. .. ._

(l O O 2

TABLE 4.3-1

{

2 5 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

CllANNEL MODES IN WillCH E CilANNEL FUNCTIONAL SURVEILLANCE CllANNEL Q CilECK__ CALIBRATION _ TEST REQUIRED m FUNCTIONAL UNIT N.A. N.A. S/U(1) N.A.

1. Manual Reactor Trip Linear Power Level - liigh 5 0(2,4),M(3,4), M 1, 2 2.

Q(4)

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

3. Logarithmic Power Level - liigh S and

• R M 1, 2

4. Pressurizer Pressure - High S

{ 1,2 and *

" S R M

5. Pressurizer Pressure - Low ,

4 k R M 1, 2

6. Containment Pressure - liigh S S R M 1, 2 and *

7. Steam Generator Pressure - Low s R H 1, 2 S

8. Steam Generator Level - Low Local Power Density - liigh S P(2,4),R(4,5) M,R(6) 1, 2 9.

9 0 TABLE 3.3-3 ,

I g Ef!GIf1EERED SAFETY FEATURE ACTUATIO:t SYSTEf t IrtSTRUf1EtiTATIO 1 9 r11tiIt1UM g CHANilELS APPLICABLE g TOTAL NO. CHA!iNELS TO TRIP OPERABLE MODES ACTION FUtlCTIO :AL UttIT OF CHAf!NELS_

h 1. SAFETY II JECTI0il(SIAS) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 8 H a. flanual (Trip Buttons)

N l b. Containment Pressure -

High 4 2 3 (d) 1, 2, 3 9*

c. Pressurizer Pressure - 1,2,3(a)

Low 4 2 3 (d) 9*

2. CONTAII; MENT SPRAY (CSAS)

a. l1anual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 8 w

2 b. Containment Pressure --

w High - High 4 2(b) 3 [d) 1, 2, 3 10 I

.L

~

3. CONTAINMEi4T ISOLATION (CIAS)

a. t1anual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 8

)

b. Containment Pressure -

High 4 2 3[d) 1, 2, 3 9*

i t

f t 4 L

~

{

~

j

i TABLE 3.3-3 (Continued)

EtiGIflEERED SAFETY FEATURE ACTUATI0ri SYSTEM IriSTRUMErlTATI0il

=

T' T1ItilMUM i E TOTAL T10. CHAfi!4ELS CHAtitiELS APPLICABLE Q FUf1CTI0t:AL UtlIT OF CHAtirELS TO TRIP OPERABLE MODES ACTI0ft N

4. MAlti STEAM ATJD FEEDWATER

- ISOLATION (MSIS) i - a. Manual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 8

b. Steam Generator 4/ steam 2/ steam 3/ steam 1, 2, 3 9*

I Pressure - Low generator generator generator (d)

{ 5. C0t1TAlt:MErlT C00LIriG (CCAS)

. a. Itanual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1, 2, 3, 4 b M b. Containment Pressure -

Higl: 4 2 3 (d) 1, 2, 3 9*

[

i l M c. Pressurizer Pressure -

Low 4 2 3 [cl) 1,2,3(a) 9*

l l 6. RECIRCULATI0t1(RAS)

a. Itanual (Trip Buttons)(c) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 8 per train per train per train j

i

! b. Refueling Water l Tank - Low 4 2 3 (d) 1, 2, 3 9*

. r. .. .. ., ... . .

, , . _ . p ., ;. , .. , ... ,,

1 TABLE 3.3-3 (Continued) g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM IriSTUMErlTATIOrl 5

x 5

MINIflUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E

G 7. LOSS OF POWER m a. 4.16 kv Emergency Bus Undervoltage (Loss of Voltage) 2/ Bus 1/ Bus 2/ Bus 1,2,3 8

b. 460 volt Emergency Bus Undervoltage (Degraded Voltage) 1/ Bus 1/ Bus 1/ Bus 1,2,3 8

8. EMERGEf!CY FEEDWATER (EFAS)

a. Manual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 8 per S/G per S/G per S/G

{

Y b. 3G Level and Pressure C (A/B) - Low and AP (A/B) - High 4/SG 2/SG 3/SG (ch 1, 2, 3, 4 9*

l

c. SG Level (A/B) - Low 1 and No S/G Pressure -

! Low Trip (A/B) 4/SG 2/SG 3/SG[d) 1, 2, 3, 4 9*

4 I

?,'

hf .

i

-- *-e,-- . .

i 9  :

L l TABLE 3.3-3 (Continued) 1

TABLE NOTATION (a) Trip function may be bypassed in this MODE when pressurizer  !

y pressure is below 400 psia, bypass shall be automatically removed when pressurizer pressure is > 500 psia. ,

(b) An SIAS signal is first necessary to enable CSAS logic.

t I i (c) Remote manual not provided for RAS. These are local manuals at t i each ESF auxiliary relay cabinet.

~

g) Nom //y low chue/s win be mainkined i,i L 2n OPCAABLF coni/fjon. //oweyy, esse dGISERE inoperable channel may be bypassed for up to  !

90 days from the time of initial loss of OPERABILITY if the inoperable component is located outside of containments 6228446 Tne inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition. T/f e G2K$h inoperable channel may be bypassed until the next refueling if the inoperable component is located inside of containment.

g The provisions of Specification 3.0.4 are not applicable.

t ACTION STATEMENTS p3 ACTION 8 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel

, to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least g HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN W within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Y w ACTION 9 - With the number of OPERABLE channels one less than the F M'. .d'#"" g'"g Mwtdesh w M h *, operation may proceed provided the folloping conditions are satisfied:

Of*W W j[M k One of f/c two

% d22it inoperable channel 5is placed in edest* the j

't.. e m edeed tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> gna s

y-g p g a+-wa wn , e s Ty m'!4M fW &2:_^"?Me wc ' &

Q t ML1M = =M & % & -c &.C " W M#2'M"+MMMMM^!M M'&

y'",66M*::  : k":" *R?!&M!&& 4" gggig

.h *S4%$Cd4MM'#, ,,

G p- q Withinonehour,alignctionallogicunitsreceiving an input from h channel are also placed

h. ,

in the 'et.d edt,44 % M 5 P M 5 Clos tripped , &td d'04.

AfK AN.S A S - t/N/ 7" Q -

3/4 3-/V

- _ _ _ _ _ _ _ _ _ . . =-

~ -

~

. 1 TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATf0N TRIP VALUES g

E:= ALLOWABLE 9 TRIP SETPOINT VALUES FUNCTIONAL UNIT f

E 1. SAFETY INJECTION (SIAS) Not Applicable

a. Manual (Trip Buttons) Not Applicable Q

m i 19.024 psia

b. Containment Pressure - lligh 1 18.4 psia 17GG l]l1.3 57

> Tf44Lpsia (1) >768Q psia (1)

c. Pressurizer Pressure - Low

2. CONTAINMENT SPRAY (CSAS)

Not Applicable Not Applicable

a. Manual (Trip Buttons)

Containment Pressure -- liigh-liigh 1 23.3 psia 1 23.624 psia b.

3. CONTAINMENT ISOLATION (CIAS)

Not Applicable Not Applicable

a. Manual (Trip Buttons)

Containment Pressure - liigh 1 18.4 psia i 19.024 psia b.

w

~

.. - am. ___

_ . .a a. &__<._ . n.w- _ anam w .. . -

O O O  !

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES s ALLOWABLE FUNCTIONAL UNIT. . TRIP VAtuF. VALUES E 4. MAIN STEAM AND FEEDWAT E ISOLATION (MSIS)

Z a. Manual (Trip Buttons) Not Applicable Not Applicable N 7f/ 7 A9. 6 / 3 ti . Steam Generator Pressure - Low >_'72R psia (2) > 7t)6 4 psia (2) l

}

I

5. CONTAINMENT COOLING (CCAS)

a. Manual (Trip Buttons) Not Applicable Not Applicable  !

b. Containment Pressure - High 1 18.4 psia 1 19.024 psia 17&G ' 1712.7 57

c. Pressurizer Pressure - Low >.'1940. psia ( 1 ) > 7686 A psia (1)  :

R 6. RECIRCULATION (RAS)

L

a. Manual (Trip Buttons) Not Applicable Not Applicable

[t I' C b. Refueling Water Tank . Low 54,400 1 2,370 gallons between 51,050 and 58,600 1 (equivalent to 6.0 1 0.5% gallons (equivalent to indicated level) between h Rtt and M 86%

indicated / level) 6.999

7. LOSS OF POWER

a. 4.16 kv Emergency Bus Undervoltage (Loss of Voltage) 3120 volts (4) 3120 volts (4)

b. 460 volt Emergency Bus Undcrvoltage 423 + 2.0 volts

~

423 + 4.0 volts (Degraded Voltage) with an 8.0 1 0.5 with an 8.0 1 0.8 second time delay second time delay

'o .-

g 1ABLE 3.3-4 (Continued),

ENGINEERED SAFELY ffATURL ACTUA110N SYSTEM INSTRUMENTA110N Trip VALUES ALLOWABLE g FUNCTIONAL UNIT TRIP VALUE VALUES

[ 8 EMERGENCY FEEDWATER (EfAS)

a. Manual (Trip utittons) NotApplicable Not A licable

b. Steam Generator (AAB) Level-Low >

90 b (3) > h dt. (3)

~ %344 l

c. Steam Generator Ap-liigh (SG-A > SG-B) < M psi

^

< 4tMb si

d. Steam Generator Ap-liigh (SG-B > SG-A) psi psi

e. Steam Generator (A&B) Pressure - Low > 7fR psia (2) >7tEk psia (2)

~ ']SI ~ 31'l.4I3 during l A ressuriter Prwurt R (~1TValue may be decreased manually, to a minimum of > 100 psia , as pikssbas,,,ed re, ,,, << acHen m sure !E infeduc ed , provided

1. " the margin between the pressurizer pressure and tIiis value is maintained at < 200 pst ; the setpoi y shall be increased automatically as pressurizer pressure is increased until Ihe trip setpoint is reaciien.

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

Jaro.n a plwe) reatucHon in slenon

,~

(2) Value may be decreased manually atstenin gcnerator-tmessuee-D,genero}or pressure-reduce 41,, provided the m between the steam generator pressure and this value is maintained at < 200 psi; the setpoint

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

3- (3) % of the distance between steam generator upper and lower level instrument nozzles.

(4) Inverse time relay set value, not a trip value. lhe zero voltage trip will occur in 0.75 1 0.075

@ seconds.

r+

6 i

t

I l .i 3/4.4 REACTOR C00LAN* SYSTEM j REACTOR COOLANT LOOPS 4

I LIMITING CONDITICN FOR OPERATION 3.4.1 Both reactor coolant loops and both reactor coolant pumps in each f loop shall be in operation.

1 1 APPLICABILITY: As noted below, but excluding MODE 6.*

^

i Wr t R Ve!M D C PE R' 4-n e n) ***

ACT10N: GNSnil "A IUs **" +~ ceac.he ecol- / paanps in ep c. h an

, be in a+ less + HOT ,$TANDBY wHhin one Gar.

Da e -e ico A nar.en***

a MODES. 1 and 2:

i

a. With one reactor coolant pumo not in operation, STARTUP and/cr continued POWER OPERATION may ::roceed crovided THER"AL POWER is restricted to < **" of RATED THER!iAL POWER and the set-

. point for the Linear Pcwer Level - High trip has been reduced to the value specifiec in Specification 2.2.1 for operation j with three reactor coolant pumps operating.

g_. b. With two reactor coolant pumps in ooposite loops not in 0: era-mion, STARTUP and/or continued POWER OPERATION may croceed 2

provided THER".AL POWER is restricted to < ***,' of RATED THER:tAL i

POWER and the set:oint for the Linear Po er Level - Hich trip j <

has been reduced to tne value specified in Specification 2.2.1 for operation with two reactor coolant pumps operating in i opposite loops.

j

c. With two reactor coolant pumps in the same loop not in coera-tion, STARTUP and/or continuec POWER OPERATION may proceed provided the water level in both steam generators is maintained above the Steam Generator Water Level-Low trip set:cint, the THERMAL POWER is restricted to < **% of RATED THERMAL POWER, and the setpoint for the Linear Power Level - High trip has been reduced to the value specified in Specification 2.2.1 for operation with two eactor coolant pumps operating in the same loop. -

l 1

"See Speciai Test Exception 3.10.3 g

j **These values lef t blank pending NRC approval of M analyses for coera tion with less than four reactor coolant pumps aceratinc.

J W .

NRC

? ael ! copc4 appecval cyerahen s<fely ana is nof lyses.allev<A in ModosIand;;;yedsng I

ARKANSAS - UNIT 2 3/4 4-1 i n ., . . . . - - . - .

, . . - - - ~ ~ - -~

l

.s i

MODE 3: t ,

f REACTOR COOLANT SYSTEM Operation may proceed provided two -

t reactor coolant loops are in operation with at least one reactor coolant pump i

ACTION: (Continued) in each loop. With less than one 1

, INSERT reactor coolant pump in each locp in operation have at least ene pumo in

. MODES'4 g 4 and 5: each loop in coeration within one hour i nr he in at least HOT SHUTDOWM within the~ next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Operation may proceed provided at least one reactor coolant loop is in operation with an associated reactor coolant pump or shutdown cooling pump.* The provisions of Specifications 3.0.3 and 3.0.4 are not applicable All reactor coolant pumps and shutdown cooling pumps may bE de-ener; Ied for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, provided no operations are pennitted which could cause dilution of the reactor coolant system baron concentration.

a F.e Q

, SURVEILLANCE RECUIREMENTS

! i

! 4.4.1 The Reactor Protective Instrumentation channels specified in the applicable ACTION statement above shall be verified to have had their trip setpoints changed to the values specified in Specification 2.2.1

for the applicable number of reactor coolant pumps operating either

a. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after switching to a different pump combination if switch is made while operating, or

b. Prior to reactor criticality if switch is made while shutdown.

l l

'l ARKANSAS - UNIT 2 3/4 4-2

7

~

^

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

> TABLE 3.6-1 CONTAINMENT ISOLATION VALVES 5

m PENETRATION ISOLATION E NUMBER VALVE NUMBER FUNCTION TIME (SEC)

Q m A. C0flTAINMENT ISOLATION 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) 1 20 2SV-5843-2 RCS & Pressurizer Sample Isolation (outside) 1 20 2P9 2CV-6207-2 II.P. Nitrogen to SI Tanks (outside) 1 20 2P14 2CV-4821-1 CVCS L/D Isolation (inside) 1 35 2CV-4823-2 CVCS L/D isolation (outside) 1 20 2P18 2CV-4846-1 RCP Seal Return Isolation (inside) < 25 R

a cv- 4947- 2. %i-Mtl4.-2.- RCP Seal Return Isolation (outside) 7 20 2P31 2CV-2401-1 Containment Vent lleader (inside) 7 20

{ 2CV-2400-2 Containment Vent lleader (outside) 1 20 cn 2P37 2SV-5878-1 Quench Tark Liquid Sample (inside) 1 20 2SV-5871-2 Quench Tank Liquid Sample (outside) < 20 2SV-5876-2 SI Tanks Sample Is.olation (outside) 1 20 2P39 2CV-4690-2 Quench Tank Makeup & Demin Water Supply Isolation (outside) 1 20 2P40 2CV-3200-2 Fire Water Isolation (outside) 1 20 2P41 2CV-6213-2 L.P. Nitrogen Supply Isolation (outside) 1 20 2P51 2CV-3852-1 Chilled Water Supply Isolation (outside) 1 20 2P52 2CV-5236-1 CCW to RCP Coolers Isolation (outside) 1 20 2P59 2CV-3850-2 Chilled Water Return Isolation (inside) 1 20 2CV-3851-1 Chilled Water Return Isolation (outside) 1 20 2P60 . 2CV-5254-2 CCW from RCP Coolers Isolation (inside) 1 20 2CV-5255-1 CCW from RCP Coolers Isolation (inside) 1 20 g.,,'

L

j m. - _ -_ m <-- ._ n-we _

1 A

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m le1

,: ^ .C

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'. C' CJ 3 e 3 C I. -1 W CuC -

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C Cu - +Jr w C- t Z EO C; WC > U: @

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[Wm]

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~ ;< i s,. Q

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• I:=

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W wC c

.J wi l' C3 w!C' -

< M i

H J:=l d!d

~

O w;=1 -

b 3

Cl'O LWi i x f'

+J O

=

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> -e =;or uC m ci e-wi<t Z +J Z Cl c mi.= l' > - c ; Db

& i>- -E C O l  :& 2 % -- 0 l

' W'jW 9 0 D O J G., ba C CD L

< cA U B t B = C. a l 3 O C1 E l C LJ C' .M 3 l .C.

J g~Z +J - N m C L J*J  %%

CL

'-@ +>

<f E.E o ==

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> C C l O r w 4

ARXANSAS - UNIT 2 3/4 7-3 a' .~- w-_-- .-. . . , _ , __ *T

e

?

l PLAIT SYSTEMS-1 EMERGE? ICY FEECWATER SYSTEM LIMITIriG C0!iDITI0t1 FOR OPERATI0t1

w 3.7.1.2 Two emergency feedwater pumps and associated flow paths snall Y be OPERABLE with

a. One motor driven pump capable of being powered frca an OPERABLE emergency bus, and

b. One turbine driven pump capable of being powered frcm an

OPERABLE steam supply system.

j APPLICABILITY: MODES 1, 2 and 3.

ACTIO!1:

! With one emergency feedwater pump inoperable, restore the inoperable pump to OPERASLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTCOW'i wir" n the l_ next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

i 1

12 SURVE'LLA? ICE REQUIREME?iTS Ii l ;

i 4.7.1.2 Each emergency feedwater pump shall be demonstrated OPERABLE:

4 a a. At least once per 31 days by:

1. Verifying that the turbine driven pump develops a l discharge pressure of > 1200 asig at a flow of > E60 gpm.

a ;" Oc 56Z dsU 3'On Sa?;'b PMCC 7a i: I ' -P '? '""

j hELETE) '6o5 pais cou U.c ,, v . . . ; 3x d i: ; 2600 --

The pro-visions of Specification 4.0.4 are not applicable.

2. Verifying that each val re (manual, power operated or 4 automatic) in the ficw path that is not locked, sealeo, or otherwise secured in position, is in its correct j position.

4

.2 l1

_ O j ARKA:iSAS - UtiIT 2 3/4 7-5 l

~ ,_

7-

i PLAfiT SYSTEMS \.

~..

p.

ACTIVITY

\

j LIMITING CONDITIC'1 FOR CPERATICN t-t l

1

. 3.7.1.2 The scecific activity of the secondary coolant system shall be .

1.h f aci/ gram COSE EOUIVALENT I-131.

c.c 46 APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

0.04G With the s:ecific activity of the secondary ccolant system > S-tJ uCi/

l gram COSE EQUIVALENT I-131, ce in at least HOT STANCEY witnin 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and in COLD SHUT:0'4N within :ne following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

I h

SURVEILLANCE RECUIREMENT5

~Q

,- f O..

4.7.1.4.1The specific activity of the secondary ecolant system shall be

! determined :c be within the limit by performance of the sancling and analysis program of Table 0.7-2.

V.7. l . 4. 2 l The total primary to secondary leakage shall be determined to be less than MD 100 gpd at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. ,

2 i

1 6 -

i in, s

O ARKANSAS - UNIT 2 3/4 7-8 1

= &&

j i

l d c i REFUELItlG OPERATI0!iS REFUELIriG PdACHINE OPERASILITY l

LIMITING CONDITION FOR OPERATION

.] [onsul d 3.9.6 The refueling machine shall be used for movement of TEh-sc. fuel

i assemblies and shall be OPERABLE with

J

a. A minimum capacity of 3750 pounds, I

b. An overload cut off limit of < 100 pounds plus the comoined l weight of one fuel assembly, cne part length CEA, and the

, grapple in the " fuel only" region, and 1

i c. An overload cut off limit of < 100 pounds plus the comoined I

weight of one fuel assembly, one part lengtn CEA, tne grapole, and the hoitt box in the " fuel plus hoist box" region.

, APPLICASILITY: During movement.of CEAs or fuel assemblies witnin the i reactor pressure vessel.

ACTION:

With the requirenents for refueling machine OPERABILITY not satisfied,

, i suspend its use from operations involving the movement of CEAs and fuel i assemblies within the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

4 4

i SURVEILLAjiCE RECUIRE?>ENTS

[ccien]

l 4.9.6 The refueling machine shall be demonstrated OPERABLE within 72 1 hours prior to the start of movement of h fuel assemolies within the reactor pressure vessel by performing a load test of at least 3750 pounds and demonstrating automatic load cut offs wnen the crane loads exceed 100 pounds plus the applicable loads.

l l 3

i d

! 1

?O ARKANSAS - UNIT 2 3/4 9-7

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

' 3/4.1 REACTIVITY CONTROL SYSTEMS BASES i

b 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN j A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operat'ng conditions, 2) the reactivity transients associated with costulated accident conditions are controllable witnin acceptable limits, and 3) the reactor will be maintained sufficiently l subcritical to preclude inadvertent criticality in the shutdown concition.

l

- SHUTDOWN MARGIN requirements vary throughout core life as a function

! of fuel depletion, RCS bcron c:ncentration, and RCS T . The most j restrictive condition occurs at EOL, with T at no 182d operating i temperature, and is associated with a postui h,ed steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTCOWN MARGIN tf 5.0% ak/k is required to control the 3

reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is 1 based upon this limiting condition and is consistent with FSAR safety I analysis assumptions. With T < 200'F, the reactivity transients result-i

'} ing from any costulated accidME are minimal and a Kak/k shutdown margin provides adequate protection.

5%

- 3/4.1.1.3 BOR0f, DILUTION i A minimum flow rate of at least 30C0 GPM provides adequate mixing, prevents stratification and ensures tnat reactivity cnanges will be

, gradual during boron concentration reductions in the Reactor Ccolant System. A ficw rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9,975 cubic feet in approximately 25 minutes. The reactivity change rate associated with boron ccncen-tration reductions will therefore be within the capability of operator recognition and control .

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)

The limitations on MTC are provided to ensure th&t the assumotions used in the accident and transient analysis remain valid througn each fuel cycle. The surveillance requirements for ceasurement of the MTC during each fuel cycle are adequate to confinn tne MTC value since tnis l

coefficient changes s1 wly due principally to the reduction in RCS bcron j concentration associattd with fuel burnup. The confirmation that ne

measured MTC value is within its limit provides assurances that the l coefficient will be maintained within acceptable values throughout each 3 fuel cycle.

l ARKANSAS - UN!T 2 B 3/41-1 w

A REACTIVITY CONTROL SYSTEMS ,

I BASES I

) 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY 0

ej This specification ensures that the reactor will not be r..ade y critical with the Reactor Coolant System average temperature less than f 525 F. This limitation is required to ensure 1) the moderator temperature Y .

coefficient is within its analyzed temperature range, 2) the protective f  ; instrumentation is within its normal coerating range, 3) the pres' 'er l is mable of being in an OPERABLE status with a steam bubble, and he reac w. pressure vessel is above its minimum RT mperature.

NDT 3/4.1.2 BORATION SYSTEMS 1

r The coron injection system ensures that negat1ve reactivity control is available during each mode of facility operation. The components required to perforn this 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 4

OPERABLE diesel generators, b With the RCS average temperature above 200'F, a minimum of two separate and redundant baron injection systems are provided to ensure single functional capability in the event an assumed failure renders cne of the systems inoperable. Allowable out-of-service periods ensure tha' minor component repair or corrective action may be ccmpleted without

, undue risk to overall facility safety from injection system failures during the repair period.

ihe boration capability of either system is sufficien/to provide a SHUTDOWN MARGIN from expected operating conditions of M Ak/k after xenon decay and cooldown to 200*F. The maximum expected boration cap-ability requirement occurs at EOL frca full power equilibriun xenon conditions and requires boric acid soluti.on from the boric acid makeup tanks in the allowable concentrations and volumes of Specification 3.1.2.8 or'MAC. gallors of 1731 ppm borated water from the refueling water tank.

5 % 455 _

With the R:S temperature below 200*F one injection system is acceptable without single failure consideration on the basis of the stable reactiv ty 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.

~.-

,j ARKANSAS - UNIT 2 8 3/4 1-2 Y- . . .. - - - -

REACTIVITY CONTROL SYSTEMS

. BASES j 50% The boron capability required below 200 F is based upon providing a

W k/k SHUTDOWN MARGIN after xenon decay and cooldown from 200*F to 140'F.

4 This condition requires either Mgallons of 1731 ppm borated water

! from the refueling water tank or[ boric acid solution from the boric acid makeup tanks in accordance with/the requirements of Specification 3.1.2.7.

J 8185-The contained water volume limits includes allowance for water not available because of discharge line location and other physical j characteristics. The3h%0. gallon limit for the refueling water tank

. isbaseduponhavingan[indicatedlevelinthetankofatleast2".

35,250 The OPERABILITY of one boron injection system during REFUELING

l. ensures that this system is available for reactivity control while in MODE 6.

The limits on contained water volume and boron concentration of the RfT also ensure a pH value of between 8.9 and 11.0 for the solution 1

recirculated within containment after a LOCA. This pH band minimizes the

] evolution of iodine and minimizes the effect of chloride and caustic g stress corrosion on mechanical systems and components.

V i 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power i distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of CEA mis.lignments are limited to acceptable levels.

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

i The ACTION statements npplicable to a stuck or untrippable CEA, to two or more inoperable CEAs, and to a large misalignment (> 19 inches) of two or more CEAs, require a prompt shutdown of the reactor since any of these conditions may be indicative o# a possible 1,oss of' mechanical functional capability of the CEAs and in the event of a stuck or untrip-jable CEA, the loss of SHUTDOWN MARGIN.

For small misalignments (< 19 inches) of the CEAs, there is 1) a small effect on the time deoendent long tenn power distributions rela-tive to those used in generating LCOs and LSSS setpoints, 2) a smell effect on the available SHUTDOWN MARGIN, and 3) a small effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION l'

ARKANSAS'- UNIT 2 B 3/4 1-2 N - w - .m -,- ,.-m -, ,,,p_ . ,.y, , , , , _. ,,

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

F 1

t. .a f : . .? ce"a :..=. D '. C'. O.~.U'^".'.'*.".. T: .u. .

an R " .t r -.b J/*.4.1  ; 7....-

A . e t.m M h.iri. 7.a.-- tt The limitation en linear nea: rate ensures :nat in :ne event of a eu e

LOCA, the peak tem erature of the fuel claccing will not exceed 2200 F.

t Either of :ne two core ;cwer distribution monitoring systers, :ne Core C;erating Limit Su:ervisory Sys:e- (COL 55) and ine Lccal : we-

,v ens 1:y cnannels in :ne tore rrctecticn ca,tculators turus), Oravice adetuate acnitoring of :ne core power cistricution and are ca:atie Of verifying 2na: the linear nea rate cces not exceed its its limits.

.he C 4 ba ;erforms this function by continuousiv roni:Orinc :ne core

, wa.r ,... . .r . l g . '. .= . .>i-.12

.'m 3 .= ---

sm a. .-c.w . v^ a. =..-, o i.a. me- . r. a. -

d s .o n e i r.a, **

.. . -o a. .= l l .w a .a l a. . a .a k l i - oc .= . n a. .= . ..=.a. .  :==-.--

.. . s. m'.- .a. =..m --

o 2-cr _"a_ l c w *..n i. s c_ = l c u 1.= .a.d. - v,, a. r l a v a l = c o r a_ s . .n .= *. .. . a.

.. 1 4. m i .-2

^#

v.

~

r. i 3. ra.

3.1-1 are not exceecec.

The COLS5 calculated core :caer anc the CCL55 calculated core

, ;cwer Operating limits based cn linear neat rate are centinucusly monitored and displayec :0 ne a:erator. A COL 55 alarm is anr.unciated I

/;

O i r. *una. avan *. .r.aw * ...

i h j s- - rg

.. y j c a s .= d a. -", .=

• a.

for nCr~al ste=cy state C eration.

• a. cova . a_ - ^ w a_ r a_ x .- =. a. .- -3

.c

~.,=.".#.r.*.**."a. . o

..n a. w. r a. .o w a. r.

l # #a a_ .= r . . e^ .= *. r .= *. a. ".. .a_ .= '. i. rg-Normal rs=c'Or ;cwer *ransients or vc . .=*.in l. ni .. i .

'4 i . 4. *.

e equi ment failures wnicn d0 not require a react 0r tri; may result in 1 tnis care ;cwer c:erating limit bein; excesced. In t.e avent :nis y occurs, COL 55 alarms will te annuncia:e:. 'f :ne even wnicn :auses :na

,z- 4 - .- w =. , x , ,_ a_ =. e. r a_ v 't .2 . - :- .en4.ivos .en wn 4 ... ,..,..,-. . .....=_ .,s.a_

C UL b 'I n 1

.. _ . .o o. .

l sa . f. e- *.,y 1 4 ,. . .24. . , .= re.=w*.-. .

• rl' n.

. will -a_ #. n i *. i 2

• a. d.

. . "wy  ?...a. .

a_~= ."*.C r P ra .a. .*.i Va.

I ...u e . r _-a_ c. . . 3. ..<.r o r. mLa. .. n . i t. . . .3..] - , , l .2 . 4. r..n. s.

r: . a_ *i 2...  : . o_ _s e. or e .a .. r.s c_ 1 4. . 4. .

i n u- l u d. e. s .a. - a. r . r i .= *.a. J a c_ - a r a i. , . < .

> n d. . e .. .= s . .. #..=c r2 - a- -_ =_ .c 2 .= r.v . .^ .- . . ". 4. a.

c..:f c:. m . . 4. s.o n .

c 1 =.y o 1 . n.

.. .. . s. .. o. . ~= x i _ . . . 'i i n =_ .= r - oo_ .= *. r =. . a. . = i - 'a '. a. *. .= d. by COLSS is greater than or ecual to tna: existing in :ne ccre. To ensure

• c h a *. * . . a. da_s.-n e

4

. ara34.n .- .= # a. . /

. 1' s .~=

. 1' rn*. .= i . ,a_ e , * ..b.e Cm - . . . - - +a -a- f' o 3- ,-

j y program inciuces a%measurednt uncertainty factor ofT.S'L an _

eagineering uncertainty factor of 1.03, a THERFAL PC'a'ER measure en uncertainty fac:cr of 1.02 and at:ropriate uncertain *y anc :eralty

,ac: Ors ter riux peaking augmentation and red bow.

Parameters required c maintain the c:erating lini: Ocwer level based on linear heat ra*e, margin to CNE and Octai core ;cxer are also monit; ed by the CPCs.

. Therefore, in the event tna: :ne COL 55 is not being used, c;eration witnin tne limits of Figure'27 -4L can te maintainec

~

j l

3 by utilizing a ; redetermined lccal ;cwer density margin and a ::ai core

power limit in the CPC trip channels. The atove listec uncertainty and 4 re

..ilty factors are also included in the CPCs, 3. 2- 4

. O l ARKANSAS - UNIT 2 B 3/4 2-1 1

i

~~

Es-, -

,n ._ - , --

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

~

( POWER DISTRIBUTION LIMITS C BASES I c 3/4.2.2 RADIAL PEAYING FACTORS (y l PLAN % MDIAL FfMNG FAc7CRS

! Limiting the values of the p'.ana, ,adi;l pak4% # 2ctc.rs (.c) used in the COLSS and CPCs to values equal to or greater than the E PyNAR RADIAL PEAXIn FACTDRS j measured giana, redici pnki ; futr s(y) c;ovides assurance that the limits calculated by COLS5 c:d the CPCs remPin valid. Data from the p i

PfAXW6 F44 gre detectors are used for determining the measured' pia.. .c-midi.LL , ggt rmir.; 'r:7 - pgegc vg requirements for determining l thge peanu,fsgrgggj rovides assurance that the

.am. r m , __

f used in COLSS and the CPCs remain valid throughout the fuel cycle. Determining the measured ;lan;r ad4 _ PLANAR RApm.

FPur6 9croseca.in; fr ~ t after each fuel loading prior to exceeding 70% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

i j 3/4.2.3 AZIMUTHAL POWER TILT - T a s

J r

( The limitatio:1s on the AZIMUTHAL POWER TILT are provided to ensure s.

that design safety margins are maintained. An AZIMUTHAL POWER TILT greater than 0.10 is not expected and if it should occur, operation is restricted to only those conditions required to identiff the cause of

, the tilt. The tilt is nomally calculated by COLSS. The surveillance requirements specified wnen COLSS is out of service provide an acceptable means of detecting tne presence of a steady state tilt. It is necessary to explicitly account for power asymetries because the radial oe2 king factors used in the core power distribution calculations are based on an untilted power distrioution.

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 location is of the form:

P tiltI untilt

=1+T q gcos(e-ej _

t l where:

T is the peak fractior.al tilt amplitude a' the core periphery q

l 9 is the radial normalizing factor e is the azimuthal core location f

Go is the azimuthal core location of maximum tilt '

i ARKANSAS - UNIT 2 B 3/4 2-2 .

• ==we e.

-ww a

1 2

e n..a. r e s . .2:4.U

. a. c U4. = :a .i tn.. i. T.44uri.e C. *. 7. .C

-n .n c. ... .a.

J 1

< p.ll./p c

e un.ll. 1s .w o

...e ... 1% . . a. ., ..,. a - e. a.

. 3 .n e..., . t. .n 1. . .. .. .a.

q presence ci a tilt to the pcwer a :nat ICCa!ien witr. no ti,l!.

a j g , -

.j..t. u ,i x,,-9.-r...

i

. enobi J

Ihe limita* ion on CNER as a func :cn cf */ AL SHAPE !N'EX re:re-a s a. r. '.s

. 3 r. . n s a. -V a

• 4. v a.

. a. n V a. l.- . a. .r #. c . a r.= . d. r. 3 -^

. .. r~~ i . 4. . o s - .. - co.4. s- *. a. . . w '. *. ;*.

I the safety analysis assuccticns anc wnicn nave ccen analytically ce in-

strated adecuate

c maintain an accectable minimum CNER :nrcugrcut all 1 anticicated cceraticnal cccurrences, cf wnicn :ne less cf ficw transient i s *o n * .. *. ^ s- . li i.i.n3 m 0 ."a. . .= *. i C n . #. . . . a. . a. r a. w i. .o- .= '. 'i:. 0 . = * . .r .=.*.va.

• 9is

w. li-4.*.

m . 4. .- a. s .= s 2 " r .=. n a. *.o" a * .a r. =. . . a. -. . .= 't a. . . . i r. d. m" .

r -". . 'sa.. " : :. a 4. '1 '. . c.

24 in . . . a. a. ,/ c a .e. c ;, .a 1.wss m;

4. n . ..e.: o se.e.o..

4 a

i. ~sn2.4i n a. . .. c. *twa 4 C i +..h e r . #.

. -..r.. .a *wc .-. ra. r . w a. r. d. i s . . i "". t. i.. n -u- m4 . .. r .i.n , e. .s s . a ~. e , . . . a.

i Core Ocerating Limi: Sucerviscry Syster (CCL55) anc :ne CNER cnar.r.eis in Ce r e e r .a. r. *.ie n ^. = ' r".. .': a .. r s ( C r r.3 )' , - r. ~s . w-4 a. <

4

• ..n a.

. 2 . c- r. o. .= ~. a. .vn4.... ,r4.n, .

• • a .. r a. -. .w a. r ' i. *2 r i " ".. *. #. .- r. .= r.d. .= r. a .=.. c-.1.=. .~#. ". e . 4. #. .<.

. . ~ ' r. - . . . =. . . a~ . N : :. .

dces r,c*. ".clas. .

a 4, . .e 1 4u . 4. *. .: . 'he

, , CLC.. - a.

. . r #. - ~~ s . . ', 2

1. . "< . .- *..4. r. ' v. . n.

.ine wucby

. . ..cni.- 'ir.y *.o- a. -a w ra. . r.w a r '. 4. . t. r. 4 "i.r. . 4 .- a .r. o =o- r. .= 't ~. J 'L .a * '. o- . ..- ^ . a.

j cperatir.s limit ccrrec;cnding :c :ne ailcwable 7.inirur CNEP. Reac:cr

• e.3 cceration at or belcw :nis calculatec ceaer level assures na: the limits cf Figure"??6-2. are no: violated. The CCLSS calculaticn of ccre 1

- aer.e..r a. r. 2.. .4. . ~, l i .. i ' *. a s n- .. e -...'N:.:. i -lu a.t a . . r . ^..

o . 4. .= ?. =. '; r~- o. . . =..

'. . '..y . = ~ . . .

3 penalty fac:crs necessary tc crovice a 95/95 ccnficer.ce levei :na: .e g,19

r. ,., r a. - .~.a.,, 2r. .en4..o r so v:- s c.e ,i a. s a. . w , .N. - - ., .o. l .- e .. .-,., 17s . .2 ] r.. . ,i. .2 2. 4 cy COLSS, is less :nen er ecuai tc cna: whicn wcuid actually be recuired i r. *.o* a. .-.r a. . To e..n s u ra, *"=t. o ... ." a. c 4. : . r=

. . ' . . *- .. 3 - =. #. a *. v i s r .= ". r. * =. i. r a ". , -

y

+. n a. e.r.a.ca .r- .. u.a ..

..w r2- 4. n e. l , s.a. e. .

. 2 na N. . -e .2so. ra~ c.

. . . ... . a. .4 .

2 . n . .,, e. ., c.. . . r. xy of T.R an encineerinc uncertainty #actcr of 1.03, a THE 9L !'aER

/.053

~

q c~a .. = e " r c. - c- o *. 'a .a ~ . r * = i n .,v ' #. .= c *.. . v^ #. 1. c.2 a r.d. .. =- -r. - . . r *. =

• a.

. "e n r.a. r. *..= 4. n *.y a ad o.

penalty fac:crs for flux ceasing augrentation and roc bcw.

a rarameters re:uire. s to maintain the rargin to u .v.: anu tctal ccre

cwer are alsc renitored by tne C?Cs. Therefcre, in tr.e event tha

*;j;.3.1- 4 COLSS is not being used, cceraticn within the limits of Figure'?? s1.can 1

be r.intained by utilicing a prede:erained C GR as a functicn of A!.!AL SHA?E INCEX and by enitcring the CFC tric cnennels. The above listec uncertainty and penalty factors are 1150 included in the CPC.

1 a

~

.I O

ARKANSAS - UNIT 2 8 3/4 2 3

" - W ~" " '

r w ,

, , w w w w ,_ y-r -

'mm n , - ~ ~ "Y

a POWER DISTRI3UTION LIMITS j BASES ,

j 3/4.2.5 RCS FLOW RATE 1

l This specification is provided to ensure that the actual RCS total j flow rate is maintained at or above the minimum value used in the LOCA safety analyses.

l I N E AVERAGE COOLANT TEMPERATURE @ELETEl l

j This specification is pr  ::.a une assumptiens used for

= id.

the initial cenai+'-

::, u.e cucn safety ar.alyses e-.

REPLACEMENT BASES 3/4.2.5, 3/4.2.7 and 3/4.2.8 1 >

1 3/4.2.6 REACTOR CCOLANT COLD LEG TEMPERATURE ,~~

I inis specification is provided to ensure that the actual value of Reactor coolant, cold leg temperature is maintained within the range of values used in the safety i analyses.

l , 3/J.2.7 AXIAL SHAPE INCEX This specification is provided t] ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses.

3/4.2.8 PRESSURIZER PRESSURE ~

This specification is provided to ensure that the actual value of pressurizer

, pressure :s maintained within tr.e range of values used in the safety analyses.

I t

i i

l l .

'7, I

ARXANSAS - UNIT 2 B 3/4 2-4

'* - e y .-+-e f

_.____-._?_.,.-_. _ -. ,,._ , _ , , , . , _ _ - --

," 7/4.4 REACTOR COOLANT SYSTEM i

BASES

,, I i

3/4.4.1 REACTOR COOLANT LOOPS i

l- The plant is designed to ocerate with both reactor coolant locos and I associated reactor coolant pumos in operation, and maintain DNER above j J.2N M during all normal operations and anticicated transients. STARTUP I and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for tne Pcwer

Level-High, Reactor Coolant Flcw-Low, ar.c Thermal Margin /Lew Pressure i trips have been reduced to their specified values. Reducing tnese trio

{ setpoints ensures that the DNBR will be maintained aboveMcuring y*g j three pump operation and that during two pumo oceration the core void i

fraction will be limited to ensure parallel channel flow stability witnin the core and thereby prevent premature CNE.

A single reactor coolant loco with its steam generator filled above the low level trip set:oint provides sufficient heat removal capacility 1 for core cooling while in MODES 2 and 3; however, single failure consi-

derations require plant cooldown if cocconent repairs and/or corrective i actions cannot be made within tne allowable cut-of-service time.

( v 3/4.4.2 and 3/4.a.3 SAFETY VALVES Wo,0cc

' The pressurizer coda safety valves operate to prevent the RCS from being pressurized aoove its Safety Limit of 2750 psia. Each safet/ valve 4

is designed to relieve 3M,iGO lbs per hour of saturated steam at the val /e setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutcown. In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connectec to the RCS, provides overpressure relief caca-bility and will prevent RCS overpressurizati)n.

During operation, all pressurizer code safety valves must be OPERAELE to prevent the RCS from being pressurized nove its safety limit of 2750 psia. The ccmbined relief capacity of these valves is sufficient to limit the Reactor Coolant System pressura to within its Safety Limit of 2750 psia follcwing 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 loss of turbine) and also assuming no operation of the steam dump valves.

ARKANSAS - UNIT 2 B 3/4 4-1

I REACTOR COOLANT SYSTEM j BASES 1

1 d

The M steam aenerator tube leakace limit of i W., ivi ;I ' : t"--

j scoci a ;.u 1 5 ec. w a :d ; th; do::;: ::rt-ibutic" '" - t"e *" * ' " M a j niii Le l u.. i kd a a 2.T.e l l f r;.c t i on o f N r t 100 ' "i t: 4" the e"e"* '#

4 e i wc1 a me. gcr.crator tube ru;'ture or stear '#ne tre9 '"e '

• 7 limit 15 COndialenL wiin v.t c 5 a v..@ i.i G O 3 uSCd . ' the ar,alyE i E C # t*"e i uuu i.2: . "he 0.5 GPM Tecia r li-i t per steam generator ensures that steam generator tube integrity is maintained in the event of a main

steam line rupture or under LOCA conditions.

I PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since

it ,nay be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE B0UNDARY LEAKAGE i, requires the unit to be promptly placed in COLD SHUTDOWN.

1 3/4.4.7 CHEMISTRY g the limitations on Reactor Coolant System chemistry ensure that g corrosion of the Reactor Coolant System 's minimized and reduce the j potential for Reactor Coolant System leakage or failure due to stress -%

(

corrosion. Maintaining the chemistry within the Steady State Limits 1 y provides adequate corrosion protection to ensure the structural integrity j of the Reactor Coolant System over the life of the plant. The associated S effects of exceeding the oxygen, chloride and fluoride limits are time J and temperature dependent. Corrosion studies s ow that operation may be continued with contaminant concentration leveis in excess of the Steady i

State Limits, up to the Transient Limits, for the soecified limited time i intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

1 The surveillance requirements provide adequate assurance that con-

! centrations in excess of the limits will be detected in sufficient time to take corrective action. ,

3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not

, exceed an appropriately small fraction of Part 100 limits following a 5

3 3

/

drl ARKANSAS - UNIT 2 B 3/4 4-4 1

4 W WW WM m ,

-- - ---yy.

f'WMG f 3 - we , -

REACTOR COOL ANT SYSTER

, (

BASES i

} steam ,enerator tube rupture accident in conjunction with an assumed J steady state primary-to-secondary steam generator leakage rate of%Q, loo 6PD j "Si% and a concurrent loss of offst% electrical power. The values for the limits on specific activity repr2sent interic limits based upon a l parametric evaluation by the NRC of typical site locations. These i values are conservative in that specific site parameters of the Arkansas i Nuclear One site, such as site boundary location and meteorological con-ditions, were not considered in this evaluation. The NRC is finalizing i site specific criteria which will be used as the basis for the reevaluation

! of the specific activity limits of this site. This reevaluation may l result in higher limits.

' l t The ACTION statement pemitting POWER OPERATION to continue for

! '; limited time periods with the primary coolant's specific activity > l .0 pCi/ gram f0SE EQUIVALENT I-131, but within the allowable limit shown on l

Figure 3. 4-1, accommodates possible iodine spiking phenomenon which i may occut .following changes in THERMAL F0WER. Operation witn specific j activity levels exceeding 1.0 pCi/ gram DOSE EQUIVALENT I-131 but within

the limits shown on Figure 3.4-1 must be restricted to no more than 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> per year (approximately 10 percent of the unit's yearly operating time) since the activity levels allowed by Figure 3.4-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor c# up to 20 following

, . a postulated steam generator tube rupture. The reporting of cumulative operating time over 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any 6 month consecutive period with > 1.0

i pCi/ gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission

! evaluation of the circumstances prior to reaching the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit.

> Reducino T to < 500*F prevents the release of activity should a steam generator ^dbe rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on todine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

l l 3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with-stand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by nomal load transients, reactor trips, and startup and shutdown operations. The various categories m .

ARKANSAS - UNIT 2 B 3/4 4-5

- , - - , , - . _ ,, lQ - . , . . - - , .

1 1

1 PLANT SYSTEMS

(

i s

l BASES i

U =

maximum number of inoperable safety valves per

.j operating steam line 4

125 =

Power Level-High Trip Setpoint for two loop operation i

1 =

Power Level-High Trip Setpoint for single loop 1 operation with two reactor coolant pumos operating in the same loop i e

}h X = Total relieving capacity of all safety valves per 4 steam line in lbs/ hour (7,399,580 lbs/hr) 4 Y =

Maximum relieving capacity of any one safety valve il in lbs/ hour (1,508,360 lb3/hr) 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEM

, The OPERABILITY of the emergency feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 F from I

normal operating conditions in the event of a total loss of off-site power.

Each emergency feedwater pump is capable of deliveringswfficien}2 t teL-feedwater flow v. 455 3,.. 2. a p e;;.. : ;' '1:: ;;i: t: s e +-'--a d

of tne amem. 9c:.;a_.2. ~'; ::;;;4 t;  : tr##i-4,

• to ensure that i adecuate feedwater flow is available to remove decay heat and reduce tne Reactor Ccolant System tem;erature to less than 350'F when the shutdown cooling system may be placed into operatica.

4 3/4.7.1.3 CONDENSATE STCRAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS a t HOT STANDBY conditions for one hour with steam discharge to atmosphere with concurrent with total loss of off-site :wer. The con-tained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

J 1

~ . -

il ARKANSAS - UNIT 2 B 3/4 7-2 1

l w .-. - -.

1

!O i

! PLANT SYSTEMS I

1 -*tr.q dn.

4 4

3/. .

,/.1.4. -. C.it. ,/ . i.,f

.)

4 The 11m1:ations on secondary syster speci,.1c activity ensure nat i

U tne resultant off-site raciation dose aili be limited c a snail fraction 9 of 10 C.:R Par 100 Timits in :ne even of a steam line ructure. This dose also includes the effects of a coinciden:'TTO-s%' crimary :c secondary

• "." e l e = 'r. i. r. *.n a_ 2- . a. .a . . 3 e n = . = . ^ r v^ #. .a. 2. #. #. a. . *. u. e. e..=r ' #. . a. .ar.

• 2 J concurrent lors of cffsite electrica,i ;cwer. ines values are ccnsisten; with the assump:icns used in the acciden; analyses.

ICO&rD I

a

, 3/4.7.l.cs va?

.ata i' CT.

. rr.v. a t * *i i^

T. .cLon . ,.m. iin r

• t. ". r_ .q 1

or :ne main steam 1 scla:1cn valves ensures ine Or:re,_7,ii s..

• ..h .= .*. .o m. o ra. .*. . =. r. e o e e. . = = . . ,a.c. c.r.=.. r '.'.'i ".'. u a~~

. . m eon i.r. ...a_ a v c_ a. . *. . #. =.

taule. e. ,_2m. . r . ,. . a. .- ,. is e. a. ,.a;.<. a.s. . s ,s, - i. n. '. a .- =. . . a.

4 s ,o e _, ... ,i ; e. a_ ,,

, -.. .a r e .

p o s i .4. ". a. ra = c. .i v 4. .,v e #. #. ae. .s c #. .*..a. :. a.. =. .'.o . r..^ o l = . . . <.. v e. .c.m c . i. d. wn .

associa:ec witn the bicwdown, anc 2) limi: the pressure rise oithin i

J-containment in the event the steam line ructure occurs within contain-

, O ment. The OPERASILITY of :ne main steam isolation valves witnin ne closure times of the surveillance recuirerents are ccnsisten sitn the assumptions usec in :ne acciden analyses.

1 i

3 / , . 7.1. c- S tC C.

aun .-.RY ,a,n. i:.R C.,..1 5,,.d- i -..

.t i

n' *.s*.

. nrn. e, r.= ... n '.1 1 ". a. c. ~. r. . " . . =. d. .a ~ ~ r i. ~a c. 2 .- . ru x 4. . .= .=. ' y .b. .s. # 4.. r e. . ~-

I

• month 2 o #. c . a. r.= '.'. co a .= '. a. r. i a. i . 4. a l + ".. 4. . 4. .= l '. *.y-

• o a. s '. .= .+ 1 i s ". .. . c. ~-"..r4.2

. . . .. .. c.

l i ...i . 2 r. .". . a. 2 a. . n .. s.= r ,v w a. *.a. r. . ". . = - . i s . r," . .= - 2. o. .a. r s 2. . . d. r. a. . =. . c. r. - i v =. . .a.

acprocriate frecuencies for monitoring -hese carameters. The resuits cf this tes crceram will ce submitted :a :ne Ccamission cr review. . . .

.ne uommission .:

will :n.en issue a revision :D :nis scecifica:1cn scec1?ying .

> the ,ilmits on :ne cnemistry parameters ang e :ne trequencies icr ment:cring these carameters.

o,.

inis test program will include an analysis One_cnemican . con .. ,

..ine analysis sna,si 1centity stitutents of :n= condenser cooling water.

the various tra es of ions which ucon concentration in tr.e condenser may have the potential fer inducement for stress corrosion in the steen generator tubing. The tes program snail analyce ccacentration ;nentmena i

and the concentration rates in the steam generator and One secondary water system.

j 1(:::)

1 ARKANSAS - UNIT 2 B 3/4 7-3 N

- ~ ~ - - -

9- -..,m._ -. -- , _ , - - - _ , ..y- --,,__7---

o

_. _ = -

1

( REFUELING CPERAT10NS 'v' BASES i

3/4.9.5 C0fmuNICATIONS l The requirement for comunications capability ensures that refueling 8 station personnel can be promotly informed of significant changes in the J facility status or core reactivity condition during CORE ALTERATIONS.

3/4.9.6 REFUELING MACHINE OPERASILITY l The OPERABILITY requirements for the refueling machine ensureg that

1) the refueling machine will be used for movement of CEAs entfuel assemblies and that it has sufficient load cacacity to lift ah i fuel assembly, and 2) the core internals and pressure vessel are pro-tected from excessive lif ting force in the event they are inadvertently engaged during lifting operations.

1 i 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the ncminal i.s weight of a fuel assembly, CEA and associated handling tool over other fuel assemblies in :ne storage pool ensures that in the event tnis load is drc; ped (1) the activity release will be limited to that contained in i 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.

3/4.9.6 COOLANT CIRCULATION The requirement that at least one shutdcwn cooling loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor oressure vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core tr> minimi:e the effects of a boron dilution incident and prevent boren scratification.

A  :

ARKANSAS - UNIT 2' B 3/4 9-2

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

i 1

L i

DESIGil FEATURES VOLUME

5.4.2 The total water and steam volume of the reactor coolant system is 10,295 + 400 cubic feet at a nominal T avg of 545 F.

-U 6.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The metecrological tcwer shall be located as shown on Figure 5.1-1.

5.3 FUEL STCiRAGE CRITICALITY - SPENT FUEL 5 6. I. )

T9:4. The spent fuel storage racks are designed and shall be maintained with a ncminal 12.8 inch center-to-center distance between fuel assemblies having a maximum enrichment of 4.3 weignt percent U-235 placed in the i

storage racks to ensure a k,,f ecuivalent to < 0.95 when ficoded with I unborated water. The k,,, Of 1 0.95 includes a conservative allcwance 4 of 1.7% .k/k for uncertiinties as described in Section 9.1.2.3 of the FSAR. In addition, fuel in th'e storage pool snall have a U-235 loading of 1 47.8 grams of U-235 ::er axial centimeter of fuel assembly.

CRITICALITY - NEW FUE'.,

, I 4. I .'1.

M The new fuel stcrage racks are designed and shall be :aintained I

with a nominal 25.0 incn center-to-center distance between new fuel assembliessuchthatK,{,willnotexceed0.98whenfuelhavingamaximum enrichment of 3.7 weign ' 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 w$ter. Ice calculated K*"- incluces a conserva-

) tive allowance of 1.0". 2.k/k for uncertainties.

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

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

M 5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components icentified 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

~F - m -

.c . - -._ , . . . _ .

13.0 References 13.1 Section 1.0 References (1-1) " Final Safety Analysis Recort", Arkansas Nuclear One Unit 2, Arkansas Power & Light Ccmpany, Docket No. 50-368.

13.2 Section 2.0 References (2-1) Arkansas Pcwer & Light Company, Arkansas Nuclear One-Unit 2 Startup Test Report to NRC, NFP-6, Decket rio. 50-368, July 31, 1979.

13.3 Section 3.0 References (3-1) CENPD-256-P-A, " Test Fuel Rod Irradiaticn, 16 X 16 Nuclear Reactor", August 1977.

(3-2) CEN-50-P, Rev. 1-A, "Burnacle Poison Irradiatien Test",

August 1977.

13.4 Section 4.0 References (4-1) " Final Safety Analysis Report", Arkansas Nuclear One Unit 2, Arkansas Power & Light Company, Occket Nc. 50-368.

(4-2) CENPD-187, "CEPAN Methcc of Analyzing Creep Collapse cf Oval Cladding", June 1975.

(4-3) CEN-96-(A)-P, Rev. 1, "ANO-2 Reacter Operation with Modified CEA Guide Tubes and Lengthened Upper Guide Structure Flow Channels", July 12, 1978.

(4-4) CENPD-139-P-A, "C-E Fuel Evaluation Model Tooical Report",

July 1, 1974 (4-5) NUREG-0418, " Fission Gas Release frcm Fuel at Hign Surnup",

March 1978.

13-1

13.5 Secten 5.0 References (5-1) CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report",

July ~, 1974 (5-2) CENPD-153-P, Rev. 1-P-A, " INCA /CECOR Pcaer Peaking Uncertainty", May 1980.

(5-3) W. R. Caldweil, "PDO-7 Reference Manual", WAPD-TM,-678, January 1967.

(5 4) T. Ober, et al., " Theory, Capabilities and Use of the Three-Dimensional Reactor Operation and Control 'imulatcr (ROCS)", Nucl. Sci. and Eng., 54 (605), 197'.

(5-5) Letter, A. E. Lunavall, Jr. (BG&E) to R. W. Reic (NRC),

February 23, 1979.

(5-6) Safety Evaluation Report, Calvert Cliffs Nuclear Power Plant, Unit No. I, Docket No. 50-317, June 14, 1979.

(5-7) Letter, Robert E. Uhrig, (FP&L) to Victor Stello (NRC),

February 22, 1979, "St. Lucie Unit 1, Decket he. 50-335, Proposed Amendment to F4cility Operating License DPR-67.

(5-8) Safety Evaluation Repert, Florida Power & Light, St. Lucie Unit No. I, Decket No. 50-335, May 27, 1979.

(5-9) Letter, Robert W. Reid, (NRC) to A. E. Scherer (C-E),

Decemoer 17, 1979.

(5-10) A. Jcnsson, et. al., " Discrete Integral Transport Theory Extended to the Case with Surface Sources",

Atomkermenergie, Bd. 24, 1974 13-2

13.5 Section 5 References (Continued)

(5-11) A. Jonsson, et al., " Verification of a Fuel Assembly Spectrum Code Based on Integral Transport Theory", Trans.

Am. Nucl. Soc., 28 (778), 1978.

(5-12) Baltimore Gas & Electric Ccmpany, Docket No. 50-317 "Calvert Cliffs Unit No. 1, Cycle 2 Amendment to Facility Operating License", March 14, 1977.

(5-13) Omaha Public Power District, Docket No. 50-285 " Fort Calhoun Station Ui. i t No. 1, Cycle 5 Amendment to Facility Operating License", August 3, 1978.

13.6 Section 6.0 References (6-1) CEr?O-161-P, " TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core", July 1975.

(6-2) CENPD-162-P-A, " Critical Heat Flux Correlation fcr C-E Fuel Assemblies with Stancarc Spacer Grids Part 1, Uniform Axial Power Distributien", April 1975.

(6-3) CENPD-206-P, " TORC Coce, Verification and Simplified Modeling Methods," January 1977.

(6-4) CEN-143(A)-P "CPC/CEAC Software Modificrtions for Arkansas Nuclear One - Unit 2", December 1980, Appendix A.

(6-5) CEN-123(F)-P, " Statistical Combination of Uncertainties Part 2", February 1980.

l (6-6) CEN-124(B)-P, " Statistical Combination of Uncertainties

! Part 2" March 1980.

13-3

(6-7) CEN-139(A)-P, " Statistical Combination of Uncertainties; Combination of System Parameter Uncertainties in Thermal Margin Analyses fer Arkansas Nuclear One Unit 2", November 1980.

(6-8) Supplement 3-P (Proprietary) to CENPD 225P, " Fuel and Poison Rod Bowing", June 1979.

(6-9) Letter, D. B. Vassalo (flRC) to A. E. Scherer (C-E), datec June 12, 1978, 13.7 Section 7.0 References To be Supplied later with Section 7.0 (7-1) CENPD-135-P, "STRIKIN II, A Cylindrical Gecmetry Fuel Rod Heat Transfer Program," August 1974 (7-2) CENPD-190-A, "CEA Ejection, C-E Method for Control Element Assembly Ejection," July 1976.

(7-3) GEMP-482, H. C. Brassfield, et al., " Recommended Property and Reactor Kinetics Data for Use in Evaluating a Light Wat.er-Cooled Reactor Loss-of-Coolant Incident Involving Zircalvy-4 or 304-55, clad 002 ," April 1968.

(7-4) Idaho Nuclear Corporation, Monthly Report, NY-123-69, October 1969.

(7-5) Idaho Nuclear Corpcration, Monthly Report, HAI-127-70, March 1970.

(7-6) "CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply Sy tem," CENPD-107, April 1974, Proprietary Information.

(7-7) "ATWS Model Modifications to CESEC," CENPD-107, Supplement 1, September 1974, Proprietary Information.

13 4 t . .

~. .

(7-8) "AWTS Mocels Modification to CESEC," CENPD-107, Supplement 1, Amendment 1-P, November 1975, Proprietary Information.

(

(7-9) "ATWS Model for Reactivity Feedback and Effect of Pressure on Fuel," CENPD-107, Supplement 2, Sectember 1974, Proprietary Information.

(7-10) "ATWS Model Modifications to CESEC," CENPD-107, Supplement 3, August 1975.

(7-11) "ATWS Model Modificaticns to CESEC," CENPD -107, Supplement 4-P, December 1975, Proprietary Infccmation.

(7-12) "CPC/CEAC Software Modifications for Arkansas Nuclear One-Unit 2," CEN-143(A)-P, December 1980.

(7-13) " Final Safety Analysis Repcrt," Arkansas Nuclear One Unit 2, Arkansas Power & Lignt Company, Docket No. 50-368.

13.8 Section 8.0 References (8-1) " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors", Federal Register, Vol. 39, No. 3, Friday, January 4, 1974 (8-2) CENPD-132, " Calculative Methods for the C-E Large Break LOCA Evaluation Model," August 1974 (Proprietary).

CENPD-132, Supplement 1, " Updated Calculative Methods for the C-E Large Break LOCA Evaluation Mcdel," December 1974 (Proprietary).

CENPD-132, Supplement 2, " Calculational Methods for the C-E Large Break LOCA Evaluation Model", July 1975 (Prcprietary).

(8-3) CENPD-135, "STRIKIN-II, A Cylindrical Gecmetry Fuel Rod Heat Transfer Program," April 1974 (Proprietary)

CENPD-135, Supplement 2, "STRIKIN-II, A Cylindrical 13-5

Ge: retry Fue1 R:: :-e at Transfer r:cra (v :ificatien),"

re c ru a ryre: ,a .,,- (crc:rietary)

CENFO-135, Su::lerent 4, "ST:. KIN-::, 1 Cylincrical '

Ge retry Fusi Rc Heat Transfer Pr: gram," au gus: 1975 (Prc:rie:ary)

CENPD-135, Su::lemen 5, "$TRIKIN-!!, A Cylincrica' Ge retry Fuel Rc Heat Transfer Pr: gram", A:ril 1977 (Proprietary).

(. 4)

u. .'; p r,.. - 1.0 9 , "C :. .." a. i .:v .21 " > +. . . r. te~ .4.'," .-l .v

.'. i. .: 7 (Pr::rietary)

(S-5) " Final Safety Analysis Re: et," Arkansas Nuclear Cne Uni: I 2, Arkansas 00 oer an Lign: C:r:any, Occ<e: N:. 50-355.

(3-5) CEN:D-132, " PARCH - A FC . T P  ?.- I '.' Cigital r:gra-  ::

Evaluate Pcci Boiling, Axial R:: ar.: C001an Heatu ,"

August 1974 (Pr::rietary).

r.d. *pn; 'us, -

1, a-r-m-c r - - - ---- "r '

suco1e ent -

r e'. s i e-.'. i s D i g t' *. a 1 Prcgra- to Evaluate ::c! 5ciling, 1.x l a l 4:c an C: clan:

Heatu:" (Vocificaricr4s), February 1975 (Pr::rietary) c ', e n., _ .l .. : ,.e-N : l e. ..c. r, *.. :,( .: ) , -::Rv"9

. . 1 .:^

t . -m::'? ."i . '. , i *. = 1 Pr:gra ic Evaluate P:01 Boiling, axial R: an: :::ian:

Featup" January 1977 (Proprietary).

13.9 Secticn 9.0 Peferences (9-1) CEN-39(A)-P, "CPC Pec:ecticn Algorithm 5 ftware Cnange Prececure," Revision 2, CeceTeer 21, 1978.

(9-2) CEN-39(A) P, "CPC Protecticn Alg:ritnm Soft are Cnange Precedure Supple ent," Su:pleren: 1-P, Revisien 1, January 5, 1979.

(9-3) CEN-143(A)-P, "CPC/CEAC Scftware vocificaticns f:r *r<ansas Nuclear One - Unit 2," Cecember 19SO.

13-5

(9-4) CENPD-170-P, " Assessment of the Accuracy of PWR Safety System Actuation as Performed by the Core Protection Calculators (CPC)," July 1975, and Supplement 1-P, November 1975.

,10 Section 10.0 References (10-1) CENPD-384, "The Evaluation and Demonstration of Methods for Improved Nuclear Fuel Utilizaticn," First Semi-Annual Progress Report: January-June 1980, for the Department of Energy, October 1980, to be published.

\

13-7