ML20098F133

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Proposed Tech Spec Changes Re Cycle 7 Reload
ML20098F133
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 09/26/1984
From:
ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML20098F132 List:
References
NUDOCS 8410020363
Download: ML20098F133 (21)
v • d • e


Text

.

LIST OF FIGURES Number Title Page 2.1-1 CORE PROTECTION SAFETY LIMIT 9a 2.1-2 CORE PROTECTION SAFETY LIMITS 9b 2.1-3 . CORE PROTECTION SAFETY LIMITS 9c 2.3-1 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINT 14a 2.3-2 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS 14b 3.1.2-1 REACTOR COOLANT SYSTEM HEATUP AND C00LDOWN LIMITATIONS 20t 3.1.2-2 REACTOR COOLANT SYSTEM NORMAL OPERATION-HEATUP LIMITATIONS 20b 3.1.2-3 REACTOR' COOLANT SYSTEM, NORMAL OPERATION C00LDOWN LIMITATIONS 20c 3.1.9-1 LIMITING PRESSURE VS. TEMPERATURE FOR CONTROL R0D DRIVE OPERATION WITH 100 STD CC/ LITER H2O 33 3.2-1 BORIC ACID ADDITION TANK VOLUME AND CONCENTRATION VS. RCS AVERAGE TEMPERATURE 35a 3.5.2-1 ROD POSITION LIMITS FOR FOUR-PUMP OPERATION FROM 0 EFPD TO EOC - ANO-1 48b 3.5.2-1B DELETED 3.5.2-1C DELETED 3.5.2-1D DELETED 3.5.2-2A R0D POSITION LIMITS FOR THREE-PUMP OPERATION FROM 0 EFPD TO E0C - ANO-1 48c

  • 3.5.2-28 ROD POSITION LIMITS FOR TWO-PUMP OPERATION FROM 0 EFPD TO EOC - ANO-1 48d 3.5.2-2C DELETED 3.5.2-2D DELETED
  • Pages 48c1 through 48c7 were deleted.

, iv Amendment No. 52, 71 m!!!akSMjh P

3.5.2-2E DELETED 3.5.2-2F DELETED

~3.5.2-2G DELETED 3.5.2-2H DELETED 3.5.2-3 OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 EFPD 10 EOC - ANO-1 48e 3.5.2-38 DELETED 3.5.2-3C DELETED 3.5.2-3D DELETED 3.5.2-4 LOCA LIMITED MAXIMUM ALLOWABLE LINEAR HEAT RATE 48f 3.~5.2-4A APSR POSITION LIMITS FOR OPERATION FROM 0 EFPD TO APSR WITHDRAWAL - ANO-1 48g 3.5.2-48 APSR POSITION LIMITS FOR OPERATION AFTER APSR WITHDRAWAL ANO-l' 48h 3.5.2-4C DELETED 3.5.2-4D DELETED

'3.5.4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE INDICATION 53a 3.5.4-2 INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TILT INDICATION 53b 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c 4.4.2-1 -NORMALIZED LIFT 0FF FORCE - HOOP TENDONS 85b t.

4.4.2-2 NORMALIZED LIFT 0FF FORCE - DOME 1ENDONS 85c 4.4.2-3 NORMALIZED LIFT 0FF FORCE - VERTICAL TENDONS 85d 6.2-1 MANAGEMENT ORGANIZATION CHART 119 6.2-2 FUNCTIONAL ORGANIZATION FOR PLANT OPERATION 120 v Amendment No. 52, 80

l DNBR of 1.'3 corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been. considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop' was assumed in reducing the pressure trip set points to correspond to the elevated location where the pressure was actually-measured.

The curve presented in Figure 2.1-1 represents the conditions at which a l minimum DNBR greater than 1.3 is predicted. The curve is the most

.estrictive combination of 3 and 4 pump curves, and is based upon the maximum possible thermal power at 106.5% design flow per applicable pump status. This curve is based on the following nuclear power peaking factors (2) with potential fuel densification effects:

F" = 2.83; F = 1J1; = 1.65.

q H The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification:

1. Tge 1.3 DNBR limit produced by a nuclear power peaking factor of F = 2.83 or the combination of the radial peak, axial peak and p8sitionoftheaxialpeakthatyieldsnolessthan1.3DNBR.
2. The combination of radial and axial peak that prevents central [

fuel melting at the hot spot. The limit is 20.5 kW/ft. I Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.

The flow rates for curves 1, 2, and 3 of Figure 2.1-3 correspond to the c:gected einimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump maximum thermal power combinations shown in Figure 2.1-3. The curves of Figure 2.1-3 represent the conditions at which a minimum DNBR greater than 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation. The local quality at the point of minimum DNBR is less than 22 percent (1).

Amendment No. 21, 52 8

l 1

Using a local quality limit of 22 percent at the point of minimum DNBR as a basis for curve 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.

The DNBR as calculated by the BAW-2 correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.

-The maximum thermal power, as a function of reactor coolant pump operation is limited by the power level trip produced by the flux-flow ratio (percent flow x flux-flow ratio), plus the appropriate calibration and instrumentation errors.

For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation. Curves 1 and 2 of Figure 2.1-3 are the most restrictive because any pressure-temperature point above and to the left of this curve will be above and to the left of the other curve.

REFERENCES (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May 1976.

(2) .FSAR, Section 3.2.3.1.1.c Amendment No.- 21, 31, #3, 52, 57 9

i

)

l 1

j Core Protection Safety Limits - ANO-1 Figure 2.1-2 Thermal Power Level,

- 120

(-31,112) (27.7,112) i

(-42,102.1 1 AccEP m LE  :

l 4 PL"4P l l l OPE?aTIO:: -- 100 l l

[(-31,90.6) (27.7, 90,6f g (35.9, 96.4)

I '

('-42,89.7)I l ACCEPTABLE I '

I 4 .uD 3 -- 80 l 1 l hp[._7g).s l (35.9,75) i I [(-31,63.3) (27.7.63.3)!

l . T l

l

- - 60 I l ACCEPTABLE l

(-42,53.4)* g 4, 3, x D 2 l g

[ PL".4P l l OPEP.ATIO:: l g(35.9,47.7)

I 40 l

g l l l 1 I i I i

l l I I i 1 I i I -

- 20 I j l l 1 1

I I I I I i i i i I I I i t t I f I  ! II I

. -60 -40 -20 0 20 40 60 Reactor Poyer Imbalance, %

Amendment No. 5, 22, 22, #3, 52, 57, 7I 9b

l Core Protection Safety Limits - ANO .1 Figure 2.1-3 2600 2400 E

f 2200 , ,

b 1 th-V%

j 2000 3

o b

c.2 3

1800 , f 1600 560 550 600 620 640 660 Reactor Outlet Temperature, 'F CURVE RPM POWER PUMPS OPERATlHG (TYPE OF LIMIT) 1 374,880 (1005)* 112% FOUR PUNPS (DNER LIMIT) 2 280,035 (74.75) 90.6% THREE PUMPS (DNBR LIMIT) l 3 184,441 (49.25) 64.1% ONE PUNP IN EACH LOOP (0UALITY LIMIT)

  • 106.55 OF DESIGN FLOW l

l i Amendment No. 21 9c

The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are as follows:

1. Trip would occur when four eactor coolant pumps are operating if power is 107 percent and reactor flow rate is 100 percent or flow rate is E3.5 percent and power level is 100 percent.
2. Trip would occur when three reattor coolant pumps are operating if power is 80 percent and reactor flow rate is 74.7 percent or flow rate is 70 pe cent and power level is 75 percent.
3. Trip would occur when one reactor coolant pump is operating in each loop (total of two puops operating) if the power is 52 percent and reactor flow is ^9.2 percent or flow rate is 45.8 percent and the power level !s 49 percent.

The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

No pe621ty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage. For safety analysis calculations the n,aximum calibration and instrumentation errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits. The reactor power imbalance (power in top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power level

[ trip associated reactor power-to-reactor power imbalance boundaries by 1.07

i. percent for a 1 percent flow reduction.
8. Pump Monitors In conjunction with the power inbalance/ flow trip, the pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant Amendment No. 22, 32, #3, 52, 67 12

_. __ _ ___ _. ___ . _ ._ _. .,_ __ , _. __. ~ _ . . _

Protective System Maximum Allowable Setpoints ANO-1, Figure 2.3-2 Thermal Power Level, % FP

- 120

(-12,.107) _ (12, 107)

-100!

l ACCEI TABLE l

(-28'913 I 4 Ptn P l I OPERi IION I l 1 l (22, 84.13)

[ (-12,80) (12,i30)

-80 l I ACCETTABLE I I

[ l 3 ANE 4 l 1 I I PtDIP l

(-28, 64) l lOPERITION l t I

i 1

-- 60 l 1 (22, 57.13)

[

l(-12,5:p l 1(12,852)

I 1 l

1 I I i .

- 0

(-28, 36) ACC" Tr 7 l 2, 31, AND 4 , ,

I PIRIPl I (22, 29.13) -

I OPERhTION l g I I I l

l l- -

- 20 l l l l ,

1 l l I 1 I I I I I.

l 1 I 1 l , t! I l

-60 -40 -20 0 20 40 60 Reactor Power Imbalance, %

Amendment No. 5, 2Z, 3Z, #3, 32, 67, 71 14b i .:.

Tcblo 2.3-1 Ructer Pratmetirn System Trip Sstting Limits Four RC Pumps Operating Three RC Pumps Operating One RC Pump Operating (Nominal Operating (Nominal Operating in Each Loop (Nominal) Shutdown Power - 100%) Power - 75%) Operating Pcuer - 49% Bypass Nucicar power, % of ' .104.9 104.9 104.9 5.0" rct*_d, max NuclgarPowerbasedon 1.07 times flow minus 1.07 times flow minus 1.07 times flow minus Bypassed flow and imbalance, reduction due to reduction due to reduction due to

% of rated, max imbalance (s) imbalance (s) imbalance (s)

Nuclear Power based on NA NA 55 Bypassed pump monitops, % of rzttd, max High RC system 2300 8 2300 2300 1720 pressure, psig, max Lcw RC system 1800 1800 1800 Bypassed prassure, psig, min d d d Varitble low RC 11.75 T

-5103 11.75 T -5103 11.75 T -5103 Bypassed system pressure,.

psig, min RC temp, F, max 618 618 618 618 High reactor building 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 prassure, psig, max psia) a Automatically set when other segments of the RPS (as specified) are bypassed.

Rractor coolant system flow.

c Thn pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two pump operation.

d T is given in degrees Fahrenheit (F).

ut Amendment No. 2, 2Z, A3, A9, 52, 57 15

/

Boric Acid Addition Tank Volume and Concentration Vs RCS Average Temperature - ANO-1 Figure 3.2-1 GPERATIOT1 ABOVE AtID TO THE LEFT 6000 0F THE CURVES IS ACCEPTABLE

/

5000 - 8700 PPM J

S

/

S- 4000 - 9500 PPM--

/

o-l

'E 3000 - 10,000 PPM

.e

.t E

12,000 PPM 2000 -

T3 ~

.S

=

b 1000 -

0- ' ' ' '

200 300 400 500 600 700 RCS Average Temperature, F TEMP. F REQUIRED VOLuiiE, GAL.

8700 PPM 9500 PPM 10,000 PPM 12,000 PPM 579 6436 5863 5554 4589 532 5289 4817 4564 3769 500 4488 4087 3872 3199 400 2434 2218 2101 1737 300 986 898 851 705 200 0 0 0 0 Amendment No. 33, #3, 52, 68, 71 35a

- + -

L

6. If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2 operation above 60 percent of the thermal power allowable for the reactor coolant pump combination may continue provided the rods in the group are positioned such that the rod that was declared inoperable is contained within allowable group average position limits of Specification 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.3.

3.5.2.3 ..The worth of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the Control Rod Position Limits defend in Specification 3.5.2.5.

3.5.2.4 Quadrant tilt:

1. Except for physics tests, if quadrant tilt exceeds 3.1% power shall be' reduced immediately to below the power level cutoff (92% FP). Moreover, the power level cutoff value shall be reduced 2% for each 1% tilt in excess of 3.1%. For less than 4 pump operation, thermal power shall be reduced 2% of the thermal power allowable for the reactor coolant pump combination for each 1% tilt in excess of 3.1%.
2. Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be reduced to less than 3.1% except for physics tests, or the following adjustments in setpoints and limits shall be made:
a. The protection system maximum allowable setpoints (Figure 2.3-2) shall be reduced 2% in power.for each 1%

tilt.

b. The control rod group and APSR withdrawal limits shall be reduced 2% in power for each 1% tilt in excess of 3.1%.
c. The operational imbalance limits shall be reduced 2% in power for each 1% tilt in excess of 3.1%.
3. If quadrant tilt is in excess of 25%, except for physics
- tests or diagnostic testing, the reactor will be placed in the hot shetdown condition. Diagnostic testing during power
. operation with a quadrant power tilt is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3.5.2.4.1 above.

4'. Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15% of rated power.

Amendment No. 6, fl, 31, #3 47 I

I l

,3.5.'2.5 Control rod positions:

1. Technical Specification 3.1.3.5 (safety rod withdrawal) does

, not prohibit the exercising of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.

2. Operating rod group overlap shall be 20% 15 between two sequential groups, except for physics tests.

J

![

6 j.

4 5

Amendment No. 47a i

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

G s

Except for physics tests or exercising control rods,_(a) the

~

3.

control rod withdrawal limits are specified on Figures 3.5.2-1, 3;5.2-2A-and 3.5.2-28 for 4, 3 and 2 pump operation respect 1vely; and (b) the axial power shaping control rod

, _ withdrawal limits are specified on Figures 3.5.2-4A and 3.5.2-48. .Itl any of these control' rod position . limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4. Except for physics tests, power shall not be increased above the power level cut-off of 92% of'the maximum allowable power level unless one of the following conditions is satisfied:
a. - Xenon reactivity is within 10% of the equilibrium value for operation at the maximum allowable power level and asymptotically approaching stability,
b. Except.for. xenon free startup, when 3.5.2.5.4a applies, 60 the reactor has operated within a range of 87 to 92% of the maximum allowable power for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 40% rated power.

Except for physics tests, imbalance shall be maintained within the envelope defined by_ Figure 3.5.2-3. If the imbalance is not within the envelope defined by Figure 3.5.2-3. corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reactor power shall be reduced until imbalance limits are met.

~

3.5.2.7 The control rod drive patch' panels shall be locked at all times

. :with limited access to be authorized by_the Superintendent.

Bases The power-imbalance envelope defined in Figure 3.5.2-3 is based onL(1) LOCA analyses which.have defined the maximum linear heat rate (see Figure 3.5.2-4), such that the maximum cladding temperature will not exceed the Final Acceptance Criteria and (2) the Protective System Maximum Allowable Setpoints (Figure 2.3-2). Corrective measures will be taken immediately-ahould the indicated quadrant tilt, rod position, or imbalance be outside i their specified boundaries. Operation in a situation that would cause the Final Acc-ptance Criteria to be approached should a LOCA occur is highly

- improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at.their limitsswhile l

Amendment No.'6, 22, JZ, #3, 52 48 i

E- *"y'-* yd- 4t w yy- y-p,yyv- ypwy -wy -y'yw, yi y _,-pga ryrwyisiW

The quadrant power. tilt limits set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using the. definition of

. quadrant power tilt given in Technical Specifications, Section 1.6. These limits, in conjunction with the control rod position limits in Specification 3.5.2.5.3, ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel

.densification.

The quadrant tilt and axial imbalance limits in Specifications 3.5.2.4 and 3.5.2.6, respectively, apply when using the plant computer to monitor the limits. The 2-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service. Additional uncertainty is applied to the limits when other monitoring methods are used.

During the physics testing program, the high flux trip setpoints are administratively set as follows to ensure that an additional safety margin

!4 provided.

Amendment No. 52 48a1

Rod Position '.imits for 4-Pump Operation From 0 EFPD TO EOC ---- ANO-1 Figure 3.5.2-1 110 100 - (231 2' 102I G(

(300, 102) 90 (27540) 80 -

(265,78

I y 70 - SHUTCCW?I MARGIll LIMIT

=

y 60 -

OPEP.ATIO!!

- RESTRICTED 50 CPEP.AT:Cn !:1 (158.5,28)g (230, 45)

. THIS REGIC:1 IS 40 t - NOT ALLCWED 5

30 -

PEF.M!SSIELE OPEP.UING 20 -

REGIC3 10 0

'd 20 40 60 80 100 120 140 160 180 200 220 240 E50 2:0 JC0 0 20 40 60 30 100 i e i i GRCUP 7 0 20 *0

. 60 80 100 t . f f f f GROUP 6 0 20 40 60, 80 100 GROUP 5 Rod index, ; WO P

Amendment No. 5, 22, (Z, #3, 52, 7I 48b-l

_ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _z _ _ ' _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _

d Rod Position Limits for 3-Pump Operation From 0 EFPD TO E0C ---- ANO-1 Figure 3.5.2-2A 110 '

100 -

90 -

80 -

(233, 77) - (280.3,77 (300,77) 5

, ' 70 - SHUTCCWN MARGIN LIMIT l (270,67' y 60 -

(250,58) 50 -

OPEPATICH IN OPERATIO!!

. THIS REGION 15 RESTRICTED 40

~h - NOT ALLOWED 5 (158.5,36) (230,35) 30 -

0 -

PEP.MISSIBLE 10 -

(81.5,9) OPERATIfiG REGIO!!

Q y, ' '

f  ? f f f f  ?  !  !  !  ? e 0 20 40 60 80 100 120 140 160 180 200 220 240 250 230 300 0 20 40 60 80 100 f  ! I f f f

GRCUP 7.

0 20 40 60 80 100 t f f f f f GROUP 6 0 20 80 100 i r 40 60, , ,

GROUP 5 Rod Index. WO Amendment No. 2Z, 31, 42, 52, 72 48c

R00 POSITION LIMITS FOR 2 - PUMP OPERATION FROM 0 EFPD TO E0C ---- ANO-1 Figure 3.5.2-2B 110 100 -

90 -

80 -

y 70 -

es 3

~

60 -

',;; ' 50 -

OPEPATION IN

( ' )^ '

= -

=

(200' :2) 40 THIS REGICN IS SHUTD0'Ali GPERATIO;l (275, 45) b - NOT ALLOWED MARGift RESTRICTED (265,c3) i LI:4IT 30 (158.5,26) 20 . (235,25)

OPERATitiG 10 - REGI0tl o[IU'UI , , (81.),8.,5), , , , , , , ,

0 20 40 60 80 100 120 140 160 180 2C0 220 240 250 220 JC0 0 20 4,0 60 80 1C0 GROUP 7 0 20 . 40 60 80 100 t f f f f 1 GROUP 6 0 2,0 4,0 60, 80 100 t , ,

. GROUP 5 Rcd Index. : WO Amendment No. 14, JZ, #3, 52, 71 48d

, -, -,e -- -., .e - , , . 4,g --s,.e- ,-er _.,,a.m

<-a-, y q - -nre - - , , , . , , -

,,p.,--v,- w-, - - , - , ,

Operational Power Imbalance Fnvelope for Operation From 0 EFPD TO EOC EFPD ---- ANO-1 Figure 3.5.2-3

.. 110

- iw])r (9, 102)

(-15,102)Ns/ -

-(-15, 92' %

,_ J[)(10,92)

(-20,80) _-

Sq[) (10, 30)

__ 70 RESTRICTED RESTE!CTED REGION REGIC:t 60 t

(-20, 50'() ((__ 5c() (12, 50)

N 13 -

40 b- -

30 8

c.

20 10

(-20,0) (12,0)

' ' ' e irs . .

(3

~

-50 -40 -30 -20 -10 0 '0

. 20 30 40 50 Axial Power Imbalance f

Amendment No. 22, 52 48e 9

9

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

LOCA Limited Maximum '.110wat:1e Linerar Heat Rate Figure 3.5.2-4 d 21 k

a 20 -

J 19 -

c:

a g 18 -

=

~

h 17 :. Balance of g Cycle 3 16 -

2 j 15 -

First 1,000 "

o mwd /mtu 14 - .'

{

s i 13 -

R

2 12 ' ' ' ' '

0 2 4 6 8 10 12

~

Axial Location of Peak Power from Bottom of Core, ft l

Amendment No. A3, 52, 72 48f I

i f s I

APSR Position Limits for Operation Fron 0 EFPD to APSR Withdrawal ---- ANO-1 Figure 3.5.2-4A 110 (9.5.102) (35, 102) 100 -

q, ,_.TRIC..D r.o u.

aggIc;;

~

- p' ' ) (35,90) 90 <

80 -

7 (40,75)

E 70 g (0,70)

S o 60 PEFlilSSIBLE

    • OPEPAT!!!G

{ '

REGIO:1 e

40 - p 1 0, 40) 30 -

20 10 0

O 10 20 30 40 50 60 70 80 90 100

  • . Withdrawn .

Amendment No. 2Z, #2, 52, 72 489

- _ - _ _ _ _ - ____________________l____________ ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _

2 APSR Position Limits for Operation After APSR Withdrawal ---- ANO-1 Figure 3.5.2-4B 110 100 -

90 _

80 -

APSR INSERTION ':07 ALLO'.-lED et 70 -

IN THIS TIME INTERVAL E

60 -

e s 50 -

C g 40 -

30 -

20 -

10 0 , , , , , , , , ,

0 10 20 30 40 50 , 60 70 80 90 100

% Withdrawn ~

Amendment No. 3Z, 43, 52, 7Z 48h f

v. - e- ~ . . ~ , . - - , - , - - - - . - - ~ - - , - , , - , - - - - - - - - . . - - , , ,,n

kt BAW-1840 l

August 1984 l

l ARKANSAS NUCLEAR ONE, UNIT 1

- Cycle 7 Reload Report -

Babcock &Wilcox

  • a McDermott company

.- - - . .- .-. _- .- - . . _ . ..- . . _ . ._ - . . . - . - . . - . . . - . . _ _ . _ - .-

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