ML20206C465

Amend 113 to License DPR-51,permitting Operation of Facility for Cycle 9 & Modifying Variable Low Pressure Reactor Trip Setpoint

ML20206C465
Person / Time
Site:	Arkansas Nuclear  (DPR-51-A-113)
Issue date:	11/08/1988
From:	Calvo J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20206C469	List: ML20206C465 ML20206C471 ML20206F517
References
NUDOCS 8811160288
Download: ML20206C465 (34)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION o

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.I wAsHWG TON, D. C. 20066 s.,.....)

ARKANSAS POWER AND LIGHT COMPANY DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UN_IT 1 ANENDMENT TO FACILITY OPERATING LICENSE Amendment No.113 License No. DPR-51 n gulatory Comission (the Comission) has found that:

1.

The Nuclear e

A.

The applications for amendment by Arkansas Power and Light Company (the licensee) dated July 20 and August 31, 1988, comply with the i

standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; C.

The facility will operate in conformity with the applications, as arrended, the provisions of the Act, and the r,ules and regulations of the Comission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be

+

conducted in compliar.ce with the Comission's regulations; D.

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

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

G8111602G8 881109 ADOCK0".OOg,g3 DR 7

2 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No.

DPR-51 is hereby amended to read as follows:

2.

Techn,1c,al, Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.ll3, are hereby in orporated in the license. The licensee shall operate the faci'lity in accordance with the Technical Specifications.

3.

The license amendment is effective as,of its date of issuance.

FOR THE NUCLEAR REGULATORY C0 mlSS10N N

fev Jose A. Calvo, Director Project Directorate - IV Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Char.ges to the Technical Specifications Date of Issuance: November 8, 1988

0

.o ATTACHMENT TO LICENSE AMEi!DMENT f;0.113 FACILITY OPERATING LICENSE NO. OPR-51 DOCKET NO. 50,-3,13 Revise the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

REMOVEPAGE)

INSERTPAG_E}

iv iv v

v 7

7 8

8 9

9 9a 9a 9b 9b 9e 9c 11 11 12 12 13 13 14a 14a 14b 14b 15 15 47 47 48 48 48b 48b 48c 48c 48d 48d 48e 48e 4Sf 48f 48g 489 48h 48h 48i 481 48j 48j 48k 48k I

481 481 1

48m 48m 48n 48n 480 48p 48q 48r 48s 48t 48u 48v

, 114 114 115 115

80 6

LIST OF FIGURES Number Title Pace 2.1-1 COREPROTECTkONSAFETYLIMITS 9a 2.1-2 CORE PROTECTION SAFETY LIMITS 9b l

2.1-1 CORE PROTECTION SAFETY LIMITS 9c l

2.3-1 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINT 14a 2.3-2 PROTECTIVE SYSTEM MAXIMUM ALLOWADLE SETPOINTS 14b 3.1.2-1 REACTOR COOLANT SYSTEM HEATUP AND C00LDOWN LIMITATIONS 20a 3.1.2-2 REACTOR COOLANT SYSTEM NORMAL OPERATION-HEATUP 20b LIMITATIONS 3.1.2-3 REACTOR COOLANT SYSTEM, NORMAL OPERATION C00LDOWN 20e LIMITATIONS 3.1.9-1 LIMITING PRESSURE VS. TEMPERATURE FOR CONTROL ROD DRIVE 33 OPERATION WITH 100 STO CC/ LITER H O 2

3,2-1 BORIC ACID ADDITION TANK VOLUME AND CONCENTRATION VS. RCS 35a AVERAGE TEMPERATURE 3.5.2-1A ROD POSITION SETPOINTS FOR FOUR-PUMP OPERATION FROM 0 48b to 27+10/ 0 EFPD - ANO-1 CYCLE 9 3.5.2-1B R00 POSITION SETPOINTS FOR FOUR-PUMP OPERATION FROM 48c 27+10/-0 TO 360+50/-10 EFPD - ANO-1 CYCLE 9 3.5.2-1C.

ROD POSITION SETPOINTS FOR FOUR-PUMP OPERATION AFTER 48d 360+50/-10 EFPD - ANO-1 CYCLE 9 3.5.2-2A R00 POSITION SETPOINTS FOR THREE-PUMP OPERATION FROM 0 48e TO 27+10/-0 EFPD - ANO-1 CYCLE 9 3.5.2-2)

ROD POSITIDN SETPOINTS FOR THREE-PUMP OPERATION FROM 48f 27+10/-0 TO 160+50/-10 EFPD - ANO-1 CYCLE 9 3.5.2-2C ROD POSITION SETPOINTS FOR THREE-PUMP OPERATION AFTER 48g 360+50/-10 EFPD - ANO-1 I

ALendment No. 52, 72, 92, 185. 113 iv

3.5.2-3A ROD POSITION SETPOINTS FOR TWO-PUMP OPERATION FROM 0 48h TO 27+10/-0 EFPD - ANO-1 CYCLE 9 3.5.2-3B ROD POSITION SETPOINTS FOR TWO-PUMP OPERATION FROM 48i 27+10/-0 TO 360+50/-10 EFPD - ANO-1 CYCLE 9 3.5.2-3C R00 POSITION SETPOINTS FOR TV)-PUMP OPERATION AFTER 48j 360+50/-10 EFPD - ANO-1 CYCLE 9 3.5.2-4A OPERATIONAL POWER IMBALANCE SETPOINTS FOR OPERATION FROM 48k 0 TO 27+10/-0 EFPD - ANO-1 CYCLE 9 3.5.2-4B OPERATIONAL POWER IMBALANCE SETPOINTS FOR OPERATION FROM 481 27+10/-0 TO 360+50/-10 EFPD - ANO-1 CYCLE 9 3.5.2-4C OPERATIONAL POWER IMBALANCE SETPOINTS FOR OPERATION AFTER 40m 360+50/-10 EFPD - ANO CYCLE 9 3.5.2-5 LOCALIMITEDMAXIMUMALLOWABLELINEARNEATRATE 49n 3.5.4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE 53a INDICATION 3.5.4-2 IWCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TILT 53b INDICATION 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c 3.24-1 NYDROGEN LIMITS FOR ANO-1 WASTE GAS SYSTEM 110be 4.4.2-1 NORMALIZED LIFTOFF FORCE - HOOP TENDONS 85b 4.4.2-2 NORMALIZED LIFTOFF FORCE - DOME TENDONS 85c 4.4.2-3 NORMALIZED LIFTOFF FORCE - VERTICAL TENDONS 85d l

t i

Amendment No. 52, 88, si ff, NJ, v

97,Zil. 03

~

2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE 1

Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.

]

Objective i

To maintain the integrity of the fuel cladding.

l Specification i

2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the j

locus of points established in Figure 2.1-1.

If the actual i

pressure / temperature point is below and to the right of the j

pressure / temperature line the safety limit is exceeded.

J 2.1.2 The combination of reactor thermal power and reactor power imbalance (power in the top half of the core minus the power in the j

bottom half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points for the specified flow set forth in Figure 2.1-2.

If the actual-reactor-thermal power / reactor power-imbalance point is above the line for the specified flow, the safety limit is exceeded.

Bases To maintain the integrity of the fuel cladding and to prevent fission i

product release, it is necessary to prevent overheating of the cladding i

under normal operating conditions.

This is accomplished by operating l

'within the nucleate boiling regime of heat transfer, wherein the heat j

transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature.

The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling l

(DN8).

At this point there is a sharp reduction of the heat' transfer coefficient, which could result in high cladding temperatures and the l

l possibility of cladding failure.

Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature and pressure can be related to DNSthroughtheuseofacriticalheatflux(CHF) correlation.

The i

BAW-2(1) and BWC(2) correlations have been developed to predict DNB and the location of DNS for axially uniform and non-uniform heat flux distributions.

The BAW-2 correlation applies to Mark-B fuel and the BWC correlatten applies to Mark-8Z fuel.

The local DN8 ratio (DN8R), defined I

as the ratio of the heat flux that would cause DNB at a particular core i

location to the actual heat flux, is indicative of the margin to DNB.

The j

minimum value of the DNBR, during steady-state operation, normal j

operational transients, and anticipated transients is limited to 1.30 l

(BAW-2) and 1.18 (BWC).

l l

j Amendment Ne. 21, 113 7

I

A*DNBR of 1.30 (BAW-2) or 1.18 (BWC) corresponds to a 95 percent l

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 for the allowable RC pump combination has been considered in determining the core protection safety limits.

The curve presented in Figure 2.1-1 represents the conditions at which the l

DNBR is greater than or equal to the minimum allowable DNBR for the limiting combination of thermal power and number of operating reactor coolant pumps.

This curve is based on the following nuclear power peaking factors (3) with potential fuel densification effects:

l FN = 2.83; Fh=1.71;FN = 1. 65.

g 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:

N 1.

The DNBR limit produe;ed by a nuclear power peaking factor of F

= 2.83 or the combination of the radial peak, axial peak and 4 position of the axial peak that yields no less than the DNBR limit.

2.

The comoination of radial and axial peak that prevents central fuel melting at the hot spot.

The limit is 20.5 kW/ft.

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 expected minimum 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 curve; of Figure 2.1-3 represent tle conditions at which the DNBR limit is l

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 (BAW-2)G) or 26 percent (BWC)(2).

l r

e i

I Amendment No. 2I, 52, 52,113 8

Using a local quality limit of 22 percent (BAW-2) or 26 percent (BWC) at the point of minimum DNBR as a basis for curves 2 and 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 or the BWC correlation continually l

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 calibra",fon 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 1.30 (BAW-2) or 1.18 (BWC) or a local quality at the point of minimum DNBR less than 22 percent (BAW-2) or 26 percent (BWC) for that particular reactor coolant pump situation, Curve 1 of Figurs 2.1-3 is the most restrictive because I

any pressure-temperature point above and to the left of this curve will be above and to the left of the other curves.

REFERENCES (1)

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

(2)

BWC Carrelation of Critical Heat Flux, BAW-10143P-A, April, 1985.

(3)

FSAR, Section 3.2.3.1.1.c.

l n

Amendment No. 2Z, 3Z, 43, 52, 87, 92,113 9 i

2400 3200 f

ACCEPTABLE E

OPERATCN so f

I 2000 k

f UNACCEPTABLE OPERATCN o

1800 p

1600 480 000 420 440 860 REACTOR OUTLET TEMPERATURE. *F 4

CORE PROTECTION SAFETY LIMIT FIGURE N0. 2.1-1 Amendment No. 22.113 94

+

Core Protection Safety Limits - ANO-1 Figure 2.1-2 THERMAL POWER LF. VEL

%FP

--140 i

-120

(-3 3.04,112.0) ^

, (33.04,112.0) i lACCEPTABLE l

4 PUMP l

p (45.2 7,100.5 5)

-100 OPERATION l

(-33.04,90.75) l (33.0,4,90.75) lACCEPTABLE

(-62.32,84.45) l 4 & 3 PUMP

-80 p (45.27,79.30) l OPERATICN

(-6 2.32,63.20) (k

^l l

l

- (33.$4,64.08)

I l ACCEPTABLE l

-60 I

4,342 PUMP (45.27,52.63) l OPERATION l

-40

(-62.32,36.53)4 i

-20 l

l I

e t

i i

t i

-60

-40

-20 0

20 40 60 REACTOR POWER IMBALANCE, %

i l

j Amendment No. I, 22, 31, dJ, 52, 27, 72,12.113 Sb i

o

_,_m

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

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

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

4 2200

{

1 r

LT

(

so 2

2000 i

iO 1800 ff i

1600 680 000

$20 440 460 l

REACTOR OUTLET TEMPERATURE.Y r

CURVE OPM POWER PUMPS OPERATING (TYPE OF LIMIT)

(

i 1

374,840 (100%)

• 142%

FOUM PUMPS (DNBR LNff) t

{

2 280,035 (74.7%)

SOA%

THREE PUMPS (QUALITY LNTT)

(

3 184,441 (40.2%)

43.7%

ONI PUMP N EACH LOOP (QUAlffY LNff) l

• 104A% OF DESON FLOW I

Amendment No. ZZ, 92,113 Se i

1

2.3 LIMITING SAFETY SYSTEM SETTINGS, PRDTECTIM INSTRUMENTATION Applicability Applie: to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.

Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.

Specification 1

2.3.1 The reactor protection system trip setting limits and the l

permissible bypasses for the instrement channels shall be as stated in Tablo 2.3-1 and figure 2.3-2.

i Bases The reacter protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a i

reactor trip if any one of these conditionA deviates from a prestlee.ted operating range to the degree that a safety limit may be reached.

The trip setting limits for protection systtm instrumentation are listed in Table 2.3-1.

The safety analysis has been based upon these protection system instrumentation trip setpoints plus calibration and instrumentation errors.

1 Nuclear Overpowar A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from rea:tivity excursions too rapid to be

, detected oy pressure and temperature measurements.

During normal plant operation with all reactor coolant putps operating, reactor trip is initiated when the reactor power level reaches 104.9

)

percent of rated power.

Adding to this the possible variation in trip t

setpoints due to calibration and instrument errors, the maximum actual i

l power aj. which a trip would be actuated coJ1d be 112%, which is the value i

used in the safety analysis.

1 l'

A. Overpower Trip Based on Flow and Imbalance The power level trip setpoint produced by the reactor coolant system flow is based on a power-to-flow ratio which has been t

established to accommodate the most severe thermal transient

~

considered in the design, the loss-of-coolant-flow accident from I

high power. Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DNBR of less than 1.30 (BAW'2) or 1.18 (BWC) should a low flow condition exist due to any

[

electrical malfunction.

j Amer.Nent No. II, 43, 57,113 11 a

. _. _ ~

The power level trip setpoint 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.

Tne power level trip setpoint produced by the power-to-flow ratio provides overpower DNS protection for all modes of pump operatf 6n.

For every flow rate there is a maximum permissible power level, and for every power level there is a minimum perufssible low flow rate.

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

No penalty in reactor coolant flow through the cere was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outaje.

For safety analysis calculations the maximum 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 arn 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 assaciated reactor power-to-reactor power imbalance boundaries by 1.07 percent for a 1 percent flow reduction.

B.

Pump Honitors In conjunction with the power imbalance / flow trip, the pump monitors prevent the einimum core ONBR from decreasing beloi 1.30 (BAW-2) or 1.18 (8WC) by tripping the reactor due to ti.;

l loss of reactor coolant Amendment No. 22, II, #3, 52, 57, 92,113 12

pump (s).

The pump monitors also restrict the power level for the number of pumps in operation.

C.

RCS Pressurc During a startup accident from low power or a slow rod withdrawal froa high power, the system high pressure trip is reached before.

the nuclear overpower trip setpoint.

The trip se,tting limit shown in Fi psig)gare 2.3-1 for high reactor coolant system pressure (2355 has been established to maintain the system pressure below the safety limit (2750 psig) for shy design transient.(a)

The low pressure (1800 psig) and variable low pressure (13.89T

-6766)tripsetpointshowninFigure2.3-1havebeenestablishS8g to maintain the ONB ratio greater than or equal to the minimum allowable DNB ratio for those design accidents that result in ;

pressure reduction.(2,3)

To account for the calibration and instrumantation errors, the accident analysis used the safety limit of Figure 2.1-1.

D.

Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (618F) shown in Figure 2.3-1 has been established to prevent i

excessive core coolant temperatures in

• a operating range.

Due to calibration and instrumentation err s, the safety analysis used a trip setpoint of 620F, E.

Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor builoing or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

F.

Shatdown Bypass In order to provide for control rod drive tests, 2nro power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system.

The reactor protection system segments which can be bypassed are shown in Table 2.3-1.

Two conditions are imposed when the bypass is used:

1.

A nuclear overpower trip setpoint of 55.0 percent of rated power is automatically imposed during reactor shutdown.

2.

A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.

Amendment No. 2, 21, H, 87, IN,11313

o.

2500 P:2355 PSIG T= 618 'F 2355 2300 I

ACCEPTABLE OPERATION uI 2100 W

!E 5

8o 1900 o

Pa (1329T -6766) PSG Uh' ACCEPTABLE g

Q OPERATON I

P=1800 PSIG 1700 1500 560 580 600 620 640 660 RE ACTOR OUTLET TEMPERATURE, *F PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINT Figure 2.3-1 Amendment No. 27. 49, 67. 204. 113 14a

Protective System Maximum Allowable Setpoints ANO-1, Figure 2.3-2 THERMAL POWER LEVEL, % FP

--140

-120 g

(-184,107)-

(184,107)

I

-100 ACCEPTABLE 4 PUMP (34.7,90.1)

OPERATON

-80 m (184,79.9)

(- 18.0,79.9)

(-51.0,744),

I l

ACCEPTABLE l

l 3 & 4 PUMP OPERATON.

60 i

I

(-184,52.6) I (184,52.61)

(-514,4 6.9) q ACCEPTABLE 2,3 & 4 PUMP --40 OPERATON

' (34.7,35.7) 1 I

(-51.0.19.6)

-20 g

I i

i

[

I i

i i

i, 40

-40

-20 0

20 40 60 REACTOR POWER NBALANCE, %

l Amendment No. I, 2I, JI, 42, 52, 57, 7I, 52 14b 113

. - = _ _

l Tabla 2.3-1 1

i Reactor Protection System Trip Setting Limits f

~

One Reactor Coolant Pump l

Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop

~

1 Operating (Nominal Operating (Nominal (Nominal Operating Shutdeun j

Operating Power - 100E)

Operating Power - 75%)

Power - 495)

Bypass Nuc1:ar power, % of 104.9 104.9 104.9 5.0 *I I

r:ted, max Nucigar Power based on 1.07 times flow minus 1.07 times flow minus 1.i,1 times flow minus Bypassed flow and labalance, reduction due to reduction due to reduction due to

% cf rated, max imbalance (s) imbalance (s) isbalance(s)

Nuc1:ar Power based on NA NA 55 Bypassed pumpmonitops,%of r:ted, max High RC system 2355' 2355 2355 1720' pressure, psig, max Low RC system 1800 1800 1800 Bypassed pressure, psig, ein d

d d

Verfable low RC 13.89 T" -6766 13.89 T

-6766 13.89 T

-6766 Bypassed l system pressure, psig, min RC temp, F, max 618 018 618 618 3

)

High reactor building 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 4

pressure, psig, max psia)

1. utomatically set when other segments of the RPS (as specified) are bypassed.

A b : actor coolant system flow.

R

!CThe 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.

i dT,,g is given in degrees Fahrenheit (F).

Amendment No. 2, II, #3, if, 52, 57, ff. IN,113 15 f

I

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 defined in specification 3.5.2.5.

3.5.2.4 Quadrant tilt:

1.

Except for physics tests, if quadrant tilt exceeds A.12%,

l reduce power so as not to exceed the allowable power level for the existing reactor coolant pump combination less at least 2% for each 1% tilt in excess of 4.12%.

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 4.12% except for physics tests, or i

the following adjustments in setpoints and limits shall be made:

a.

The protection system maximum allowable setpoint:

(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 4.12%.

l l

L c.

The operational imbalance limits shall be reduced 2%

in power for each 1% tilt in excess of 4.12%.-

l 3.

If quadrant tilt is in excess of 25%, except for physics tests or diagnostic testing, the reactor will be placed in the hot shutdown condition.

Diagnostic testing during power cperation 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. d, 22, 3Z, 43, 52, 285 11347

1 j

3.

Except for physics tests or exercising control rods, the control rod withdrawal ifmits are specified on Figures 3.5.2-1(A-C), 3.5.2-2(A-C), and 3.5.2-3(A-C) for 4, 3, and 2 pump operation respectively.

If the applicable 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 or exercising axial power shaping rods (APSRs), the following limits apply to APSR position:

i Up to 410 EFPD, the APSRs may be positioned as necessary for transient imbalance control, however, the APSRs shall be fully withdrawn by 410 EFPD.

After 410 EFPD, the APSRs shall not be reinserted.

With the APSRs inserted after 410 EFPD, corrective measures shall be taken immediately to achieve the full withdrawn position.

Acceptable APSR 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 />.

3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not l

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.

Excepc for physics tests, imbalance shall be maintained within the envelope defined by Figure 3.5.2-4(A-C).

If the imbalance is not within the envelope defined by Figure 3.5.2-4(A-C),

j c

corrective measures shall be taken to achieve an acceptable Imbalance.

If an acceptable imbalance is not achieved within 4 l

hours, 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 I

with limited access to be authorized by the Superintendent.

Bases i

The power-imbalance envelope define:: D Figure 3.5.2-4(A-C) is based on l

l LOCA analyses which have defined the maximum linear heat rate (see Figure I

3.5.2-5,), such that the s,aximum cladding temperature will not exceed the l

Final Acceptance Criteria.

Corrective measures will be taken isnediately I

should the indicated quadrant tilt, rod position, or imbalance be outside c

their specified boundaries.

Operation in a situation that would cause the Final Acceptance Criteria to be approached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and fabalance) must be at their limits while f

Amendment No g 2Z, 3Z, 43, 52, 92, 185 48

Rod Position Setpoints for 4-Pump Operation From 0 to 27+10/-0 EFPD ANO-1 Cycle 9 Figure 3.5.2-1A I

110 (44.1.102) 070A,102) -

100 OPERATCN N TNS

[

REGON 5 NOT J

^

90 SHUTDOWN 0664,90)

W.RG N LMT O

g4gg,73) 70 I

OPERATON at 60 RESTRCTED a'

Y

~

g (41.5,48)

Q 124,@

l l

40 30

~

PERMESELE g

OPERATNG REGCN l

l 20 (5J.13) i 10 0

O 20 40 SO 80 100 120 140 160 180 200 220 240 260 280 300 O

20 40 SO 80 100 t

f i

1 i

i GROUPT 0

to 40 40 80 100 i

t I

i t

i I

GROUP 6 0

20 40 00 80 100 a

1 I

I i

J ROD NDEX, % WD i

l Amendment No. 5, 2Z, 4Z, 43. 52, 7Z. 92. III 48b 113,

i

Rod Position Setpoints for 4-Pump Operation From 27+10/-0 to 360 +50/-10 EFP0 ANO-1 Cycle 9 i

Figure 3.5.2-1B l

t 110 MU (189.5.102)

(%0,102) 100 SHUTDOWN MARGN M

M (264.0,90)

=

80 (244.0,78)

CPERAMN 1 70 RESTRCTED g

OPERATCN N THG REGCN 2 NOT N 60 alt. OWED

1. 50 d

(79.5,48)

(133.0,48)

$2' 30 PERM $$5LE 20 OPERATNG (31.5.13)

REGCN 10 (0.6.3) l 0

i e

i t

i 1

e e

i e

i e

i i

j 0

20 40 SO 80 100 120 140 160 180 200 220 240 260 280 300 t

0 20 40 60 30 100 I

I f

f I

I i

GROUP 7 0

20 40 80 30 100 i

i i

i i

GROUPG

?

YTYY 190 RODDCEX.% WD GROUP 5 t

t I

Amendment No. If5 113 48c I

h I

Rod Position Setpoints for 4 - Pump Operation l

Af ter 360 +50/-10 EFPD ANO-1 Cycle 9 l

Figure 3.5.2-1C I

f i

1,'O SHUTDOWN (1593,102)

(264A,102) -

.(300,102) l 100 MARGN i

I LMf7 (264A90) l 30 l

OPERATION I

30 RESTRIOTED l

(244A78) 70 OPERATCN N TH6 REGCN 6 NOT 8

60 ALLOWED ut 50 (79.5,48) 40 1

30 PERMESBLE 20 OPERATNG i

REGON 10 W 13)

I (0.6.3) l 0

O 20 40 to 30 100 120 140 160 ISO 200 220 240 260 280 300 i

O 20 40 00 80 100 i

l l

i I

i t

I GROUP 7 0

20 40 e0 s0 100

)

I I

I I

I i

GROUPG I

O 20 40 90 80 1p 1

i t

i I

GROUP 5 ROD NDEX, % WD I

l l

\\

t

{

Amendment No. Igg.113 48d r

e O

e Rod Position Setpoints for 3-Pump Operation From 0 to 27+10/-0 EFPD -- ANO-1 Cycle 9 l

Figure 3.5.2-2A 110 100 90 30 h

'I OPERATON N THE 70 REGON 6 NOT SHUTDOWN g

60 yAgon i

(2484.58) p g

50 OPERATON RESTRCTED 40 (41.5.36)

(2124.35.5) i 30 20 PERM 6 SELE (5.5,9,75)

O jo REGCN IO'# 75) f 0

0 20 40 60 80 100 120 140 160 180 200 220 240 260 200 300

?

???

Y '?o GROUP 7 k

YiYY

'Y ORolP 6 i

O 20 40 00 SO 100 t

t I

t t

i GROLP 5

.ROO PCEX, % WD t

Amendment No.22, 3I, 52, 71, 92, If5 48e 113 i

Rod Position Setpoints for 3-Pump Operation From 27+10/-0 to 360 +50/-10 EFPD -- ANO-1 Cycle 9 Figure 3.5.2-28 110 100 30 SHU7DOWN E

so uARon 3

(16oA 77)

(269.0,7 ?) j

-.(300,77) y 70

~

OPERATION N THE

[

REGCN 6 NOT 60 PERAMN e.OwrD QM.0,5&

Go 4o

~

(1984,35.5)

=

(

to OesmTwo nescu 10 (31.5,0.75) i g 5475)

O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 l

0 s,o d,o eo ao,oo B

I i

i GROUP 7

)

Y YYY

'iO anou, e 0

20 40 30 go too j

f f

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t i

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I l

I l

Amendment No.III, j33 48f l

l l

l I

Rod Position Setpoints for 3-Pump Operation Af ter 360 +50/-10 EFPD -- ANO-1 Cycle 9 l

Figure 3.5.2-2C 110 100 90 SHUTDOWN

""RON 80 tm (160.4,77)

(269.0,77) -

(300,77) 70 OPERATON N THS

.0,67) 8 60 PEGON 6 NOT 8

ALLOWED (244 2 )

OPERATON 50

t RESTRCTED 2

40 (79.5,36)

(198.0.35.5) 30 i

20 PERMtSSIBLE OPERATING REGION 10 5 475),

,O 20 40 30 30 100 120 140 160 130 200 220 240 260 280 300 Y

Y GROUP 7 YTYY 1Y oRew a I

0 2,0 4,0 00 8,0 10,0 l

f OROLP5 ROO PCEX, % WV I

L i

l Amendment No. If5,113 48g l

i i

I e

l 1

Rod Position Setpoints for 2-Pump Operation From 0 to 27+10/-0 EFPD -- ANO-1 Cycle 9 l

Figure 3.5.2-3A l

i 110 l

100 90 Ia 70 g

'i

.0 OPERATION N THE l

l g

MEGON 5 NOT (85.7.52)

(271.3.52)-

(30042)

(266A44) l T N 40 N WN RESTRCTED G48A38) pgggg 30 LMT (41J.24) c12A23)

PERM 63str 10 (5.5,6.5)

OPERATNG (0,11),

REGCN 0

0 20 40 40 40 100 120 140 100 180 200 220 240 2 0 200 300 p

0 20 40 80 ao 100 i

e I

t I

t ORoup 7 0

to 40 to 80 100 i

e a

t e

t

[

amoup e 0

20 40 00 80 100 i

i e

t a

ROD M % W ORoup 5 l

t Amendment No. 2I JI, 42, 52, 72, 52, 185 48h l

11$

Rod Position Setpoints for 2-Pump Operation From 27+10/-0 to 360 +50/-10 EFPD -- ANO-1 Cycle 9 Figure 3.5.2-38 l

110 100 90 80 Ig 70 g

SHUTDOWN 4 60 O

MARON LNIT (162.4.52)

(269.3,52) -

(300,52) g OPERATON N THis (264.0,44)

I k 40 REGCN 16 NOT ALLOWED

"'M)

RESTRCTED 30 i

(79.5.24)

(198.0,23) 20 4

PERM 45 SALE 10 (o,3,1)

OPERATNG (31.5,6.5)

REGCN g

i e

t i

1 t

I t

t t

t i

t t

0 to 40 60 80 100 120 1 44 960 140 200 220 240 260 280 300 0

to 40

.0

.0 500 i

t t

t t

I GROUP 7 dP Y Y

'?*

i oROuP e y

2,0 4,0 yy y

OROUP5 MOD NDEX, % WD 1

i l

Amendment No. 285. 113 481 l

O Rod Position Setpoints for 2-Pump Operation After 360 +50/-10 EFP0 -- ANO-1 Cycle 9 l

o Figure 3.5.2-3C d

I 110 100 90 t

90 70 8

60 SHUTDOWN l

8 MARGN 162.4,52)

(269.3,52)-

(300,52)

M 60 (264.0,44) 40

• OPERATON N THIS OPERATION (244.0.38)

REGCN 5 NOT RESTRCTED ALLOWED i

30 (79.5,24)

(194.0.23) 20 PERMIS$8LE OPERATNG (33.5,6.5) l (0,3.1) <

0 0

20 40 60 SO 100 120 140 160 180 200 220 240 260 280 300

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20 40 to to 100 t

i e

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i a

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C 20 40 00 s0 100 i

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GROUP 5 ROO NDEX, % WD l

j J

l l

i Amendment No. Iff. 113 48j l

i

s Operational Power Imbalance Setpoints for Operation From 0 to 27+10/-0 EFPD -- ANO-1, Cycle 9 l

Figure 3.5.2-4A 9

. 110

(-14.90,102) "

-100

(-18.95,32)

-90

-80 (23.19,80)

(-23.78.80)

RESTRICTED j--70 RESTRICTED REGION 2

REGION g

l- -60 N

(-30.16,50) A h

$--50 A (26.04,50) r 5

g- -40 m

h k-30 2

-to I

10 i

e

. 40 20 -10 0

10 20 30 40 60 AXLAL POWER.NBALANCE, %

Amendment No. 21, 52, if,185 113 48k l

Operational Power Imbalance Setpoints for Operation From 27+10/-0 to 360 +50/-10 EFPD -- ANO-1, Cycle 9 l

Figure 3.5.2-4B e

1 110 I

(-20.21.102) 7

- (21M,102)

--100

(-22.92)

,g (21.74A2) if (23.41,80)

(-28.95,80)

- 80 RESTRCTED l

REGCN REGCN u,, so 8

N (26.29.50)

(-31D6.50)

W

1. --50 g

E W

W

-40 g-1 30 l

-20 l

-10 t

1 1

i f

i 1

1 t

t

-60

-40

-30 10 0

10 20 30 40 So i

AXRt. POWER NSA1.ANCE. %

r I

Amendment No. Iff,113 481 l

1 Operational Power Imbalance Setpoints for Operation Af ter 360 +50/-10 EFPD -- ANO-1, Cycle 9 l

Figure 3.5.2-4C 110

(-20.21,102) -

, (21.5a,102)

-"100

.(-25.62.92)

(25.03,92)

.g

(-28.95.80)

- 80 (27.54,80) h- -70 RESTRCTED RESTRCTED REGCN REGCN kn--so g

a', --50 0 (31.20.50)

(-31.06,50)6 l

.4o j

2 i

.so

.3

-10 t

t t

I 1

i f

i f

f

• 40 20 -10 0

10 20 30 40 SO AXML POWER DASALANCE, %

)

Amendment No. 285. 113 48s i

3

.~

LOCA Limited Maximum Allowable l

Linear Heat Rate Figure 3.5.2-5 b

20 14 I

I

/

\\

i g

[

l I

E 16

/

r

=

/

8

/

i4 i

B S

Y 1

l 12 0-1000 mwd /mtU x

3 AFTER 1000 mwd /mtU 10 0

2 4

4 4

to 12 AXtAL LOCATON FROM BOTTOM OF CORE, ft.

Amendment No. 43, 52, 72, if, 185 48n l

113

4 i.-

1 5.3 REACTOR Specification 5.3.1 Reactorfore 5.3.1.1 The reactor core contains approximately 93 metric tons of l

slightly enriched uranium dioxide pellets, The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of 177 fuel assemblies.

Each fuel assembly is fabricated with 208 fuel rods.

(1,8) Starting with Batch 11, a reconstitutable fuel assembly design is implemented.

This design allows the replacement of up to 208 fuel rods in the assembly.

5.3.1.2 The reactor core approximates a right circular cylinder with an equivalent diameter of 128.9 inches and an active height of 144 inches.

The active fuel length is approximately 142 inches.(8) l 5.3.1.3 The average enrichment of the initial core is a nominal 2.62 weight percent of 23 U.

Three fuel enricheents are used in the initial core.

The highest enrichment is less than 3.5 weight percent 2 sg, 5.3.1.4 There are 60 full-length control rod assemblies (CRA) and 8 i

axial power shaping rod assemblies (APSRA) distributed it. the reactor core as shown in FSAR Figure 3-60.

The full-length CRA contain a 134-inch length of silver-indium-cadminum alley clad with stainless steel.

Each APSRA contains a 63-inch length of Inconel-600 alloy.(8) 5.3.1.5 The initial core has 68 burnable poison spider assemblies witn similar dimensions as the full-length control rods.

The cladding is Zircaloy-4 filled with alumina-boron and placed in the core as shown in FSAR Figure 3 2.

L 5.3.1.6 Reload fuel assemblies and rods shall conform to the design and evaluation described in FSAR and shall not exceed an enrichment of 3.5 percent of 888U.

5.3.2 Reactor Coolant system 5.3.2.1 The reactor coolant system is designed and constructed in accordance with code requirements.(*)

l 5.3.2.2 The reactor coolant systes and any connected auxiliary sys.tems exposed to the reactor coolant conditions of temperature and pressure, are designed for a pressure of 2500 psig and a temperature of 650 F.

The pressurizer and pressurizer surge line are designed for a temperature et 670 F.(s) 5.3.2.3 The reactor coolant system volume is less than 12,200 cubic i

feet.

Amendment No IfJ,113 114

l' REFERENCE 5:

(1)

FSAR, Section 3.2.1 (2)

FSAR, Stetten 3.2.2 (3)

FSAR, Section 3.2.4.2 (4)

FSAR, section 4.1.3 (5)

FSAR, Section 4.1.2 3 15

.