ML19326C056

Amend 21 to License DPR-51,revising:core Protection Limits, Fuel Rod Bow Analyses,Tech Spec Operating Limits Based on ECCS Performance Evaluation & NRC Approved B&W Nucleate Boiling Heat Transfer Correlation Model

ML19326C056
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/21/1977
From:	Goller K Office of Nuclear Reactor Regulation
To:
Shared Package
ML19326C050	List: ML19326C049 ML19326C056 ML19326C057 ML19326C059
References
NUDOCS 8004180731
Download: ML19326C056 (34)

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1 ARKANSAS NUCLEAR ONE - UNIT NO. 1 ggj t. s..

Y AMENDMENT T0" FACILITY OPERATING LICENSE

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Accordingly, the. license is amended by changes to the Technical. ',f;3 2.

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Specifications as indicated in the attachment to this license ~.

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No. DPR-51 is hereby amended to read as follows:

(2) Technical _ Specifications M

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The Technical " Specification's. contained in' Appe'ridices.

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ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. DPR-51 DOCXET NO. 50-313 Accomplish page changes to the Appendix A portion of the Technical Specifications as noted below. The changed areas on the revised pages are identified by a marginal line.

Remove Existing Page Add Revised Page 7

7 8

8 9

9 9a 9a 9b 9b 9c 9c 11 11*

12 12 13 13 14a 14a 14b 14b 15 15 16 16 17 17 29 29*

30 30 47 47 48 48 48b 48b 48bb 48bbb 48c 48c 48cc 48ccc 48d 48d 48dd 48dd 48ddd 48ddd 48e 48e 48f 73a 73a 1 01 101*

l 102 102 1

I

• There were no changes on these pages. They are included as a matter of convenience in updating the Technical Specifications.

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F.

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SAFI lY LIMITS AND LIMITING SAFETY SYSTD1 SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during' power operation of the plant.

Obj ective s

To maintain the integrity of the fuel cladding.

Specification 2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in. Figure 2.1-1.

If the actual pressure / temperature point is below and to the right of the pressure / temperature line the safety limit is exceeded.

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 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 product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat 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 departur'e from nucleate boiling (DNB). At this point there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the po'ssibility of eladding failure. Althobgh 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 DNB through the use of the BAW-2 correlation. (1) The bah-2 correlationf has been develoi.ed to predict DNB and the location of DNB for axially uniform and non *iniform heat flux distributions. The local DNS ratio l

(DNBR), defined as the radio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients.is limited to 1.3.

A

' Amendnent No.

7

. (,.'

e DNBR of 1.3 corresponds to a 95 percent probability at a 95 percent confi-dence level.that DNB will not occur; this is considered a conservative mar-gin 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 detemining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a,v 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 minimum DNER of '1.3 is predicted. The curve is the most restrictive com-7-

binatiori 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 ;

Fq = 2.67; Fft!=1.78; N

F = 1.50 These design limit power peaking factors are the most restrictive calculated at full power for the range from all contrci rods fully withdrawn to maximum allowabic control rod insertion, and for.a the core DNBR design basis.

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 and. fuci rod bowing:

1.

The 1.3 DNBR limit produced by a nuclear power peaking factor of Fy=2.67orthecombinationoftheradialpeak,a.ialpeakand position of the axial peak that yields no less than 1.3 DNBR.

2.

The combination of radial and axial peak that prevents central fuel melting at the hot spot. The limit is 19.4 kW/ft.

Power peaking is not a directly observabic quantity and therefore limits have

.been established on the basis of the reactor power imbalance produced by the power peaking.

The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 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 naximum themal power combinations shown in Figure 2.1-3.

The curves of Figure 2.1-3 represent the conditions at which a minimum DNBR of 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of mini-mum DNBR is equal to 22 percent (1) whichever condition is more restrictive.

Amendment No.

,y 4

+,.

C.

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 minin um 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 for three pump operation is 86.0 percent due to a power icvel trip produced by the flux-flow ratio (74.7 percent flow x 1.065=

79.6 percent power) plus the maximum calibration and instrumentation error.

The maximum thermal power for other reactor coolant pump conditions is pro-duced in a similar manner.

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.lcss than 22 percent for that particular reactor coolant pump situation. Curves 14 2 of Figure 2.1-3 are the most restrictive because any pressure / temper stre 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, BArl-10000A, May, 1976.

(2) FSAR, Section 3.2.3.1.1.c 2

(

Amendment No.

)

2600 2400 i

E.

2200 E

f 5

ACCEPTABLE m

5 OPERATION o.

ea 2000 o

UNACCEPTABLE OPERATION 1800 f

~

1600 640 660 560

~580 600 620 Reactor Outlet Temperature. *F l

I ARKANSAS POWER & LIGHT CO.

FIG' NO~

ARKANSAS NUCLEAR ONE UNIT 1 CORE PROTECTION SAFETY LIMIT 2.1 -1 Amendment No.

9a

THERMAL POWER LEVEL, 5 UNACCEPTABLE OPERATION Kw/1 t LIMIT ACCEPTABLE 4 PUMP OPERATION 100 0

ACCEPTABLE 3&4 80 PUMP OPERATION a

60 ACCEPTABLE 2,3,& 4 PUMP OPERATION y

40 1

20

• MORE LIMITING THAN CALCUL ATED DNBR OR KW/FT LIMITS I

I I

I

'I l

60

-40 20 0

20 40 60 Reactor Power imoalance CURVE REACTOR COOLANT FLOW (gpm) 1 374,880 2

280,035 3

184,441 AkKANSAS POWER & LIGHT CO.

FIG. NO.

ORE PROTECTION S M H W ITS ARKANSAS NUCLEAR ONE - UNIT 1 2.1 2 l

Amendment No. E, 9b s

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2600 2400

.5 E

2200

/

=

t 2 a.

w

/

2000 a

~

1800 f

f 1600 560 580 600 620 640 660 Reactor Outlet Temperature. 'F CURVE GPM POWER PUMPS OPERATING (TYPE OF LIMIT)

~

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1 374,880 (1005)*

1125 FOUR PUMPS (DN8R L!MIT) 2 280.035 (74.75) 86.75 THREE PUMPS (ONSR LIMIT) 3 184.441 (49.25) 59.05 ONE PUMP IN EACH LOOP (00All!1 LIMIT)

• 106.55 0F DESIGN FLOW ARKANSAS POWER & LIGHT CO.

ARKANSAS NUCLEAR ONE UNIT 1 CORE PROTECTION SAFETY LIMITS FIG. NO.

2.f-3 Amendment No.

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2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION

/

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Applicability Applies 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 2.3.1 The reactor protection system trip setting li=its and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2.

Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which vill cause a reactor trip if

,J any one of these conditions deviates from a pre-selected opnating range to f

the degree that a safety limit may be reached.

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

The safety analysis has been based upon these protectica system instrumentation trip set points plus calibration and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent Ammage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.

During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 105 5 percent of rated power. Adding to this the possible variation in trip tet points due to calibration and instrument errors, the maxi =um actual power at which valueusedinthesafetyanalysis.p2".,whichismoreconservativethanthe a trip would be.wtuated could be i

A.

Overpower trip based on flov and imbalance The power level trip set peint produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the moct severe thermal transient considered in the design, the loss-of coolant flow accident from high power. Analysis has demon-strated that the specified power to flow ratio is adequate to prevent a DNBR of less than 1.3 should a lov flow condition exist due to any ele-ctrical malfunction.

11 Amendment No.

The power Icvel trip set point produced by the power-to-flow ratio provides both high power icvel and low flow protection in the event the reactor power Icyc1 increases or the reactor coolant flow rate decreas es.

Tlie 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 pemissible power level, and for every power level there is a minimum pemissible 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 reactor coolant pumps are operating if power is 106.5 percent and reactor flow rate is 100 percent or flow rate is 93.9 percent and power level is 100 percent.

2.

Trip would occur when three reactor coolant pumps are operating if power is 79.5 percent and reactor flow rate is 74.7 percent or flow rate is 70.4 percent and power icvel is 75 percent.

3.

Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.3 percent and reactor flow rate is 49.2 percent or flow rate is 46.0 percent and the power icvel is 49.0 percent.

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

No penalty in reactor coolant flow ths ugh the core was t tken for an open core vent valve because of the core vent valve surveillance program during each refueling outage.

For safety analysis calculations the maximum cali-bration and instrumentation errors for the power level were used.

l The power-imbalance boundaries are established in order to prevent reactor themal limits from being exceeded. These themal 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 1evel 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 1.065 percent for a 1 percent flow reduction.

B.

Pump monitors In conjunction with the power imbalance / flow trip, the pump moni-tors prevent the minimum core DNBR from decreasing below 1.3 by trip-ping the reactor due to the loss of reactor coolant pump (s).

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

C.

During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit Amendment No.

_s

,. c.,

shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. (2)

The _ low pressure (1800 psig) and variable-low pressure (11.75 Tout

-5103) tcip setpoint shown in Figure -2.3-1 have been established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.

(2,3)

Due to the calibration and instrumentation errors,the safety analysis used a variable low reactor coolant system pressure trip value of (11. 75 Tout-5143).

D.

Coolant outlet temperature The high' reactor coolant outlet temperature trip setting limit (619F) shown in Figure 2.3-1 has been established to prevent ex-cessive core coolant temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620F.

E.

Reactor bui? ding 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 building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

F.

Shutdown bypass In order to provide for control rod drive tests, zero 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 Tabic 2.3-1.

Two conditions are imposed when the bypass is used:

~.

1.

A nucicar overpower trip set point of <5.0 percent of rated power is automatically imposed during reactor shutdown.

2.

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

The purpose of the 1720 psig high pressure trip set point is to prevent normal. operation with part of the reactor protection system bypassed. This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated.

The overpower trip set point of <5.0 percent prevents any significant reactor power from being produced when performing the physics tests. Sufficent natural circulation (5) would be available to remoJe 5.0 percent of rated power if none of the reactor coolant pumps were operating.

Amendment No. 2,

~

.z 4

2500 P = 2355 psig

' T = 619*F ya 2300 E.

ACCEPTABLE OPERATION N

=

E 2100 Z

P = 11.75 Tout-3 5103 psig 1900 bHACCEPTABLE OPERATION P = 1800 psig 1700 1500 560 580 600 620 640 660 Reactor Outlet Temperature, F

ARKANSAS POWER L l.lGHT CO.

PROTECTIVE SYSTEM MAXIMUM FIG. NO.

ARKANSAS NUCLEAR ONE-UNIT 1 ALLOWABLE SET POINT 2,3.g Amendment No.

ua

m THERMAL POWER LEVEL, "4

120 j

UNACCEPTABLE OPERATION (106.5) i, 100

+,

%\\

ACCEPTABLE 4 y

PUMP OPERATION

'O/j (79.5) 80 ACCEPTABLE 3& 4 PUMP OPERATION 60 (52.3)

A C C E P T A B L E 2.3

& 4 PUMP OPERATION

-- 40 20 h

h S

ll 18 11 il

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-60 40

-20 0

20 40 60

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P0wer Imualance, %

ARKANSAS POWER & LIGHT CO.

PROTECTIVE SYSTEM MAXIMUM FIG. NO.

ARKANSAS NUCLEAR ONE - UNIT 1 ALLOWABLE SETPOINTS 2.3-2 Amendment No. 5, 14b s

Table 2.3-1

(,

Reseter Frstsetien Systen Trio S:tting Lielts

.g no 3

One Reactor Coolant Pump Four Resctor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop 2

Operstin~g (Nocinst~

.o

. Operating (Namins)

(Nominst Operating coersting Power - 100'.)

Orars t ing Power - 75%)

!5hu td own.

Power - 49%)

U Nuclest power,1 of 105.5

_8ypass e

rstr.d, asx

- 105.5 105.5 5.0 (3)

Nuclear power based on.

1.065 times flow minus.

1.065 times flow minus 1.065 times flow minus flow (2) and imbalance, reduction due to reduction due to typsssed

\\ of rated, esx reduction due to irbs t ance (s) imbalance (s) imbstance(s)

Nucle.ar power based on NA NA purp conitors} % of Bypassed 555 rated, esx (4

/

liigh reactor coolant 2355 2355 sy>two presscr.:, psig, 2355 1720(3)

.:.x x -

Low reactor coolant sys--

1300 1800 tem pressure, psig. nin 1800

/

Bypassed Variable low reactor (11. 75 Tout-5103) (1)

(11. 75 Tout-5103) (.1)

(11.75 Tout-5103) (1) coolant syste:2 pressure,

,yp,,,,d psig, : sin Reactor coolant temp, 619 619 F, esx 619 619 liigh reactor building 4 (13.7 psis) pressure, psig, esx 4(18.7_ psis',

4 (13.7 psis) 4(13. 7 p si s (1) T is in degrees Fahrenheit (F).

.(3) out Autonsticstly set when other segments of the RPS (ss specified) are bypassed (2) Reactor coolant syste s flow, %.

(4)

The pur:p monitors also produce a trip on: (s) loss of two reactor coolant

. ~,,

pu:sps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-punp operation.

4 t e

3.

1.lillTING CONDITIONS FOR OPERATION

^

3.I' RiiA(*l'OR. C00lNil' SYSTEM Applicability Applies' to the operating status.of the reactor coolant system.

Obj ective

~

To specify those limiting conditions for operation of the reactor coolant sys-tem which must be met to ensure safe reactor operations.

.s 3.1.1 Operational Components Specification 3.1.1.1 Reactor Coolant lPumps A.

Pump combinations permissible for given power levels shall be as shown in Table 2.3-1. _

B.

The boron concentration in the reactor coolant systcm shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.

-3.1.1.2 Steam Cencrator A.

One stea.n generator shall be operable whenever the reactor coolant average temperature is above 280 F.

t 3.1.1. 3 Pressurizer Safety Valves A.

The reactor shall not remain critical unless both pressurizer code safety valves are operable.

B.

When the reactor is suberitical, at least one pressurizer code safety valve shall be operable if all reactor coolant system

, openings are closed, except for hydrostatic tests in accord-ance with AS.stE Boiler and Pressure Vessel Code,Section III, 3.1.1.4 Reactor Internals Ven,t Valves-The structural integrity and operability of the reactor internals vent valves shall-be maintained at a level consistent with the acceptance criteria in Specification 4.1.

J f

Amendment No.

16 9

9--

t re

~

.31105 A reactor coolant pump or weay heat removal pump is requiru to be in' opera-tion before the boron concentration is reduced by dilution with makeup water.

Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor..One decay heat removal pump will circulate the equivalent of the reactor coolant system volume in one half hour or less. (1)

The decay heat removal system suction piping is designed for 300 F thus, the system can remove decay heat when the reactor coolant system is below this temperature. (2,3)

One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, cnd reactor decay heat. (4) Both pres surizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities.

The code safety valves prevent overpres-sure for a rod withdrawal accident. [5) The pressuri:ev code safety valve lift set point shall be set at 2500 psig 1 I percent alle ance for error and each

' valve shall be capable of relieving 300,000 lb/h af saturated steam at a pressure not greater than 3 percent above the set pressure.

~

The internals vent valves are provided to relieve -the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered.

Inspection and manual actuation of the internals vent valves (1) ensure operability, (2) ensure that the valves are not open during normal operation, and (3) demonstrate that the valves begin to open and are fully open at the forces equivalent to the differential pressures assuded in the safety analysis.

REFERENCES (1) FSAR, Tables 9-10 and 4-3 throu'gh 4-7.

(2)

FSAR, Section 4.2.5.1 and 9.5.2.3.

(3) FSAR, Section 4.2.5.4.

-(4)

FSAR, Sect ion 4. 3.10. 4 and 4. 2. 4.

~

(5) FSAR, Section 4.3.7.

J 4

- Amendment No..

1,

~

b.

Total reactor coolant system leakage rate is periodically determined by comparing indications of reactor power, reactor coolant tercerature, pressurizer water level and reactor coolant makeup tank level over a time interval.

All of these indications are recorded. Since the pressurizer level is maintained essentially constant by the pressurizer level controller, any coolant leakage is replaced by coolant from the reactor coolant makeup tank resulting in a tank level decrease. The reactor coolant makeup tank capacity is 31 gallons per inch of height and each graduation on the level recorder represents 2 inches of tank helpht. This inventory monitoring method is capable of detecting changes on the order of 62 gallons. A 1 gpm leak would therefore be detectable within approximately 1.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

As described above, in addition to direct observation, the means of detecting reactor coolant leakage are based on different principles, i.e., activity, sump level and reactor coolant inventory measurements. Two syste=s of different principles provide, therefore, diversified veys of detecting leakage to the reactor building.

c.

The reactor building gaseous =enitor is sensitive to lov leak rates if expected values of failed fuel exist. The rates of reactor coolant leakage to which the instrument is sensitive are discussed in 'SAR Section 4.2.3.8.

The upper limit of 30 gpm is based on the contingency of a hypothetical loss of all AC power. A 30 gpm loss of water in conjunction with a hypothetical loss of all AC power and subsequent cooldown of the reactor coolant system by the atmospheric dump system and steam driven emergency feedvater pu=p would require more than 60 minutes to c=pty the pressurizer from the combined effect of system leakage and contraction. This vill be ample time to restore both

~

electrical power to the station and makeup flov to the reactor coolant system.

References FSAR Section h.2.3.8 1

Amendment No.

29

O. -

s'

/

3.1.7.

Modcrator Temperature Coefficient of Reactivity

/

Specification The moderator temperature coefficient shall not be positive at power levcis above 95% of rated power.

Bases A non-positive moderator coefficient'at power IcVels above 95% of rated. power is specified such that the maximum clad temperatures will not execed the Final Acceptance Criteria based on LOCA analyses.

Below 95% of rated power the Final. Acceptance Criteria will not tor temperature coefficient of +0.5 x '10'ge exceeded with a positive modera-6k/k/ F corrected to 95% of rated pcuer. All other accid nt analyses as reported in the FSAR have been perforgd

. for a rtnge of moderator temperature coefficients including +0.5 x 10-4 ok/k/"F.

When the hot' : crc-pover value is corrected to obtain the hot 954 value, the following corrections will be applied.

l.

Uncertainty in isothermal measuremnnt -- The measured moderator temperature coefficient vill centnin uncertainty eving to (a) +0.2 F in the AT of the base and perturbed conditicas, and (b) uncertainty in the reactivity meacurement of +0.1 x 10 4 Ak/k.

Proper corrections will be added for thene conditions to provide a conservative moderator ccefficient.

2.

Doppler coefficient at hot cro pcVer - During mensurement of the isothermal moderator coefficient at hot zero pcver, the fuel tem-perature vill increase by' the same amount as for the moderator.

l The measured temperature coefficier.t must be increased by 0.21 r

r 10-4 (ak/k)/ F to obtain a pure moderator temperature coefficient.

3.

!!cderator temperatu' e chance - The hot zero-power measurement must r

be reduced by 0.08 x 10-4 Ak/k/*F.

This corrects for the differ-ence in water temperature from zero power (532F) and 15% power (5SOF). Above this,pover, the average moderator temper,ature re-mains 580F 1

b.

Fuel temperature interaction (power effect ) -- The moderator coef-i ficient muut be adjusted to account for the interaction of.nn aver-f age moderatcr temperature with increasing fuel tenperatures (as j

power increases). Tnic correction is -0.0022 =.10-4 Aa /AI power.

g

)

It adjust the moderator coefficient at 15% power to the ccer-ficient at any power level above 15%. For example, the power er-feet adjustment from a 15% coefficient to leo; power is

(-0.0022 x 10-")(100% - 15% ) = -0.18'l = 10-4 Aa.

n Anendment No.

30

~.

~ ---

m

+

nm_t ro l rod in i the regulat ing or a x ia.1 powe r..-haping groups e

~p it a i-Jce la red inope rable per Speci ficar ;on 4. 7.1. 2. operat ion above no percent uf the thermal power allowable for the reactor coolant ipump combinat ion may continue provided the. rods i; the group are

positioned such that the rod that was deleared in0perable is con-

.tained_within allowable group average position I r r. i t s of Specifi ca-t ion 4 ~.'. l. 2 and the wi thdrawal l i m i t.s of. spec i fi cat i on 3. 5. 2. 5. 3.

3. 5. 2.'

The worth of-single inserte'd control rods during c rit:cality are -

limited by the restrictions. of Specification 3.1 3.3 aaJ the Control-

" Rod Posit ion.l.imit s defined in Specificat ion 3. 5 2. ~.

~

3.5.2.4' QuaJrant tiit:

1.

iieept for physics tests.-if quadrant tilt exccc '

3.411 po er shall be reduced immediately to below the power lesel ec aff (see Figure.

3. 5. 2-IN and 3.5. 2-1 B l. Marcoser, the power lea 1 entoff value Ili tilt.

For

~

shall be reduced 21 for each li.t ilt' in excess or less than 4 - pump operat ion, thermal power sha1i tu reduced 2* of the thermal power allowable for the reactor coolant pump combin-ation for each 1** t ilt in exces-of 3.41'..

I 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 t i i* < hall be reduced to less than 3.41% except for physics test s, or t h..

following adjust-l

-en t s in setnoints and limit, shall be made:

ilhe protect ion system maumum allowable set pa;nt il igure a.

2.3-21 Wha 11 be reduced.2" in power foreach !

tiit.

h.

TIIe control rod' group withdrawal limits it igu 3.5.2.IA, 3.5.2-1B and 3.5.2-IC shall be reduced 2*

in p me r for each l':

tilt in-excess of 3.41" The operat ional inhalance limit s -(Figures 3.5. -3A. 3.5.2-3B c.

i and c3. 5.2-3Cl _shall be reduced 2*. in power for each I':. t i l t in excess of 3. 41*..

4 3.

If quadrant tilt'is in execss of 25*.~, except for ph.s - n es tests or diagnostic testing,'the-reactor will be placed in the hot shutdown.

condition.

1)iagnostic' testing during power operat ion wi th a quad-rant power ti1t is permitted-provided the thermai power allowable for. the reactor -coolant pump combination is rest rict'ed as stated

.in.3.5.2.4.1.above.

~4.

! Quadrant tilt shall.be monitored on a minimum frequency of once es ery two hours during ' power operation above 15'. of rated power.

3. 5. 2. 5 -- Con t ro l_ rod pos i t i ons :

1.

'lechqlcal ' Specification 3.1.3.5 (safety rod w i thdrawal) does not

-prohibit _the exercising of individual safety. rods.as required by

-lable 4.1 2 or apply to inoperable safety rod limits in Technical 7

-Specification 3.5.2.2.

2.

Operating rod group overlap shall be 25'. +5'between two sequentiat groups', except for physics tests.

Amendment No 5.

n r

~

s w

3.

Except for physics tests or exercising control. rods, the control rod withdrawal limits are specified on Figures 3.5.2-1 A. 3.5.2-1B and 3.5.2-IC for four pump operation and on Figures t.5.2-2A, 3.5.2-2B and 3.5.2-2C for three or two pump oneration.

If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod pesition. Acceptable control rod positions shall be attained within four hours.

4.

Except for physics tests, power shall not be increased above the power level cutoff (see Figures 3.5.2-1) unless the xenon reactivity is.within 10 percent of the equilibrium value for operation at rated power and asymptotically approaching stability 3.5.2.6 Reactor Power 1mbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.

Except for physics tests, imbalance shall be maintained within the envelopes defined by Figures 3.5.2-3A, 3.5.2-3B and 3.9.2-3C.

If the imbalance is not within the envelopes defined by Figures 3.5. 2-3A,3.5.2-3B and 3.5.2-3C corrective measures shall be taken to achieve an acceptable imbalance.

If an acceptable imbalance is not achieved within four 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 with limited access to be authorized by the superintendent.

Bases The power-imhalance envelopes defined in Figures 3.5.2-3A, 3.5. 2-3B and 3.5.2-3C are based on 1) LOCA analyses which have defined the maximum linear heat rate (See Fig.

3.5.2-4) such that the maximum clad temperature will not exceed the final Acceptance l

Criteria and 2) the Protective System Maximum Allowable Setpoint s (Figure 2.3-2).

Corrective measures will be taken immediately should the indicated quadrant tilt, rori position, or imbalance be outside their specified boundary.

Operation in a situation that would cause the final acceptance criteria to l

be approached should a LOCA occur is highly improbable because all of the l

power distribution parameters (quadrant tilt, rod position, and imhalance) must be at their limits while simultaneously all other engineering and un-

~

certainty factors are also at their limits.* Conservatism is introduced by application of:

l a.

Nuclear uncertainty factors b.

Thermal calibration

,i c.

Fuel densification effects d.

Hot -rod manufacturing tolerance factors

~

e.

Fuel rod bowing The 25 15 percent overlap between successive. control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke. Control rods are arranged _in groups or banks defined as follows:

~

• Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument and calibration errors. The method

.used -to define _ the operating limits is defined in plant operating procedures.

Amendment No. 6, 48

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OPERAil0NAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 TO 115 1 10 EFP0 ARKANSAS C.YCLE 2 Figure 3.5.2-3A 7

Amendment'No.

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6 Figure 3.5.2-3B Amendment No.

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ARKANSAS NUCLEAR ONE-UNIT 1 LINEAR HEAT RATE 3.5.2.4 Amendmerit f. ).

g.

Table 4.1-2 (Continued)

Minimum Equipment Test Frequency Item Test Frequency 12.

Flow limiting Annulus Verify, at normal One year, two years, on Main Feedwater operating conditions, three years, and every Line at Reactor that a gap of at least five years thereafter Building Penetration 0.025 inches exists measured from date of between the pipe and initial test.

the annulus.

13.

SLBIC Pressure Calibrate Each Refueling Period Sensors 14.

Main Steam Isolation

a. Excercise Through

a. Quarterly Valves Approximately 10%

Travel

b. Cycle

b. Each Refueling Shut-down.

15.

Main Feedwater

a. Exerci~se Through

a. Quarterly Isolation Valves Approximately 5%

Travel

.i

b. Cycle

b. Each Refueling Shut-down.

16.

Reactor Internals Demonstrate Operability Each refueling shutdown.

Vent Valves By:

a. Conducting a remote visual inspection of visually accessible sur-faces of the valve body and disc sealing faces and evaluating any

-observed surface irregu-larities.

b. Verifying that the valve is not stuck in an open position, and

c. Verifying through manual actuation that the valve is fully open with a force of 5 400 lbs (applied vertically upward).

Amendment No. f, 73a

.. =

.y.

+e 14.

. 's

.~

V' Bases The emergency power system provides power require =ents for the engineered safety features in the event of a DBA. Each of the two diesel generators is capable of supplying minimum required engineered safety features from independent buses. This redundancy is a factor in establishing testing i

intervals. The monthly tests specified above vill de=onstrate operability and load capacity of the diesel generator. The fuel supply and diesel starter motor air pressure are continuously monitored and alarmed for abnorral conditions. Starting on complete loss of off-site power vill be verified by simulated loss-of-power tests at intervals not to exce-d each refueling shutdown period.

Considering system redundancy, the specified testing intervals for the station batteries should be adequate to detect and correct any mal-function before it can result in system malfunction. Batteries vill

' deteriorate with time, but precipitous failure is extremely unlikely..

.The surveillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it fails.

Routine battery maintenance specified by the manufacturer includes ree,ularly scheduled equalizing charges in order to retain the capacity of i

the battery. A test discharge should be conducted to ascertain.the capability of the battery to perform its design function under postulated j

accident condition. An excessive drop of voltage with respect to ti=e is indicative of required battery maintenance or replacement.

Testing of the e=ergency lighting is scheduled annually and is sub, ject to review and modification if experience demonstrates a more effective test schedule.

References FSAR, Section 8

~

. ~.

I Amendment No.

101

\\,

(,

h.7 REACTOR CO:iTROL ROD SYSTEM TESTS

/

..h.7.1' Control Rod Drive Syntem Punctional Tests Applicability; Applies to the surveillance of the control rod system.

Objective To assure operability of the control rod system.

Specification b.7 1.1 The-control rod trip insertion time shall be neasured for ecch control rod at either full flow or no flow condition: following each refuelir.c outage prior to return to pearcr. The maximum control rod trip incertion time for an operable centrol rod drive me:_anism, except for the Axial Power Shaping Rods (APSRs),

from the fully withdrawn positien to 3/4 insertion (loh inches travel) shr.ll not exceed 1.46' ceconds at reactor coolant full flow condition: or 1.20 seconda for no flow conditions.

For the APSRs it chall be dcmonstrated that loss of power vill not cause rod novement.

If the trip insertion ti=c above is not met, the red shall be declared inoperab,le.

h.7.1.2 If a control rod is micaligned with its group average by nore than an inclicated nine (9) inches, the rod shall be declared inoperable and the limits of fpecification 3.5.2.2 chall apply. The rod with the grentest nicaligr. tent chall be evaluated first. The position of a red declared inoperable due to micalign=ent shall not be in-cluded in ev=puting the average position of the group for determining the operability of rods with lesser cisalignmentc.

h.7 1.3 If a control rod cannot be exercised, or if it cannot be loented with absolute or relative position indications or in or out limit lights, O

the rod shall be declared to be inoperable.

Ences The control rod trip insertion time in the total elapsed time from power interruptien at the control rod drive brenhers until the control red has completed 104 inches of travel froa the fully withdrawn position. The specified trip tice is bated upon the safety analysis in PSAR, Scetion 14.

Each control rod drive techa.ninc chall be exercised by a movement of approx-instely.tvo (2) inches of travel every two (2) veeks. This requirement shall apply to either a partial or fully withdrawn control red at reactor operating conditions. Exercising the drive ucchanic=s'in this conner provides toscurance of reliability of the techanisms.

A rod 'ic considered' inoperable if it cannet be exercised, if the trip in-s

-certion tise. is.grester than the specified allovable tice, or if.the rod Amendment No.

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