ML20127E659

Amend 66 to License NPF-6,revising DNBR Limit Used by Core Protection Calculators & Modifying Table 2.2-2, Core Protection Calculator Addressable Constants

ML20127E659
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/07/1985
From:	John Miller Office of Nuclear Reactor Regulation
To:
Shared Package
ML20127E661	List: ML20127E659 ML20127E670
References
NUDOCS 8505200154
Download: ML20127E659 (20)

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ARKANSAS POWER & LIGHT COMPANY DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2

' AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 License No.- NPF-6 d

1.

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

A.-

The application for amendment by Arkansas Power & Light Company (the licensee)datedJanuary 25, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I;

.B.

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

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

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

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

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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. NPF-6 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 66, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION s

flier, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: May 7,1985 A

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e ATTACHMENT TO LICENSE AMENDMENT NO.66 FACILTIY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overlead pages are also provided to maintain document completeness.

Remove Pages Insert Pages 2-1 2-1

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.2-6 2-6 2-7 2-7 2-9 2-9 B 2-1 B 2-1 B 2-2 B 2-2 8 2-3 8 2-3 B 2-6 B 2-6 3/4 2-8 3/4 2-8 8 3/4 2-3 8 3/4 2-3 8 3/4 2-4 8 3/4 2-4 (repositioned)

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nj I's 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE ONBR-2.1.1.1 - The DN8R of the reactor core shall be maintained > l.25.

l APPLICABILITY: MODES I and 2.

ACTION:

Whenever the DNBR of the reactor core has decreased to less than 1.25 be in NOT STAND 8Y within I hour.

PEAK LINEAR HEAT RATE 2.1.1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of the fuel shall be maintained < 21.0 kw/ft.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21.0 kw/ft, be in HOT STAND 8Y within I hour.

I ARKANSAS - UNIT 2 2-1 AmedmentNo.pf,66 n

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a SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

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

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STAND 8Y with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its ifmit within 5 minutes.

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ARKANSAS - UNIT 2 2-?

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TABLE 2.2-1 REACTORPROTECTIVhl'NSTRUMENTATIONTRIPSETPOINTLIMITS' 5

Y' FUNCTIONAL UNIT E

TRIP SETPOINT ALLOWABLE VALUES Q

1.

Manual Reactor Trip Not Applicable Not Applicable m

2.

Linear Power Level - High Four Reactor Coolant Pumps 1 110% of RATED THERHAL POWER 1 110.712% of RATED THERMAL POWER I

a.

Operating b.

Three Reactor Coolant Pumps i

Operating Two Reactor Coolant Pumps c.

Operating - Same Loop d.

Two Reactor Coolant Pumps

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. Operating - Opposite Loops u.

3.

Logarithmic Power Level -

j.

High (1)

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0.75% of RATED THERMAL POWER 1

1 0.819% of RATED THERMAL POWER 4.

Pressurizer Pressure - High 1 2362 psia

< 2370.887 psia 5.

Pressurizer Pressure - Low 1 1766 psia (2) 1 1712.757 psia (2) j :

y 6.

Containment Pressure - High 1 18.4 psia s

1 19.024 psia 1

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7.

Steam Generator Pressure - Low 1 751 psia (3) o 1 729.613 psia (3) h 8.

Steam Generator Level - Low 1 23% (4) 1 22.111 (4) i, l

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• These values left blank pending NRC approval of safety analyses for operation with less than four reactor ri coolant pumps operating.

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TABLE 2.2-1 (Continued)'

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REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS 3l!

FUNCTIONAL UNIT TRIP SETPOINT

' ALLOWABLE VALUES 5

d; 9.

Local Power Density - High 1 20.3 kw/ft (5) 1 20.3 kw/ft (5)

10. DNBR - Low

> 1.25 (5)

> 1.25 (5) l' i

11. Steam Generator Level - High 1 93.7% (4) 1 94.589% '(4)

TABLE NOTATION (1) Trip may be manually bypassed above 10-4 1 -4 of RATED THERMAL POWER.1 of RATED. THERMAL POWER; bypass sha j

removed when THERMAL POWER is 1 0 (2)

Value may be decreased manually, to a minimum value of 100 psia, during a planned reduction in

?i pressurizer pressure, provided the margin between the pressurizer pressure and this value is maintained L

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

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7 the trip setpoint is reached.

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

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(3)

Value may be decre'ased manually during a planned reduction in steam generator pressure provided the

.t margin between the steam generator pressure and this value is :aaintained at I

2 1 00 psi; the setpoint I

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

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

2 (5)

A As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measure-ment, calgulational and processor uncertainties, and dynamic allowances.

below 10- % of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER g

i of RATED THERMAL POWER.

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TABLE 2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS 1.

TYPE I ADDRESSABLE CONSTANTS POINT ID PROGRAM ALLOWABLE NUMBER LABEL DESCRIPTION VALUE 60 FC1 Core coolant mass flow rate calibration 11 15 constant 61 FC2 Core coolant mass flow rate calibration 0.0 constant 62 CEAN0P CEAC/RSPT inoperable flag 0, 1, 2 or-3 63 TR Azimuthal tilt allowance 3.1.02 64 TPC Thermal power calibration constant 3,0.80 l

65 KCAL Neutron flux power calibration constant 3.0.60 l

66 DNBRPT DNBR pretrip setpoint Unrestricted

~ 67 LPDPT Local power density pretrip setpoint Unrestricted 98 TCREF Reference cold leg temperature 525 F TCREF 1 550F 5

104 PCALIB Secondary calorimetric power 1102.0'.

f ARKANSAS - UNIT 2 2-7 Amendment No, gg, 66 a

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4 TABLE 2.2-2 (Contin'ued)

CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS II. TYPE'II ADDRESSABLE CONSTANTS POINT ID PROGRAM NUMBER.

LABEL DESCRIPTION 68 SERR0 Thermal power uncertainty bias 69 BERR1 Power uncertainty factor used in DNBR calculation 70

-BERR2 Power uncertainty bias used in DNBR calculation 71 BERR3 Power uncertainty factor used in loc 1 power density calculation

'72-BERR4 Power uncertainty bias used in local power d nsity calculation 73 E0L End of life flag 174 ARM 1 Multiplier for planar radial peaking factor 75 ARM 2-Multiplier for planar radial' peaking factor 76 ARM 3 Multiplier for planar radial peaking factor 77 ARM 4 Multiplier for planar radial peaking factor

. 78 ARMS Multiplier for planar radial peaking factor 79 ARM 6 Multiplier for planar radial peaking factor 80 ARM 7 Multiplier for planar radial peaking factor 81 SC11 Shape annealing correction factor 82 SC12 Shape annealing correction factor 83 SCl3 Shape annealing correction factor 84 SC21 Shape annealing correction factor 85 SC22 Shape. annealing correction factor 86 SC23 Shape annealing correction factor 87 SC31 Shape annealing correction factor 88 SC32 Shape annealing correction factor

' ARKANSAS - UNIT 2 2-8 Amendmen; No. 24 c

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-..1 TABLE 2.2-2 (Continued)

CORE PROTECTION CALCULATOR-ADDRESSABLE CONSTANTS II. TYPE II ADDRESSABLE CONSTANTS (Continued)

POINT ID PROGRAM NUMBER LABEL DESCRIPTION 89 SC33 Shape annealing correction factor 90 PFMLTD DNBR penalty factor correction multiplier 91

.PFMLTL LPD penalty factor correction multiplier 92 ASM2 Multiplier for CEA shadowing factor 93 ASM3 Multiplier for CEA shadowing factor 94 ASM4 Maltiplier for CEA shadowing factor 95 ASMS

. Multiplier for CEA shadowing factor 96 ASM6 Multiplier for CEA shadowing factor 97 ASM7 Multiplier for CEA shadowing factor 99 BPPCC1 Boundary point power correlation coefficient 100 BPPCC2 Boundary point power correlation coefficient

--101 BPPCC3 Boundary point power correlation coefficient 102 BPPCC4 Boundary point power correlation coefficient 103 RPCLIM Reactor power cutback time limit ARKANSAS - UNIT 2 2-9 Anendment No. JA, 66 t-.

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2.1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.1.1 REACTOR CORE

.The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kw/ft which will not cause fuel centerline melting in any fuel rod.

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

transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature. The upper boundary of the nucleate boiling regime is tenned " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of-the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding. failure.

Correlations predict DNB and the location of DNB for axially unifonn and non-unifonn heat flux distributions.

The local DNB ratio (DNBR), defined as the ratio of the predicted DNB heat flux at a par-ticular core location-to the actual heat flux at that location, is iadicative of the margin to DNB. The minimum value of DNBR durina normal operational occurrences is limited to 1.25 for the CE-1 correlation l

and is established as a Safety Limit.

Second, operation with a peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cladding integrity. Above this peak linear heat rate level. (i.e., with some melting in the center), fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods. Volume changes which accompany the solid to liquid i

phase change are significant and require accomodation.

Another con-sideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting. Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit. To account for fuel rod dynamics (lags),

the directly indicated linear heat rate is dynamically adjusted.

Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High f

ARKANSAS - UNIT 2 B 2-1 Amendment No. H.66 l

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-SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Linear Power Level trips, and limiting conditions for operation on DNBR and kw/ft margin are'specified such that there is a high degree of confidence that the specified. acceptable fuel design limits are not exceeded durin nomal operation and design basis anticipated operational occurrences. g 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from.overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

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

(The reactor vessel, steam generators and pressurizer are designed to-the 1968 Edition,-

Sunner 1970 Addenda; piping to the 1971 Edition, original issue; and the valves to the 1%8 Edition, Winter 1970 Addenda.

Section III of this Code pemits a maximum transient pressum of 110% (2750 psia) of, design The Safety Limit of 2750 psia is therefore consistent with pressure.

the design criteria and associated code requirements.

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

2.2.1 REACTOR TRIP SETPOINTS

-. The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional t. nit. The Trip Setpoints have~been selected-to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating.

the consequences of accidents.

Operation with a trip set less conserva-tive than its Trip Setpoint but within its specified Allowable Value is cceptable on the basis that the difference between each Trip Setpoint a

nd the Allowable Value is equal to or les,s than the drift allowance a

ssumed for each trip in the safety analyses.

a The CNBR - Low and Local Power Density - High are digitally generated rip setpoints based on Limiting Safety System Settings of 1.25 and 20.3 l

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

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

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F ARKANSAS - UNIT 2 B 2-2 AnendmentNo./M,66 Ha,

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES

. To maintain the margins of safety assumed in the safety analyses,.the High trips include the measurement, calculational and pro and dynamic allowances as defined in CEN-147(S)-P, " Functional Design Speci-fication for a Core Protection Calculator," January 1981; CEN-148(S)-P,

" Functional Design Specification for a Control Element Assembly Calculator,"

January (A), "CPC Methodology Changes for Arkansas N 1981; CEN-296(A)-P.."ANO-2 CPC and CEAC Data Base Listing,"; and CEN-288 October 1984, which references CEN-281(S)-P, "CPC/CEAC Software Modi for San Onofre Nuclear Generating Station Units 2 and 3," June 1984, and -P to LO-82-039, "CPC/CEAC Software Modification for Syste:n 80,"

March 1982.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protec-tive instrumentation channels and provides manual reactor trip capability.

Linear Power Level-High rapid reactivity excursions which might occur as the resul This trip initiates a reactor trip at a linear power level of 1110.712% of RATED THERMAL POWER.

Logarithmic Power Level-High The Logarithmic Power Level - High trip is provided to protect the in the event of an unplanned criticality from a shutdown con trip is initiated by the Logarithmic Power Level - High trip at a THERMAL A reactor POWER level of bypassed by the operator.1 0.819% of RATED THERMAL POWER unless this trip is ma The operator may manually bypass this trip wheri the THERMAL POWER level is above 10-"% of RATED THERMAL POWER; th i

ATED THERMAL POWER.s automatically removed when the THERMAL POWER lev R

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ARKANSAS - UNIT 2 B'2-3 Amendment No. H,66 i

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

p BASES-i Pressurizer Pressure-High The Pressurizer Pressure-High trip, in conjunction with the pres-surizer safety valves and main steam safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip.

This trip's setpoint is at 1 2370.887 psia g

which is below the nominal lift setting (2500 psia) of the pressurizer safety valves and its operation avoids the undesirable operation of the pressurizer safety valves.

Pressurizer Pressure-Low The Pressurizer Pressure-Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a Loss of Coolant Accident.

During normal operation, this trip's setpoint is set at 2.1712.757 psia.

This trip's setpoint may be manually I

decreased, to a minimum value of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the. pres-surizer pressure and this trip's setpoint is maintained at 1 00 psi; 2

this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.

Containment Pressure-High 3

3 The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint for

{

this trip is identical to the safety injection setpoint.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant.

The setpoint is sufficiently below the full load operating point of approximately 900 psia so as not to inter-fare with nonnal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manuall uring plant shutdowris,y decreased as steam generator pressure is reduced d

provided the margin between the steam generator pressure and this trip's setpoint is maintained at 2

1 00 psi; this set-point increases automatically as steam generator pressure increases until the trip setpoint is reached.

ARKANSAS - UNIT 2 B 2-4 Amendment No. 49 i

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS L

BASES l

Steam Generator Level-Low The Steam Generator Level-Low trip provides protection against a loss of feedwater flow incident and assures that the design pressure of the Reactor Coolant System will not be exceeded due to loss of the steam generator heat sink.

This specified setpoint provides allowance that there will be sufficient water inventory in the steam generator at the time of the trip to provide sufficient margin before emergency feedwater is required.

Local Power Density-High The Local Power Density-High trip is provided to prevent the linear fuel design limit in the event of any anticipated operationa The local power density is calculated in the reactor protective system utilizing the following information:

Nuclear flux power and axial power distribution from the a.

excore flux monitoring system; b.

Radial peaking factors from the position measurement for the CEAs; I

AT power from reactor coolant temperatures and coolant flow c.

measurements.

I CPC incorporates uncertainties and dynamic compensation ro These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fue1~ design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the peak LPD Safety Limit.

CPC uncertainties related to peak LPD are the same types used for DNBR calculation.

Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes i n power density), sensor time delays, and protection system equipment ime delays.

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ARKANSAS - UNIT 2 B ?-5 Amendment No. 24 l

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DNBR-Low j

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d The DNBR - Low trip is provided to prevent the DNBR in the limiting ji coolant channel in the core from exceeding the fuel design limit in the event of anticipated operational occurrences. The DNBR - Low trip incor-porates a low pressurizer pressure floor of 1750 psia. At this pressure a

a DNBR - Low trip will automatically occur.

The DNBR is calculated in o

the CPC utilizing the following information:

a a.

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

Reactor Coolant System pressure from pressurizer pressure measurement;

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Differential temperature (AT) power from reactor coolant c.

temperature and coolant flow measurements; d.

Radial peaking factors from the position measurement for the CEAs; Reactor coolant mass flow rate from reactor coolant pump speed; e.

o f.

Core inlet temperature from reactor coolant cold leg temperature measurements.

4 M

The DNBR, the trip variable, calculated by the CPC incorporates various uncertainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits. These uncertainties 4

and dynamic compensation routines ensure that a reactor trip occurs when the actual core DNBR is sufficiently greater than 1.25 such that the l

decrease in actual core DNBR after the trip will not result in a viola-U-

tion of the DNBR Safety Limit. CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithn modelling uncertainties, and computer equipment _ processing uncertainties.

Dynamic compensation is provided in the CPC calculations for the effects of coolant transport P-delays, core heat flux delays (relative to changes in core power), sensor

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time delays, and protection system equipment time delays.

The DNBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a i

CPC initiated trip.

ARKANSAS - UNIT 2 B 2-6 AmendmentNo.//f,66

POWER DISTRIBUTION LIMITS DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The DNBR margin shall be maintained by operating within the region of acceptable operation of Figure 3.2-3 on 3.2-4, as applicable.

APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER.

ACTION:

With operation outs'ide of the region of acceptable operation, as indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on DNBR; or (2) when DNBR~ limit, within 15 minutes initiate corrective action to r DNBR to within the limits and either:

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

b.

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

SURVEILLANCE REQUIREMENTS ,4.2.4.1 The provisions of' Specification 4.0.4 are not applicable.

- 4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously.

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

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

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ARKANSAS - UNIT 2 3/4 2-7 Amendment No. 79, 49 j

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POWER DISTRIBUTION LIMITS BASES ~

Ptilt /Puntilt is the ratio of the power at a core location in the

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

3/4.2.4 DNBR MARGIN

'The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety analysis assumptions and which have been analytically demonstrated adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences, of which the loss of flow transient is the most limiting.

tion of the core with a DNBR at or above this limit provides assurance thatOpera-an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.

Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory S Protection Calculators (CPCs)ystem (COLSS) and the DNBR channels in the Core

, provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate its limits.

The COLSS performs this function by continuously monitoring the core power distributfor allowable minimum! and calculating a core operating limit corresponding to the DNBR.

level assures that the limits of Figure 3.2-3 are not violated. Reactor oper 4

The COLSS e.

calculation of core power operating limit based on DNBR includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that-the core power at which a DNBR of less than 1.25 could occur, as 1,

?

calculated by COLSS, is less than or equal to that which would actually be l

required in the core.

ftained, the COLSS computer program includes an FTo ensure that the design

" factor of 1.053, an engineering uncertainty factEf of 1.03, a THERMAL POWER measurement uncertainty measurement uncertainty factor of 1.02 and appropriate uncertainty and penalty 1'

factors for flux peaking augmentation and rod bow.

Parameters required to maintain the margin to DNB and total core power areialso monitored by the CPCs.

Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-4 can be maintained by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels.

factors are' also included in the CPC.The above listed uncertainty and penalty A DNBR penalty factor has been included in the COLSS and CPC DNBR calculations to acconsnodate the effects of: rod -bow. The amount of rod i

1 in each assembly is dependent upon the average burnup experienced by that assembly.

Fuel assemblies that incur higher average burnup will experience

'f, a greater magnitude of rod bow.

Conversely, lower burnup assemblies will experience less rod bow.

In design calculations, the penalty for each batch required to compensate for rod bow is determined from a batch's maximum average _ assembly burnup applied to the batch's maximum integrated planar-radial power peak. A single net penalty for COLSS and CPC-is then determined h

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. AR.%NSAS - UNIT 2 B 3/4 2-3 Amendment No. 2A, 28, 32,66

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_P0WER DISTRIBUTION LIMITS BASES 1

a margins due to the lower radial power peaks in the 3/4.2.5 RCS FLOW RATE rate is maintained at 'or above the minimum value analyses.

3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE coolant cold leg temperature is maintained within th the safety analyses.

3/4.2.7. AXIAL SHAPE INDEX s

3-

- This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of values used in the 3/4.2.8 PRESSURIZER PRESSURE This specification is provided to ensure that the actual value of safety analyses. pressurizer pressure is maintained within the range of value s

i-t ARKANSAS - UNIT 2 B 3/4 2-4 Amendment No. 24, 32,66

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