ML20204C425

Amend 79 to License NPF-6,revising Tech Specs Re Core Protection Calculators (CPC) as Part of CPC Improvement Program

ML20204C425
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/22/1986
From:	Richard Lee Office of Nuclear Reactor Regulation
To:
Shared Package
ML20204C432	List: ML20204C425 ML20204C436
References
NUDOCS 8607310192
Download: ML20204C425 (34)

Text

'

gaucoq(o UNITED STATES c[' 3 ([g NUCLEAR REGULATORY COMMISSION g

E WASHINGTON. D. C. 20555

• N/,5

• g% y ARKANSAS POWER & LIGHT COMPANY DOCKET N0. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 79 License No. NPF-6 1.

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

A.

The application for amendment by Arkansas Power & Light Company (the licensee) dated February 27, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance (1) 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 Commission'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 Connission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions 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. 79, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

8607310192 860722 PDR ADOCK 05000368 P

PDR

_2-3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert S. Lee, Project Manager PWR Project Directorate No. 7 Division of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 22, 1986

ATTACHMENT TO LICENSE AMENDMENT N0. 79 FACILITY OPERATING LICENSE NO. NPF-6 DOCKET N0. 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 overleaf pages are also provided to maintain document completeness.

4 Remove Pages Insert Pages 2-6 2-6 B 2-2 B 2-2 B 2-3 B 2-3 B 2-6 B 2-6 B 2-7.

B 2-7 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-7 3/4 2-7 3/4 2-7a 3/4 2-7a 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4 2-10a 3/4 2-10a 3/4 3-5 3/4 3-5 3/4 3-Sa 3/4 3-Sa 3/4 3-6a 3/4 3-6a 3/4 3-6b 3/4 3-6b 3/4 3-6c 3/4 3-6c B 3/4 2-1 B 3/4 2-1 B 3/4 2-3 B 3/4 2-3 B 3/4 3-1 B 3/4 3-1 l '

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TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS a.

2 T'

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E

G 1.

Manual Reactor Trip Not Applicable Not Applicable m

2.

Linear Power Level - High a.

Four Reactor Coolant Pumps 1 110% of RATED THERNAL POWER 1 110.712% of RATED THERMAL POWER Operating b.

Three Reactor Coolant Pumps Operating c.

Two Reactor Coolant Pumps Operating - Same Loop d.

Two Reactor Coolant Pumps

{

Operating - Opposite Loops 3.

Logarithmic Power Level -

High (1) 1 0.75% of RATED THERMAL POWER 5 0.819% of RATED THERMAL POWER 4.

Pressurizer Pressure - High

< 2362 psia

< 2370.887 psia

{

5.

Pressurizer Pressure - Low 1 1766 psia (2)

> 1712.757 psia (2) 6.

Containment Pressure - High

$ 18.4 psia 1 19.024 psia 7.

Steam Generator Pressure - Low

> 751 psia (3)

> 729.613 psia (3) 8.

Steam Generator Level - Low

> 23% (4)

> 22.111 (4)

.M S

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

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 9.

Local Power Density - High 121.0 kw/ft (5) 121.0 kw/ft (5) 10 DNBR - Low 11.25 (5) 11.25 (5) 11.

Steam Generator Level - High 193.7% (4) 1 4.589% (4) 9 TABLE NOTATION (1) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is i 10-4 of RATED THERMAL POWER.

(2) Value may be decreased manually, to a minimum value of 100 psia, during a planned reduction in pressurizer pressure, provided the margin between the pressurizer pressure and this value is maintained at 1200 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is 1 500 psia.

(3) Value may be decreased manually during a planned reduction in steam generator pressure provided the margin between the steam generator pressure and this value is maintained at i 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

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

(5) As stored within the Core Protection Calculator (CPC).

Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances.

Trip may be manually bypassed below 10 4% of RATED IHERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 10 4% of RATED THERMAL POWER.

ARKANSAS - UNIT 2 2-6 Amendment No. 2#, 26, 27, gg, 79 O

e

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 pcrforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by (1) restricting fuel' operati.on 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 termed " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.

Correlations predict DNB and the location of DNB for axially uniform and non-uniform 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 indicative 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 j

would cause fuel centerline melting maintains fuel rod and cladding i

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 phase change are significant and require accomodation.

Another con-i l

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 ARKANSAS - UNIT 2 B 2-1 Amendment No. N,66

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Linear Power Level trips, and limiting conditions for operation on DNBR and kw/f t margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.

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

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, Summer 1970 Addenda; piping to the 1971 Edition, original issue; and the valves to the 1968 Edition, Winter 1970 Addenda.

Section III of this Code permits a maximum transient pressure of 110% (2750) psia) of design

. ores sure.

The Safety Limit of 2750 psia is therefore consistent with 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 unit.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety 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 acceptable on the basis that the difference between each Trip Setpoint and tne Alicwable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.25 and 21.0 l

kw/ft, respectively.

Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment.

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

i ARKANSAS - UNIT 2 B 2-2 Amendment No. 24, 66. 79

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density - High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CEN 305-P, " Functional Design Requirement for a Core Protection Calculator," July 1985; CEN-304-P,

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

July 1985; CEN-310-P, "CPC and Methodology Changes for the CPC Improvement Program," October 1985 and CEN-308-P, "CPC/CEAC Software Modifications for the CPC Improvement Program," August 1985.

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

Linear Power Level-High The Linear Power Level-High trip provides reactor core protection against rapid reactivity excursions which might occur as the result of an ejected CEA.

This trip initiates a reactor trip at a linear power level of

< 110.712% of RATED THERMAL POWER.

Logarithmic Power Level-High The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an unplanned criticality from a shutdown condition.

A reactor trip is initiated by the Logarithmic Power Level - High trip at a THERMAL POWER level of < 0.819% of RATED THERMAL POWER unless this trip is manually bypassed by the operator.

The operator may manually bypass this trip when the THERMAL POWER level is above 10-4% of RATED THERMAL POWER; this bypass is automatically removed when the THERMAL POWER level decreases to 10-4% of RATED THERMAL POWER.

ARKANSAS - UNIT 2 B 2-3 Amendment No. 49, 66, 72

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 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 2

1 370.887 psia g

which is below the nominal l'ft 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 1 200 psi; this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.

Containment Pressure-High 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 excesrive 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-fere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow.

This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is maintained at 1 00 psi; this set-2 point increases automatically as steam generator pressure increases until the trip setpoint is reached.

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

D SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 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 heat rate (kw/ft) in the limiting fuel rod in the core from exceeding the fuel design limit in the event of any anticipated operational occurrence.

The local power density is calculated in the reactor protective system utilizing the following information:

a.

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

Radial peaking factors from the position measurement for the CEAs; c.

AT power from reactor coolant tenperatures and coolant flow measurements.

The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines.

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

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

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

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

i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES DNBR - Low The DNBR - Low trip is provided to prevent the DNBR in the limiting 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.

At this pressure a DNBR - Low trip will automatically occur.

The DNBR is calculated in the CPC utilizing the following information:

a.

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

Reactor Coolant System pressure from pressurizer pressure measurement; c.

Differential temperature (AT) power from reactor coolant temperature and coolant flow measurements; d.

Radial peaking factors from the position measurement for the CEAs; e.

Reactor coolant mass flow rate from reactor coolant pump speed; t.

Core inlet temperature from reactor coolant cold leg temperature measurements.

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 and dynamic compensation routines ensure that a reactor trip occurs when the actual core DNBR is sufficiently greater than 1.25 such that the decrease in actual core DNBR after the trip will not result in a viola-tion of the DNBR Safety Limit.

CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties.

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

ARKANSAS - UNIT 2 B 2-6 Amendment No. 24, (s. 79

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES a.

RCS Cold Leg Temperature-Low

> 490 F b.

RCS Cold Leg Temperature-High 5585F c.

Axial Shape Index-Positive Not more positive than +0.6 d.

Axial Shape Index-Negative Not more negative than -0.6 e.

Pressurizer Pressure-Low

> 1785 psia f.

Pressurizer Pressure-High 52415 psia g.

Integrated Radial Peaking Factor-Low h.

Integrated Radial Peaking

-> 1.28 Factor-High

< 4.28 i.

Quality Margin-Low 50 Steam Generator Level-High The Steam Generator Level-High trip is provided to protect the turbine from excessive moisture carryover.

Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over.

This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip.

Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

i

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l ARKANSAS - UNIT 2 8 2-7 Amendment No. 24,,42, //, /9

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate limit shall be maintained by either:

a.

Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on linear heat rate (when COLSS is in service); or b.

Operating within the region of acceptable operation of Figure 3.2-1 using any operable CPC Channel (when COLSS is out of service).

APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER ACTION:

With the linear heat rate limit not being maintained as indicated by either:

1.

COLSS calculated core power exceeding COLSS calculated core power operating limit based on linear heat rate; or 2.

Operation outside the region of acceptable operation in Figure 3.2-1, when COLSS is out of service, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:

a.

Restore the linear heat rate to within its limits within one hour, or b.

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

SURVELLLANCE REQUIREMENTS

4. 2.1.1 The provisions of Specification 4.0.4 are not applicaule.

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

4.2.1.3 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 linear heat rate.

ARKANSAS - UNIT 2 3/4 2-1 Amendment No. f4' 79

THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 3/4 2-2 Amendment No. 24, 79

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POWER DISTRIBUTION LIMITS RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION 3.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F* ) shall be less than or y

c equal to the PLANAR RADIAL PEAKING FACTORS (F Limit Supervisory System (COLSS) and in the Cd/e) used in the Core Operating Protection Calculators (CPC).

. APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER.*

ACTION:

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

I c

With a F*, exceeding a corresponding F y a.

Adjust the CPC addressable constants to increase the multiplier applied to PLANAR RADIAL PEAKING FACTOR by a factor equivalent to >F* /F and restrict subsequent operation so that a margin y

to the COLSS operating limits of at least ((F" / Fc ) - 1.0] x 100%

is maintained; or Y

• Y b.

Adjust the affected PLANAR RADIAL PEAKING FACTORS (Fc ) used 'in l

y the COLSS and CPC to a value greater than or equal to the measured PLANAR RADIAL PEAKING FACTORS (F* ); or l

c.

Be in at least H0T STANDBY.

g SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F* ), obtained by l

y using the incore detection system, shall be determined to be less than C

or equal to the PLANAR RADIAL PEAKING FACTORS (F y) used in the COLSS l

and CPC at the following intervals:

a.

After each fuel loading with THERMAL POWER greater than 40%

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

At least once per 31 days of accumulated operation in MODE 1.

• See Special Test Excr.ption 3.10.2.

ARKANSAS - UNIT 2 3/4 2-4 Amendment No. 24 l

l l

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POWER DISTRIBUTION LIMITS 3/4.2.4 DNBR MARGIN LIMITING CONDITIONS FOR OPERATION 3.2.4 The DNBR limit shall be maintained by one of the following methods:

a.

Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR (when COLSS is in service, and at least one CEAC is operable); or b.

Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit ~ based on DNBR decreased by 13.0% (when COLSS is in service and neither CEAC is operable); or c.

Operating within the region of acceptable operation of Figure 3.2-2 using any operable CPC channel (when COLSS is out of service and at least one CEAC is operable); or d.

Operating within the region of acceptable operation of Figure 3.2-3 using any operable CPC channel (when COLSS is out of service and neither CEAC is operable).

APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER.

i ACTION:

With the DNBR limit not being maintained as indicated by either:

1.

COLSS calculated core power exceeding COLSS calculated core power operating limit based on DNBR; or 2.

Operation outside the region of acceptable operation of Figure 3.2-2 or 3.2-3 as applicable, when COLSS is out of service, within 15 minutes initiate corrective action to restore the DNBR to within the li.riits,, d either:

a.

Re. tore the DNBR to within its limits within one hour, or b.

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

SURVEILLANCE REOUIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.

l ARKANSAS - UNIT 2 3/4 2-7 Amendment No. M // 79 l

i

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (continued) 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 verifying at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the DNBR, as indicated on any OPERABLE CPC channel, is within the limit shown on Figures 3.2-2 or 3.2-3, as applicable.

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

ARKANSAS - UNIT 2 3/4 2-7a Amendment No. 79

4 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS DELETED ARKANSAS - UNIT 2 3/4 2-8 Amendment No. 24, 26, 32, 66

4 THIS PAGE LEFT BLANK INTENTIONALLY i

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i ARKANSAS - UNIT 2 3/4 2-9 Amendment No. 24, 79 l

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CORE AVERAGE AXIAL SHAPE INDEX s*

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1

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TABLE 3.3-1 (Continued)

ACTION STATEMENTS I

l ACTION 2 With the number of OPERABLE channels one less than the 1

Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

For the purposes of testing and maintenance, the inoperable 1

channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition, b.

Within one hour, all functional logic units receiving l

an input from the inoperable channel are also placed l

in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel, c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.

ACTION 3 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

ACTION 4 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, place the inoperable channcl in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HOT STANDBY within 6 hcurs; however, one channel may be bypassed for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for surveillance testing per Specification 4.3.1.1.1 ACTION 5 -

a.

With one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, each CEA is verified to be within 7 inches (indicated position) of all other CEAs in its group.

After 7 days, operation may continue provided that ACTION 5.b is met.

ARKANSAS - UNIT 2 3/4 3-5 Amendment No. 79

TABLE 3.3-1 (Continued)

ACTION STATEMENTS b.

With both CEACs inoperable, operation may continue provided that:

1.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the margin required by Specification 3.2.4.b (COLSS in service) or Specification 3.2.4.d (COLSS out of service) is satisfied.

2.

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

a)

All full length and part length CEA groups are withdrawn to and subsequently maintained at the " Full Out" position, except during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2 or for control when CEA group 6 may be inserted no further than 127.5 inches withdrawn.

b)

The "RSPT/CEAC Inoperable" addressable constant in the CPCs is set to both CEACs l

inoperable.

c)

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

3.

At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, all full length and part length CEAs are verified fully withdrawn, except as permitted by 2. a) above, then verify at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> that the inserted CEAs cea alfared within 7 inches (indicated les' tion) of all r~i!ec CEAs in their group.

ACTION 6 -

With three or more auto restarts of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL FUNCTIONAL TEST within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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ARKANSAS - UNIT 2 3/4 3-5a Amendment No. 24, 41, 79

g TABLE 3.3-2 9g REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES R

FUNCTIONAL UNIT RESPONSE TIME N

1.

Manual Reactor Trip Not Applicable 2.

Linear Power Level - High 1 0.40 seconds

• 3.

Logarithmic Power Level - High 1

40 seconds

• 0 4.

Pressurizer Pressure - High 1 0.90 seconds 5.

Pressurizer Pressure - Low 1 0.90 seconds 6.

Containment Pressure - High 1 1.59 seconds

[

7.

Steam Generator Pressure - Low 1 0.90 seconds 8.

Steam Generator Level - Low 1 0.90 seconds 9.

Local Power Density - High a.

Neutron Flux Power from Excore Neutron Detectors

-< 2.58 seconds

• b.

CEA Positions 1.58 seconds **

9 9

TABLE 3.3-2 (Continued)

REACT 0!!_P"0TECTIVE INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

10. DNBR - Low a.

Neutron Flux Power from Excore Neutron Detectors

< 0.39 seconds

• b.

CEA Positions 7 1.09 seconds **

c.

Cold Leg Temperature 53.79 seconds ##

d.

Hot Leg Temperature 5 1.54 seconds ###

l e.

Primary Coolant Pump Shaft Speed 1 0.80 seconds #

f.

Reactor Coolant Pressure from Pressurizer 1 3.19 seconds

11. Steam Generator Level - High Not Applicable

• Neutron detectors are exempt from response time testing.

Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

• • Response time shall be measured from the onset of a single CEA drop.

1. Response time shall be measured from the onset of a 2 out of 4 Reactor Coolant Pump coastdown.

1. 1. Based on an effective resistance temperature detector (RTD) response time of s 8.0 seconds.

If the effective RTD time constant for a CPC channel exceeds 8.0 seconds, the DNBR and LPD penalties for the affected channel (s) shall be increased by the amount indicated on Figure 3.3-1.

1. 1. 1. Based on an effective RTD response time of 5 13.0 seconds.

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ARKANSAS-UNIT 2 3/4 3-6a Amendment No. 88, 79

THIS PAGE LEFT BLANK INTENTIONALLY ARKANSAS - UNIT 2 3/4 3-6b Amendment No. 77. 79

FIGURE 3.3-1 CPC PENALTY VS. EFFECTIVE RTD TIME CONSTANT

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5 6

7 8

9 10 11 12 13 14 EFFECTIVE COLD LEG RTD TIME CONSTANT (SEC.)

ARKANSAS - UNIT 2 3/4 3-6c Amendment No. jf, 79 l

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3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 F.

Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits.

The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating limit corresponding to the allowable peak l

linear heat rate.

The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator.

A COLSS alarm is annunciated in the event that the core power exceeds the core power operating limit.

This provides adequate margin to the linear heat rate operating limit for normal steady state operation.

Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power operating limit being exceeded.

In the event this occurs, COLSS alarms will be annunciated.

If the event which causes the COLSS limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation.

The COLSS calculation of the linear heat rate limit includes appropriate uncertainty and penalty factors necessary to provide a 95/95 confidence level that the maximum linear heat rate calculated by COLSS is greater than or equal to that existing in the core.

To ensure that the design margin to safety is maintained, the COLSS computer program includes an F measurement uncertainty factor of 1.053, an engineering uncertainly factor of 1.03, a THERMAL POWER measurement uncertainty factor of 1.02 and appropriate uncertainty and penalty factors for rod bow.

l Parameters required to r..aintain the operating limit power level basea on linear heat rate, margin ta DNB and.stal core power are also monitored by the CPCs.

Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-1 can be maintained by utilizing a l

predetermined local power density margin and a total core power limit in the

(

CPC trip channels.

The above listed uncertainty and penalty factors are I

also included in the CPCs.

j ARKANSAS - UNIT 2 8 3/4 2-1 Amendment No. #, 79

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 RADIAL PEAKING FACTORS Limiting the values of the PLANAR RADIAL PEAKING FACTORS (F ) used in the COLSS and CPCs to values equal to or greater than the measured PLANAR RADIAL PEAKING FACTORS (F* ) provides assurance that the limits calculated y

by COLSS and the CPCs remain valid.

Data from the incore detectors are used for determining the measured PLANAR RADIAL PEAKING FACTORS. The periodic surveillance requirements for determining the measured PLANAR RADIAL PEAKING FACTORS provides assurance that the PLANAR RADIAL PEAKING FACTORS used in COLSS and the CPCs remain valid throughout the fuel cycle.

Determining the measured PLANAR RADIAL PEAKING FACTORS after each fuel loading prior to exceeding 70% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

3/4.2.3 AZIMUTHAL POWER TILT - T q The limitations on the AZIMUTHAL POWER TILT are provided to ensure that design safety margins are maintained.

An AZIMUTHAL POWER TILT greater than 0.10 is not expected and if it should occur, operation is restricted to only those conditions required to identify the cause of the tilt. The tilt is nonnally calculated by COLSS.

The surveillance requirements specified when COLSS is out of service provide an acceptable means of detecting the presence of a steady state tilt.

It is necessary to explicitly account for power asymmetries because the radial peaking factors used in the core power distribution calculations are based onan untilted power distribution.

AZIMUTHAL POWER TILT is measured by assuming that the ratio of the power at any core location in the presence of a tilt to the untilted power at the location is of the form:

/P

=1+T g cos (e - e )

tilt untilt q

g where:

T is the peak fractional tilt amplitude at the core periphery q

g is the radial normalizing factor e is the azimuthal core locatio~n e, is the azimuthal core location of maximum tilt ARKANSAS - UNIT 2 B 3/4 2-2 Amendment No. 24 l

l

POWER DISTRIBUTION LIMITS BASES P

/P is the ratio of the power at a core location in the presenbhlbf S"k lk 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.

Operation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of any anticipated operational occurrence.

Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core Protection Calculators (CPCs), 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 distribution and calculating a core operating limit corresponding to the allowable minimum DNBR.

The COLSS calculation of core power operating limit based on DNBR includes appropriate uncertainity 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 calculated by COLSS, is less than or equal to that which would actually be required in the core.

To ensure that the design margin to safety is maintained, the COLSS computer program includes an F measurement uncertainty factor of 1.053, an engineeringuncertain((factorof1.03,aTHERMALPOWERmeasurement uncertainty factor of 1.02 and appropriate uncertainty and penalty factors for rod bow.

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

Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-2 for CEAC operable or Figure 3.2-3 for both CEACs inoperable can be ma'.1. aired by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX cod by monitoring the CPC trip channels.

The above listed uncertainty and penalty factors are also included in the CPC.

A DNBR penalty factor has been included in the COLSS and CPC DNBR calculations to accommodate the effects of rod bow.

The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly.

Fuel assemblies that incur higher average burnup will experience 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

~

ARKANSAS - UNIT 2 B 3/4 2-3 Amendment No. 24, 26, 32, %%. 79 l

1

POWER DISTRIBUTION LIMITS BASES from the penalties associated with each batch, accounting for the offsetting I

margins due to the lower radial power peaks in the higher burnup batches.

3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the LOCA safety analyses.

3/4.2.6 REACTOR C0OLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the range of values used in the safety analyses.

3/4.2.7 AXIAL SHAPE INDEX This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses.

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

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ARKANSAS - UNIT 2 B 3/4 2-4 Amendment No. 2/, 32,66 l

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective ano ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The surveillance requirements specificd for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The measurement of response time ct the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with ec 'ufiec response times.

RTD response time is defined as the time interval required for the RTD output to achieve 63.2% of its total change when subjected to a step change in RTD temperature. The RTD response time for the Core Protection Calculator System (CPCS) is expressed as an effective time constant.

For hot leg temperatures, the effective time constant for a given CPC channel is defined as the mean time constant for averaged pairs of hot leg RTD inputs to the channel.

This is done because the CPCS utilizes the mean hot leg temperature in its calculations.

The maximum hot leg effective time constant allowable for use in the CPCS is 13.0 seconds.

For cold leg temperatures, the effective time constant to be used in Figure 3.3-1 is the maximum time constant of the two cold leg RTD inputs for a given channel.

The CPCS utilizes the more conservative cold leg temperature in the various DNBR and LPD calculations.

The maximum cold leg effective time constant allowable for use in the CPCS is 13.0 seconds.

ARKANSAS - UNIT 2 B 3/4 3-1 Amendment No. 77, 79 I

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3/4.3 INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magniture of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12. " Instrumentation for Earthquakes,"

April 1974.

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient metecrological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "0nsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of H0T STANDBY of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

ARKANSAS - UNIT 2 B 3/4 3-2 Amendment No. 33