ML20101U345

Proposed Changes to Tech Specs 2.1.1.1,2.2.1 & Tables 2.2-1,2.2-2 & 4.2.4-4 Revising DNBR Limit

ML20101U345
Person / Time
Site:	Arkansas Nuclear
Issue date:	01/28/1985
From:	ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML19269A993	List: 2CAN018505, Application to Amend License NPF-6,revising DNBR Limit. Nonproprietary & Proprietary CEN-289(A), Revised Rod Bow Penalties..., Supporting Changes Encl.Proprietary Rept Withheld (Ref 10CFR2.790).Fee Paid ML20101U345 ML20101U355
References
NUDOCS 8502070269
Download: ML20101U345 (14)

Text

7 PROPOSED TECHNICAL SPECIFICATIONS

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE DNBR 2.1.1.1 The DNBR of the reactor core shall be maintained > 1.25.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the DNBR of the reactor core has decreased to less than 1.25, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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 STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ARKANSAS - UNIT 2 2-1~ Amendment No. 24' l

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 then 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 reigme

.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 possiblity of cladding failure.

- Correlations predict DN8 and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DN8R),

defined as the ratio of.the predicted DNB heat flux at a particular core location to the actual heat flux 'at- that location,-is. indicative of the margin to DNB. The minimum value of DNBR during normal operational occurrences is limited to 1.25 for the CE-1 correlation 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 phase change are significant and require' accommodation. Another consideration 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 DN8R, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High ARKANSAS - UNIT 2 8 2-1 Amendment No. 24 1

o; l SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Linear Power Level trips, and limiting conditions for operation on DN8R 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 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 pressure. 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 c .; , -

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 conservative _than its Trip Setpoint but within its specified Allowable Value is acceptable on

. the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift a'lowance assumed for each trip in the safety analyses.-

I The DN8R.- Low and Local Power Density - High are digitally generated i' trip setpoints based on Limiting Safety System Settings of 1.25 and'20.'3 p kw/ft, respectively. Since these trips are digitally _ generated by the Core Protection Calculators, the trip values are not subject to drifts common to S trips generated by analog type equipment. The Allowable ValuesLfor these-trips are therefore the same as the Trip Setpoints.

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ARKANSAS - UNIT 2 8 2-2 Amendment No. 24-

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 DN8R - Low and Local Power Density - High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CEN-147(S)-P, " Functional Design Specification for a Core Protection Calculator," January 1981; CEN-148(S)-P, " Functional Design Specification for a Control Element Assembly Calculator," January 1981; CEN-296(A)-P, "ANO-2 CPC and CEAC Data Base Listing,"; and CEN-288(A) "CPC Methodology Changes for Arkansas Nuclear One Unit 2 Cycle 5" October 1984 which references CEN-281(S)-P, "CPC/CEAC Software Modifications for San Onofre Nuclear Generating Station Units 2 and 3," June 1984 and Enclosure 1-P to LD-82-039, "CPC/CEAC Software Modification for System 80," March 1982.

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 1 110.712% 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. r may manually bypass this trip when the THERMAL POWER level isTheoperatg%ofRATEDTHERMALPOWER; above 10 this bypass is automatically removed when the THERMAL POWER level decreases to 10 % of RATED THERMAL POWER.

ARKANSAS - UNIT 2 B 2-3 Amendment No. 49

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure-High The Pressurizer Pressure-High trip, in conjunction with the pressurizer 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 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 Pressurt-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 1712.757 psia. This trip's setpoint may be manually decreased, to a minimu?_m value of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer 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 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 interfere 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 1200 psi; this setpoint increases automatically as steam generator pressure increases until the trip setpoint is reached.

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

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 incorporates a low pressurizer pressure floor of 1750 psia. 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;

' f. 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 violation of the DNBR Safety Limit. CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modeling 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.

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ARKANSAS - UNIT 2 B 2-6 Amendment No. 24 L

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SAFET LIMIT 5ANDLIMITINGSAFETYSYSTEMSETTINGS BASES

a. RCS Cold Leg Temperature - Low > 465*F

b. RCS Cold Leg Temperature - High < 605*F

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 > 1750 psia

f. Pressurizer Pressure - High 52400 psia

g. Integrated Radial Peaking ,

Factor - Low -> 1.28

h. Integrated Radial Peaking Factor - High < 4.28

'i. Quality Margin - Low E0 Steam Generator Level-High The Steam Generator Level-High trip-is provided to protect the turbine from excessive moisture carry over.- 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.

2.2.2 CPC Addressable Constants The Core Protection' Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate-indications such as calorimetric measurements for power level and RCS flowrate and incore detector signals for~ axial flux shape, radial peaking factors.and CEA deviation penalties. Other CPC addressable ' constants allow penalization of the calculated DNBR'and LPD values based on measurement uncertainties or inoperable equipment. Administrative controls on changes and periodic checking of addressable constant values'(see also Technical Specifications 3.3.1.1 and 6.8.1) ensures that inadvertent misloading is unlikely.' The methodology for determination of CPC addressable constant values is described in MSS-NA2-P, " Arkansas Nuclear One-Unit 2 Core Protection Calculator Addressable Constant Determination Methodology" dated August

- 1981.

ARKANSAS ' UNIT 2 B 2-7 Amendment No.24,49 e

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TABLE 2.2-1 (Continued)

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

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

10. DNBR - Low 1 1.25 (5) 1 1.25 (5)

11. Steam _ Generator Level - High 5 93.7% (4) 1 94.589% (4)

TABLE NOTATION

~4 (1) Trip may be manually bypassed above_10 % of RATED THERMAL POWER: bypass shall be automatically removed when THERMAL POWER is 5 10 T 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 < 200 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 1 200 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 measuremeg% t, of calculational and processor uncertainties, and dynamic allowances. Trip may be manua} y bypassed below 10 RATEDTHERMALPOWER;bypassshallbeautomaticallyremovedwhenTHERMALPOWERis110}%ofRATEDTHER l

ARKANSAS - UNIT 2 2-6 Amendment No. 24, 26, 27

TABLE 2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS I. TYPE ~I ADDRESSABLE CONSTANTS POINT ID. PROGRAM ALLOWABLE NUMBER LABEL DESCRIPTION VALUE 60 FC1 Core coolant mass flow rate -<1.15 calibration constant 61 FC2 Core coolant mass flow rate 0.0 calibration constant 62 CEAN0P CEAC/RSPT inoperable flag 0,1,2 or 3 63- TR Azimuthal tilt allowance 11.00 l 64 TPC Thermal power calibration >0.80 l constant 65 KCAL Neutron flux power calibration >0.60 constant 66 DNBRPT DNBR pretrip setpoint Unrestricted 67 LPDPT Local power density pretrip Unrestricted setpoint 98 TCREF Reference cold leg temperature 525*F < TCREF 1 555'T 104 PCALIB- Secondary calorimetric power i 102.0%

ARKANSAS -' UNIT 2 2-7 Amendment No. 24 4

TABLE 2.2-2 (Continued)

CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS II. TYPE II ADDRESSABLE CONSTANTS POINT ID- PROGRAM NUMBER LABEL DESCRIPTION 68 BERR0 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 local power density calculation 72 BERR4 Power uncertainty bias used in local power

-density calculation 73 EOL End of life flag 74 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 L-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- SC13 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 ARKANSAS - UNIT 2 2-8 Amendment No. 24

e TABLE 2.2-2 (Continued)

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

' POINT ID PROGRAM NUMBER LABEL DESCRIPTION 88 SC32 Shape annealing correction factor 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 Multiplier 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 RPCLM Reactor power cutback time limit ARKANSAS - UNIT 2 2-9 Amendment No. 24 u

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

THIS PAGE LEFT BLANK INTENTIONALLY i-ARKANSAS - UNIT 2 3/4 2-8 Amendment No. 2#, 26, 32

POWER DISTRIBUTION LIMITS BASES P /P tilt untilt is the ratio of the power at a core location in the presence of a tilt to the power at the 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. 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 a loss of flow transient.

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. Reactor operation at or below this calculated power level assures that the limits of Figure 3.2-3 are not l violated. 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 desiga margin to safety is maintained, the COLSS computer program includes an F measurement uncertaintyfactorof1.053,anengineeringuncertaintyfactor*8f1.03,a THERMAL POWER measurement uncertainty factor of 1.02 and appropriate uncertainty and penalty factors for flux peaking augmentation and 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 3.2-4 can be maintained by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX and 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 the 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 l

ARKANSAS - UNIT 2 B 3/4 2-3 Amendment No. 24, 26, 32 t

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