ML20102B216
| ML20102B216 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/22/1992 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20102B212 | List: |
| References | |
| NUDOCS 9207280266 | |
| Download: ML20102B216 (9) | |
Text
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I PROPOSED TECllNICAL SPECIFICATION Cl(ANGES 8
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1 76 ELE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS l
FUNCTIONAL UNIT' TRIP SETPOINT ALLOWABLE VALUES
~
il.
' Manual Reactor Trip.
Not Applicable Not Applicable l
2.
iLinear Power-Level-- High
- a. ~Four Reactor. Coolant Pumps 5 110% of RATED THERMAL POWER-5 110.712% of PATED THERMAL FOhTR
.Cperating j
L.,-
.b.
Three Keactor Coolant Pumps I
' Operating
- c. 'Two Reactor Coolant Pumps OperetingL--Same Loop d.
Two Reactor Coolant Pumps' Operating - Opposite. Loops 3.
Logarithmic' Power Level -
.High (1)-
50.73% of RATED THERMAL POhT.R
$ 0.819% of RATED THERMAL POWER -
4.
Pressurizer: Pressure
.High 5 2362 psia-5 2370.887, psia 5.
- Pressurizer' Pressure - Low 2l1717.4 psia (2) 2 1686.3 psia (2) 6.-
Containment-Pressure'- High
$i18.4' psia 5 19.024 psia
-7.-
Steam Generator' Pressure
. Low 2 751 psia (3) 2 729.613 psia (3) 8.-
' Steam Generator-Level.i-! Low;
'2 23% (4) 2 20.111 (4)
,s -
- ' These values left blank pending NRC approva'l of safety analyses for operation with less than four reactor coolant pumps-operating.
ARKANSAS 1-UNIT'2-5 Amendment No. 7, ff, $5 a
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2.1 SAICTY L1 HITS AND L1_HITING SAFETY SYSTEM SITTINGS, BASES 1
I 2.1.1 REACT 0F CORE i
The restrictions of these sately limits prevent overheating of the fuel cladding and possible clat'. ding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel c4 adding is prevented by i,1) restricting fuel operation to within the nucleate boiling regime where the heat transfer coeff!.clent is large and-the cladding surface temperature is slightly above the ccolant sat 1 ration 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 v\\thAn the nucleate bofling regime of heat _
i transfer, the heat transfer coefficient is large enough no that the maximur.4 clad surface temperature is only slightly greater than the coolant
(
saturation temperature.
- he upper boundary of the. nucleate boiling regime is termed " departure-from nucleate boiling" (DNB).- At this point, there f.s a sharp reduction of the hee.t 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.
defined as the ratio of the predicted DNB heat flux
.t a particular core location to the actual heat flux at that location, is indicative of the l
margin to DNB.
The mininami 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 t
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)r fuel rod integrity would be maintained. only if + ' a design and operating conditions are appropriate throughout the
.i>o' the fuel rods.
Volume changes which accompany the solid to liquit d r.nge are ton involves'the significant and require accommodation. _Another cop..
redistribution of the fuel which depends on the_extu.
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 coaterline melting--is established as a Safety Limit..
To account for fuel rod dynamics (lags), the directly indicated linear-heat rate is dynamically adjusted.
A steady state' peak linear heat rate of 21'kw/ft-has been' established as the Safety Limit to p; event fuel centerline melting.during normal operation.
Following design basis anticipated' operational occurrencas, th transient linear heat' rate may exceed ' 21 kw/ft as long = as the fuel' ce.aerline melt temperature is not excee ed.
x ARKANSAS - UNIT l2 B 2-1 Amendment.No. 24 T66, 4
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1
. SAFETY LIMITS _AND LIMITING SAFETY __ SYSTEM SETTINGS BASES t
Limiting safety system settings foi the Low DNBR, liigh Local Power Density, liigh Logarithmic Power Level, Low Pressurizer Pressure and high Linear Power Level trips, and limiting conditions for operation on DNDR and kw/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits (i.e., DNBR and conterifne fuel melt temperature) are not exceeded during normal operation and design basis anticipated operational occurrences.
2 1.2 REACTOR COGLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant hystes, from overpreusurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The Reactor Coolant System components ato designed to Section 111 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 criterie and associated code requirements.
Tha entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to 1.11tial operation.
2mlut REACTOR TRIP SETPOINIS The Reactor Trip Setroints 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 enrare thct the reactor core and reactor coolant system are prevented from ' exceeding their. Safety Limits during normal operation and design basis anticipated operational' occurrences and te assist the Engineered Safety Features Actuation System'in mitigating the-
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consequences of accidents. -Operation with a trip sot'less. conservative than its Trip Setpoint hat within its specified Allowable Valoe is acceptable on the basis that the difference between each Trip Setpoint and
'the Allowablo Value is equal to or less than the!drif t allowance assumed for each trip in the enfety analyses.
The L"BR - Low and Local Power Density - High are digita11y' generated' trip setpoints based on Limiting Safety Systen Settings of 1.25 and 21.0 kw/ft, respectively.
Since;these trips are digitally generated byLthe Core Protection Calculators, the trip values are not subject to drif ts common to trips generated by analog type equipment. Tho' Allowable Values foi these trips are therefore the same as the Trip Setpoints.
ARKANSAS - UNIT 2 B 2-2
. Amendment No. 24, fp, 77,_
SAFETY Lly1TS AND L1H171NG SAFETY SYSTEM SETTINGS DAEES Pressurizer Pressure liigh The Pressurizer Pressure-liigh t rip, 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 actpoint is at 52370.887 psia which is below the nominal lift setting (2500 psia) of the pressurizar nafety valves and its operation avoids the undesirable operation of the pressurizer safety valves.
Pressurizer Pressure-Low-The Pressurizer Pressuro-Low trip is provided ' to trip the reactor and to assist the EngJneered Safety Features System in the event of a Loss of Coolant Accident. During normal oporation, this trip's setpoint is set at 21686.3 psia. _This trip's setpoint may be manually decreased, to a minimum value of 100 psia, as_ pressurizer pressure is reduced during plant shutdoens, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at 5200 psil this setpoint increases automatically as pressurizer pressure increases until the_ trip setpoint_is reached.
Containme.nt Pres sur e-Ilieh The Containment Pressuro-High trip providos assurance that a reactor trip is initiated concurrently with a safety injection. Tha setpoint for this trip is identical to the safety injection setpoint.
Steam Generator Presstrelaw 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-po!nt of approximately 900 psia no 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 '5200 psi; this setpoint increases automatically as steam generator pressure _ increases'until the trip setpoint is reached.
i ARKANSAS'- UNIT 2 B 2-4' AmendmentMo.37, 3-
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4 POWER DISTRIBUTION LIMITS 12EEEURIZER PRESSUrq LIMITINf),SQM),LI1Q1J,QR_OPERAIJQN 3.2.8 The average pressurizer pressure shall bc. maintained between 2025 psia and 2275 psia.
APPLICABJL1,T11 HDDE 1.
ACIl08:
With the avetogo pressurizer pressura exconding it. int.s, restore the pressure to within its limf.t-withf n 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to loss than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS n
4.2.8 The average pressurizor pressure shall be dotarmined to bo-within its limit at least once por 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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I ARKANSAS - UNIT 2 3/4 2-14 Amendment No. M U, H,
l 1
TABLE 3.3-4
=
ENGINr m n SAFETY FEATURE ACTUATION SYSTEM INSTRUMINTATION TRIP VALliES i
ALL0iiABLE FUNCTIONAL UNIT TRIP SETPOINT VAIm"E5 1.
- SAFETY' INJECTION (SIAS)
. i
. a.,. Manual:.'(Trip Buttons) hot Applicable Not Applicable l
. b.
Containment Pressure - High-5 18.4 psia 5 19.024 psia c.
Pressurizer Pressure -Low 21717.4 psia (*.)
216S6.3 psia (1) 2..
Not Applicable Not Applicable
. Manual (Trip: Buttons) a.:
- b.:
Containment Pressure -- High-High 5 23.3 psic
$ 23.490 psia
- 3..
CONTAINMENT ISCLATION-(CIAS)
- a..
- Manual-;(Trip Buttons).
Not-Applicable ~.
Not Applicable
. b.~
Containment Pressure - High 5 18.4 psia.
5 19.024 psia 1
W ARKANSAIf-: UNIT 2.
3/4 3-16
' Amendment No. If h
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o TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES AILCT4ABLE FUNCTIONAL UNIT TRIP VALUE VALUES i
4.
MAIN STREAM AND FEEDWATER ISOLATION (MSIS) a.
Manual-(Trip Buttons)
Not Applicable Not Applicable I
b.
Steam Generator Pre.ssure - Low 2 751 psia (2).
2 729.613 psia (2)
.5.
CONTAINMENT C00LINti (CCAS) l a.
Manual (Trip Buttons)
Not Applicable Not applicable b.
Containment Pressure'- High 5 18.4 psia
$ 19.024 psia c.-
' Pressurizer 'rressure - Low 2 1717.4 psia (1) 21686.3 psia (1) 6.
RECIRCULATION.(RAS) a.
Manual (Trip Buttons)'
Not Applicable Not Applicable i
b.
liefueling Water Tank - Low.
54,400 2,370 gallons between 51,050 and 58,600 t
(equivalent to 6.0 0.5*
gallons (equivalent to i
indicated level) between 5.111* and 6.3897.
indicated level)
.7 LOSS OF' POWER a.
4.16 kv Emergency Bus Undervoltage
-(Loss of Voltage) 3120 volts.(4) 3120 volts (4)
- b. _
460 volt Emergency Bus Undervoltage 423 2.0 volts 423 4.0 volts
..(Degraded Voltage)-
with an:8.0 0.5' with ar. 8.0 0.8 second time deley second time delay 3
1 ARKANSAS
. UNIT 2 3/4 3-17 Amendment No. ZjL
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POWER DISTRIBUTION LIMITS BhELD from the penalties associated with each batch, accountfog for the offsetting 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 minimuu value used in the BOCA safety analyses.
ALL.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification in ptovided to ensure that the actual value of reactor coolant cold Icg temperature is nalntained within the range of values used in'the safety analyses.
3/4.2.7 AX1AL SHAPE INDEX This specification is_provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within tho' range of values used in the safety analysec.
3/4.2.8 ERESSURIZER PRESSMBE This specification is provided to ensure that the actual value of pressurizer pressure is biaintained within the range of values used in the safety analyses.
Safety analyses cover a pressure. range from 2000 psia to 2300 psia. The upper and lower allowable 11mits:(2275.and 2025_ psia) ace adjusted by 25 psi to bound pressure instrumentation measurement uncertainty.
f-ARKANSAS -LUNIT 2
-B 3/4~2-4
-Amendme'nt No.;14 -$2, fA,
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