ML20151U397
| ML20151U397 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 04/14/1988 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20151U395 | List: |
| References | |
| NUDOCS 8805020047 | |
| Download: ML20151U397 (14) | |
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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - Tavo > 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be. qual to or greater than the limit line of Figure 3.1-lb.
APPLICABILITY:
MODES 1, 2**, 3, and 4.
ACTION:
With the SHUTDOWN MARGIN less than the limit line of Figure 3.1-lb immediately initiate and continue boration at 2 40 gpm of 2300 ppm boric acid solution or equivalent until the rcquired SHUTDOWN MARGIN is restored.
SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than the limit of Figure 3.1-lb:
a.
Within one hour after detection of an inoperable CEA(s) 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 while the CEA(s) is inoperable.
If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the witi Pawn worth of the immovable or untrippable CEA(s).
b.
When in h.':'i 1 or 2, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that l
CEA group withdrawal is within the Transient Insertion Limits of l
Specification 3.1.3.6.
I c.
When in MODE 2##, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality by verifying that the predicted critical CEA position is j
within the limits of Specification 3.1.3.6.
l d.
Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.
l Adherence to Technical Specification 3.1.3.6 as specified in Surveillance i
Requirements 4.1.1.1.1 assures that there is sufficient available shutdown margu, to match the shutdown margin requirements of the safety analyses.
See Special lest Exception 3.10.1.
With Keff 2 1.0 With Keff < l.0 CALVERT CLIFFS - UNIT 1 3/4 1-1 Amendment No.ff//7J//EE/
1
- Jpf, 8805020047 880414 ADOCK 0500 7
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O ACCEPTABLE y
OPERATION (EOC,5.0) 5.0 REGION u
E 4
24.0 0
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MINIMUM SHUTDOWN MARGIN n
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(BOC,3.5)
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R 1 3.0 z
7 UNACCEPTABLE 5
O OPERATION S
REGION 2.0 1.0 i
0.0 l
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TIME IN CYCLE i
Figure 3.1-Ib 4
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m
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:
a.
Less positive than the limit line of Figure 3.1-la, and b.
Less negative than -2.7 x 10-4 0
21 k/k/ F at RATED THERMAL POWER.
APPLICABILITY:
MODES 1 and 2*#
ACTION:
With the moderator temperature coefficient outside any one of the above limits, be in at least H0Y STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE RE0VIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements.
MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.
With Keff 2 1.0.
See Special Test Exception 3.10.2.
CALVERT CLIFFS - UNIT 1 3/4 1-5 Amendment No. /E//EE//Jpf,
UNACCEPTABLE OPERATION REGION POSITIVE MTC LIMIT LINE 0.70 o o (0.7,0.7)
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- N E 0.60-i S
O.50' U
5" 0.40 ACCEPTABLE S
OPERATION REGION s 0.30 (1.0,0.3)
E R 0.20 5
m g 0.10
$l 4
0.00 O.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 FRACTION OF RATED THERMAL POWER FIGURE 3.1-la Fraction of Rated Thermal Poweg /'F) vs.' Allowable Positive MTC Limit (10' op I
l CALVERT CLIFFS - U"IT 1 3/4 1 Sa
1.00 n -1.00,Gp 5 @ 35%
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i 0.90
(;4.90,Gp 5 0,35%
I 0.80 l
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0.75,Gp 5 9 50%
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0.70
.70,Gp 5 @ 60%
i q65,Gp5@85%
0.56,Gp 4 0 50%
I" 0.50 ILDngTar'm SF)rtTerm 0.40 gSteady State %-Steady State Insertion Insertion 0.30 l Limit Limit
\\o-0.20,Gp3060%
IGrp 5 0 25%
Grp 4 0 20%
w) 0.20
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0.00 o-0.00, Gp 3 @ 60%
Allowable BASSS h Grp 5 @ 55%
l Operating Region Groups:
5 3
1 I
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I f
I f
f I
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I f
I f
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l 0
20 40 60 80 100 0
20 40 60 80 100 0
20 40 60 80 100 136.0 108.8 81.6 54.4 27.2 0
136.0 108.8 81.6 54.4 27.2 0 136.0 108.8 81.6 54.4 27.2 0 4
2 l
I I
f I
f f
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I 0
20 40 60 80 100 0
20 40 60 80 100 136.0 108.8 81.6 54.4 27.2 0
136.0 108.8 81.6 54.4 27.2 0 i
% CEA INSERTION i
INCHES CEA WITHDRAWN t
Figure 3.1-2 CEA GROUP INSERTION LIMITS VS. FRACTION OF ALLOWABLE THERMAL POWER 1
FOR EXISTING RCP COMBINATION
l 1.10 i
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1.00
(-0.06,1.00)
- (0.12,1.00)
I UNACCPETABLE UNACCEPTABLE 0.90 OPERATION OPERATION REGION REGION 0.80 d[
0.70
(-0.3,.70)
- ACCEPTABLE
-r (0.3,.70)
OPERATION REGION 0.60 g
2 0.50
(-0.3,.50)
- 0.40 2
3 0
0.30 0.20
( 45,.15) w 0.10 0.05
-0.6
-0.4
-0.2 0.0 0.2 0.4 0.6 PERIPHERAL AXIAL SHAPE INDEX, Y 7
Figure 3.2-2 Linear Heat Rate Axial Flux Offset Control Limits CALVERT CLIFFS 3/4 2 4 1
-1.10 i
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1.00
(-0.1,1.00) *
- (0.15,1.00)
UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION 0.90 REGION REGION 0.80
(-0,3,.80) '
ACCEPTABLE
- (0.3,.80)
OPERATION REGION 0.70 se
@E 8 0.60 ho$ 0.50
(-0.3,.50)
- hb
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z g 0.40 O
b id 55 0.30 0.20
(.45,.15)
,r 0.10 0.05
-0.6
-0.4
-0.2 0.0 0.2 0.4 0.6 PERIPHERAL AXIAL SHAPE INDEX, Y 7
Figure 3.2-4 DNB Axial Flux Offset Control Limits CALVERT CLIFFS 3/4 2 11
n 2.1 SAFETY LIMITS BASES 2.1.1 REACT 0F CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which could result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 22.0 kw/ft.
Centerline fuel melting will not occur for this peak linear heat rate. Overheating of the fuel cladding is prevented by 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.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and, therefore, THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the CE-1 correlation.
The CE-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat ? lux ratio, DNBR, defined as the ratio of the heat flux that would cause FNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to the DNB SAFDL of 1.15 in conjunction with the Extended Statistical Combination of Uncertainties (ESCU).
This DNB SAFDL assures with at least a 95 percent probability at a 95 percent confidence level that DNB will not occur.
The curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show conservative loci for points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of various pump combinations for which the DNB SAFDL is not violated for the family of axial shapes and corresponding radial peaks shown in Figure 82.1-1.
The limits in Figures 2.1-1, 2.1-2, 2.1-3, and 2.1-4 were calculated for reactor coolant inlet temperatures less than or equal to 580'F.
The dashed line at 580'F coolant inlet temperature is not a safety limit; however, operation above 580*F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature.
Reactor operation at THERMAL POWER levels higher than 110%
of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in CALVERT CLIFFS - UNIT 1 B 2-1 Amendment No. EJ///E//7J, 95//19/19/95,
SAFETY LIMITS BASES Table 2.1.-l.
The area of safe operation is below and to the left of these lines.
The conditions for the Thermal Margin Safety Limit curves in Figures 2.1-1, 2.1-2, 2.1-3, and 2.1-4 to be valid are shown on the figures.
The reactor protective system in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a DNBR of less than 1.15, in conjunction with the ESCU methodology, and preclude the existence of flow instabilities.
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 pressure vessel and pressurizer are designed to Section III, 1967 Edition, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure.
The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, Class I, 1969 Edition, which permits a maximum transient pressure of 110% (2750 psia) of coroponent 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 3215 psia to demonstrate integrity prior to initial operation.
CALVERT CLIFFS - UNIT 1 B 2-3 Amendment No. JJ//J9//;f 8, 11//19/J9/99,
LIMITING SAFETY SYSTEM SETTINGS BASES operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service.
The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errcrs and response times of equipment involved to maintain the DNBR above the DNB SAFDL of 1.15, in conjunction with the ESCU methodology, under normal operation and expected transients.
For reactor operation with only two three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set; oints, the Power Level-High trip setpoints, and the Thermal Margin / Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two-or three-pump position.
Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below DNB SAFDL of 1.15, in conjunction with the ESCV methodology, during normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating.
Pressurizer Pressure-Hiah The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip.
This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.
Containment Pressure-Hiah The Containment Pressure-High trip provides assurance that a reactor trip is initiated prior to, or at least concurrently with, a safety injection.
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.
1he setting of 685 psia is sufficiently below the full-load operatir.g point of 850 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 setting was used with an uncertainty factor of 85 psi which was based on the main steam line break event inside containment.
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l CALVERT CLIFFS - UNIT 1 8 2-5 Amendment No. JJi/,4El//J, l
B5ll1171/19/19/B5,
LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water level The Steam Generator Water Level-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the pressure of the reactor coolant system will not exceed its Safety Limit.
The specified setpoint in combination with the auxiliary feedwater actuation system ensures that sufficient water inventory exists in both steam generators to remove decay heat following a loss of main feedwater flow event.
Axial Flux Offset The axial flux offset trip is provided to ensure that excessive axial peaking will not cause fuel damage.
The axial flux offset is determined from the axially split excore detectors.
The trip setpoints ensure that neither a DNBR of less than the DNB SAFDL of 1.15, in conjunction with ESCU methodology nor a peak linear heat rate which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power maldistributions. These trip setpoints were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the uncertainty associated with the excore to incore axial flux offset relationship.
Thermal Marain/ Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than the DNB SAFDL of 1.15, in conjunction with ESCU methodology.
The trip is initiated whenever the reactor coolant system pressure sigt.a1 drops below either 1875 psia or a computed value as described below, whichever is higher.
The computed value is a function of the higher of a T power or neutron power, reactor inlet temperature, and the nunber of reactor coolant pumps operating. The minimum value cf reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the mutmum CEA deviation permitted for continuous operation are assumed in the generation of this trip function.
In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.
Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.
CALVERT CLIFFS - UNIT 1 8 2-6 Amendment No. Ul/# 1//E, 711/Apl/19/M/M,
3/4.1 REACTIVITY CONTROL SYSTENS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within accept-able limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
The most limiting SHUTDOWN MARGIN requirement at beginning of cycle is determined by the requirements of several transients, including Boron Dilution and Steam Line Rupture.
The SHUTDOWN MARGIN requirements for these transients are relatively small and nearly the same.
However, the most limiting SHUTDOWN MARGIN requirement at end of cycle comes from just one transient, the Steam Line Rupture event.
The requirement for this transient at end of cycle is significantly larger than that for any other event at that time in cycle and, also, considerably larger than the most limiting requirement at beginning of cycle.
The variation in the most limiting requirement with time in cycle has been incorporated into Technical Specification 3.1.1.1, in the form of a specified SHUTDOWN MARGIN value which varies liner.rly from beginning to t cycle. This variation in specified SHUTDOWN MARGIN is conservative relati..
to the actual variation in the most limiting requirement.
Consequently, adherence to Technical Specification 3.1.1.1 provides assurance that the available SHUTDOWN MARGIN at anytime in cycle will exceed the most limiting SHUTCOWN MARGIN requirement at that time in cycle.
In MODE S, the reactivity transients resulting from any event are minimal and do not vary significantly during the cycle. Therefore, the specified SHUTDOWN MARGIN in MODE 5 via Technical Specification 3.1.1.2 has been set equal to a constant value which is determined by the requirement of the most limiting event at any time during the cycle, i.e., Boron Dilution with the pressurizer level less than 90 inches and the sources of non-borated water restricted.
Consequently, adherence to Technical Specification 3.1.1.2 provides assurance that the available SHUTDOWN MARGIN will exceed the most limiting SHUTDOWN MARGIN requirement at any time in cycle.
l CALVERT CLIFFS - UNIT 1 B 3/4 1-1 Amendment No. J///fE//77,
$5]/Jpf,
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9,601 cubic feet in approximately 24 minutes.
The reactivity change rate associated with boron concentration reductions will therefore be within the capability of operator recognition and control.
3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)
The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle.
The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
The confirmation that the measured MTC value is within its limit provides assurances that the coefficient will be maintained within acceptable values throughout each fuel cycle.
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CALVERT CLIFFS - UNIT 1 8 3/4 1-la Amendment No. 72///E//77, pf//Jpf,
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MODIFICATIONS REQUESTED ON PAGES 9-23 9-24 9-25 0F REFERENCE (A), ARE NO IDNGER APPLICABLE WITH THE PARTIAL WITHDRAWAL OF TECHNICAL SPECIFICATION 3.1.1.1 REQUEST FOR LICENSE AMENDMENT i
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