ML17215A293
| ML17215A293 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  (DPR-67-A-063, DPR-67-A-63) |
| Issue date: | 03/01/1984 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17215A294 | List: |
| References | |
| NUDOCS 8403120217 | |
| Download: ML17215A293 (25) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON D C 20555 FLORIDA POWER 5 LIGHT COMPANY DOCKET NO. 50-335 ST.
LUCIE PLANT UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
63 License No.
'I 1.
The Nuclear Regulatory Commission (the Cormission) has found that:
A.
B.
C.
D.
E.
The application for amendment by Florida Power 8 Light Company, (the licensee) dated January 20, 1983 as supplemented, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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 The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8403120217 840301 PDR IIIIDDCK 05000335 P
I I
I
2.
Accordingly, Facility Operating License No.
DPR-67 is amended by changes to the Technical Specifications as indicated in the Attach-ment to this license amendment by amending paragraph 2.C(2),
and by adding a
new paragraph 2.C(4) to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
6>
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
(4)
Prior to reaching 38,000 MWd/MTU peak assembly, the
'licensee. must use. an approved, method to show that Combustion Engineering fuel will not experience creep collapse unless the new Exxon Corporation methodology has been approved for use by the staff and its results are valid for Cycle 6.
3.
The license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Operating Reactors Branch ¹3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
March I, 1984
ATTACHMENT TO LICENSE AMENDMENT NO.
63 TO FACILITY'OPERATING L'ICENSE'NO DPR-"67"""'"'"'""""""'""'""
DOCKET NO. 50-335 Replace the following pages of the Appendix "A" Technical Sepcifications with the enclosed pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
~Pa es B '2-1 B 2-3 B 2-5 B 2-7 3/4 1-1 3/4 1-5 3/4 2-2 3/4 2-5 B 3/4 1-1 B 3/4 2-1 B 3/4 4-1
2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit 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 is prevented by maintaining the steady state peak linear heat rate below the level at which centerline fuel melting will occur.
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 Exxon XNB correlation.
The XNB 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 flux
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 1.22 using the XNB DNBR correlation.
This value corresponds to a
95 percent probability
.at a
95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions'he curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the minimum DNBR is no less than the DNBR limit for the family of axial shapes and corresponding, radial peaks shown in Figure B 2.1-1.
The limits in Figure 2.1-1 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 PO'WER levels higher than 112%
of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in Table 2.1-1.
The area of safe operation is below and to the left of these lines.
ST.
LUCIE - UNIT 1
B 2-1 Amendment No. g7, gg,
~ 2.0 1.6 z0 I
1.2 I-0 K
0.6 X
0.4 ROD RADIALPEAK 1.44 1.44 1.51 1.56 1.56 0
0 10 20 30 40 50 60 70 80 90 100 PERCENT OF ACTIVE CORE LENGTH FROM BOTTOM Figure B2.1-1 Axial Power Distribution for Thermal Margin Safety Limits
SAFETY LIMITS BASES The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.
The reactor protective system in combination with the Limiting Conditions for Operationis 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 the DNBR limit 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 of the ASME Code for Nuclear Power Plant components which permits a maximum transient pressure of 110K (2750 psia) of design pressure.
The Reactor Coolant System piping, valves and fittings are designed to ANSI B 31.7, Class I which permits a maximum transient pressure of 110K (2750 psia) of component 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.
ST.
LUCIE - UNIT 1
B 2-3 Amendment No. gg, 63
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow-Low (Continued) 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 errors and,response times of equipment involved to maintain the DNBR above the DNBR limit under normal I
operation and expected transients.
For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip setpoints, 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 the DNBR limit during normal I
operational transients and anticipated transients when only two or three reactor coolant pumps are operating.
Pressurizer Pressure-Hi h
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 con-current operation with the power-operated" relief valves avoids the undesir-able operation of the pressurizer code safety valves.
Containment Pressure-Hi h
The Containment Pressure High tr ip pyovides assurance that a reactor trip is initiated concurrently wi th 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 sub-sequent cooldown of the reactor coolant.
The setting of 600 psia is sufficiently below the full-load operating point of 800 psig so as not ST.
LUCIE - LiNIT 1
B 2-5 Amendment No. 3g, P5, gg, 63
LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Pressure-Low (Continued) 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 + 22 psi in the accident analyses.
Steam Generator Water Level - Low 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 design pressure of the reactor coolant system will not be exceeded due to loss of steam generator heat sink.
The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to provide a margin of more than 10 minutes before auxiliary feedwater is required.
Local Power Densit -Hi h The local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring,'s provided to ensure that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a consequence of axial power maldistributions.
A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2.
The AXIAL SHAPE INDEX is calculated from the upper and lower ex-core neutron detector channels.
The calculated setpoints are generated as a function of THERMAL POWER level with the allowed CEA group position being inferred from the THERMAL POWER level.
The trip is automatically bypassed below 15 percent power.
The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment per-nitted for continuous operation are assumed in generation of the set-points.
In addition, CEA group sequencing in accordance with the Specifications
- 3. l. 3. 5 and 3. l. 3. 6 is assumed.
Finally, the maximum insertion of CEA banks which can occur during any anticipated operational ccurrence prior to a Power Level-High trip is assumed.
ST.
LUCIE - UNIT 1
B 2-6 Amendment No.
27
LIMITING SAFETY SYSTEM SETTINGS BASES Thermal Mar in Low Pressure The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than 'the DNBR limit.
I The trip is initiated whenever the reactor coolant system pressure signal drops below either 1887 psia or a computed value as described below, whichever is higher.
The computed value is a function of the higher of aT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX.
The minimum value of reactor coolant, flow rate, the maximum AZIMUTHAL POWER TILT and the maximum 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.
The Thermal Margin/Low Pressure trip setpoints include appropriate allowances for equipment response time, calculational and measurement uncer-
- tainties, and processing error.
A further allowance of 30 psia is included to compensate for the time delay associated with providing effective termina-tion of the occurrence that exhibits the most rapid decrease in margin to the DNBR limit.
As metric Steam Generator Transient Protective Tri Function ASGTPTF The ASGTPTF consists of Steam Generator pressure inputs to the TM/LP calculator, which causes a reactor trip when the difference in pressure between the two stdam generators exceeds the trip setpoint.
The ASGTPTF is designed to provide a reactor trip for those events associated with secondary system mal-functions which result in asymmetric primary loop coolant temperatures.
The most limiting event is the loss of load to one steam generator caused by a single main steam isolation valve closure.
The equipment trip setpoint and allowable values are calculated to account for instrument uncer tainties, and will ensure a trip at or before reaching the analysis setpoint.
ST.
LUGIE - UNIT 1
B 2-7 Amendment No. g7, gg, 44/~
LIMITING SAFETY SYSTEM SETTINGS BASES Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15'A of RATED THERMAL POWER.
This trip provides tur bine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these valves.
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 Protec-tion System.
Rate of Chan e of Power -Hi h
The Rate of Change of Power-High trip is provided to protect the core during startup operations and its use serves as a backup to the administra-tively enforced startup rate limit.
Its trip 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.tri p setting is required to enhance the overall reliability of the Reactor Protection System.
U ST.
LUCI E - UNI 1
B 2-8 Ame.ndment No. 4 8
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 BORATION CONTROL SHUTDOWN MARGIN - T
> 200'F av LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be
> 3.6g ak/k.
APPLICABILITY:
MODES 1, 2*, 3 and 4.
ACTION:
With the SHUTDOWN MARGIN < 3.6g, ak/k, immediately initiate and continue boration at
> 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.
SURVEILLANCE RE UIREMENTS 4.1.1.1 a.
C.
1 'he SHUTDOWN MARGIN shall be determined to be
> 3.6g bk/k:
Within one hour after detection of an inoperable CEA(s) and at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s-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 withdrawn worth of the immovable or un-trippable CEA(s).
When in MODES 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 CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3.1.3.6.
When in MODE 2
, at least once during CEA withdrawal and at least once per hour thereafter until the reactor is critical.
Prior to initial operation above SX RATED THERMAL POWER after each fuel loading, by consideration of the factors of e
- below, with the CEA groups at the Power Dependent Inserti on Limits of Specification
- 3. 1.3.6.
See Special Test Exception 3..10. 1.
With K
> 1.0.
With K
( 1.0.
eff ST.
LUCIE - UNIT 1
3/4 1-1 Amendment No. g7, gP, 63
REACTIVITY CONTROL SYSTEMS
'SURVEILLANCE RE UIREMENTS Continued e.
When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by con.-
sideration of the following factors:
l.
2.
3.
4.
5.
6.
Reactor coolant system boron concentration, CEA position,*
Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0X ak/k at least once per 31 Effective Full Power Days (EFPD).
This comparison shall consider at least those factors'stated in Specification 4.l.l.l.l.e, above.
The predicted reactivity values shall be adj'usted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.
- For Modes 3 and 4, duri ng calculation of shutdown margin with all CEA's verified fully inserted, the single CEA with the highest reactivi ty worth need not be assumed to be stuck in the fully withdrawn position.
ST.
LUCIE - UNIT 1
3/4 1-2 Amendment No.
45
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 a
~
b.
C.
The moderator temperature coefficient (MTC) shall be:
Less positive than 0./ x 10 ak/k/'F whenever THERMAL POWER is
< 70Ã of RATED THERMAL POWER, Less positive than 0.2 x 10 hk/k/'F whenever THERMAL POWER is
> 70K of RATED THERMAL POWER, and Less negative than -2.8 x 10 hk/k/'F at RATED THERMAL POWER.
APPLICABILITY:
MODES 1
and 2*8 ACTION:
With the moderator temperature coefficient outside any one of the above limits, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE RE UIREMENTS 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.
- W)th K ff > 1.0.
PSee Special Test Exception 3.10.2.
ST.
LUCIE - UNIT 1
3/4 1-5 Amendment No. g7, 63
REACTIVITY CONTROL-SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each'uel cycle:
a.
Prior to initial operation above 5X of RATED THERMAL POWER, after each refueling.
b.
At any THERMAL POWER, within 7 EFPD after initially reaching a
RATED THERMAL POWER equilibrium boron concentration.
c.
At any THERMAL POWER, within 7 EFPD after reaching a
RATED THERMAL POWER equilibrium boron concentration of 300 ppm.
ST.
LUCIE - UNIT 1
3/4 1-6 Amendment No. 27
3/4.2 POWER -DISTRIBUTION LIMITS
'LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed the limits shown on Figure 3.2-1.
APPLICABILITY:
MODE 1'.
ACTION:
With the linear heat rate exceeding its limits, as indicated by four or more coincident incore channels or by the AXIAL SHAPE INDEX outside of the power dependent control limits of Figure 3.2-2, 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 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 RE UIREMENTS
- 4. 2. 1. 1 The provisions of'pecification 4. 0. 4 are not, applicable.
4.2.1.2 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring system.
4.2.1.3 Excore Detector Monitorin S stem - The excore detector moni-toring system may be used for monitoring the core power distribution by:
a.
Verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the full length CEAs are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Limit of Specification 3.1.3.6
~
b.
Verifying at least once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limits shown on Figure 3.2-2.
ST.
LUG IE - UNIT 1
3/4 2-1 Amendment No. g7, 32
POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued-C.
Verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2, where 100 p'ercent of maximum allowable power represents the maximum THERMAL POWER allowed by the following expression:
where:
1.
M is the. maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination.
'2.
N is the maximum allowable fraction of RATED THERMAL POWER as determined by the F
curve of Figure'.2-3.
T xy 4.2.1.4 Incore Detector Monitorin S stem
- The incore detector monitor-
¹ ing system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:
- a. 're adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 days of accumulated operation in MODE l.
b.
Have their alarm setpoint adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alarms:
1.
A measurement-calculational uncertainty factor of 1.07,*
2.
An engineering uncertainty factor of 1.03, 3.
A linear heat rate uncertainty factor of 1.01 due to axial fuel densification and thermal expansion, and 4.
A THERMAL POWER measurement uncertainty factor of 1.02.
¹If the core system becomes inoperable, reduce power to M x N within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and monitor linear heat rate in accordance with Specification 4.2.1.
- An uncertainty factor of 1.10 applies when in LOAD FOLLOW OPERATION.
ST.
LUCIE - UNIT 1
3/4 2-2 Amendment No. 77, g7, gg, 63
THIS PAGE INTENTIONALLY LEFT BLANK ST.
LUCI E - UNIT 1
3/4 2-5 Amendment No. g7, gg, gg, 63
POWER DISTRIBUTION LIMITS TOTAL PLANAR RADIAL PEAKING FACTOR F LIMITING CONDITION FOR OPERATION 3.2.2 The calculated value of F, defined as F
=
F (1+T ), shall be T
T xy'y xy q
'imited to 1.70.
APPL ICABILITY:
MODE 1*.
ACTION:
With F 1.70, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:
T xy a
~
b.
Reduce THERMAL POWER to bring the combination of THERMAL POWER and F to within the limits of Figure 3.2-3 and withdraw the full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification
- 3. 1.3.6; or Be in HOT STANDBY.
SURVEILLANCE RE UIREMENTS 4'.2.
1 The provisions of Specification
- 4. 0.4 are not applicable.
4.2.2 '
F shall be calculated by the expression F
=
F (1+T
) when T
T xy yy xy q
in non-LOAD FOLLOW OPERATION and by the expression F
= 1.03 F
(1+T
)
T xy xy q
when in LOAD FOLLOW OPERATION.
F shall be determined to be within xy its limit at the following intervals:
a.
Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading, b.
At least once per 31 days of accumulated operation in MODE 1, and c.
Within four hours if the AZIMUTHAL POWER TILT (T
) is
> 0.03.
q
- See Special Test Exception 3.10.2.
ST.
LUCIE - UNIT 1
3/4 2-6 Amendment No. g7,
)$,4 8
3 4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4. 1. l. 1 and 3/4. l. 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 acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent'riticality in the shutdown condition.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg The most restrictive condition occurs at EOL, with Tavo at no load operating temperature, and'is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.
In the analysis of this accident, a minimum SHUTDOWN MARGIN of 3.6%'k/k is required to control the reactivity transient.
Accordingly, the SHUTDOWN MARGIN required by Specification
- 3. 1. 1. 1 is based upon this limiting condition and is con-sistent with FSAR accident analysis assumptions.
For earlier periods during the fuel cycle, this value is conservative.
With Tavq ( 200'F, the reactivity transient resulting from a boron dilution event with a partially drained Reactor Coolant System requires a
2% ak/k SHUTDOWN MARGIN and restrictions on char ging pump operation to provide adequate protection.
A 2% bk/k SHUTDOWN MARGIN is 1.0% ak/k conservative for Mode 5 operation with total RCS volume present, however LCO 3.1:1.2 is written conservatively for simplicity.
3/4.1.1.3 BORON DILUTION AND ADDITION 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 changes in the Reactor Coolant System.
A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 11,400 cubic feet in approximately 26 minutes.
The reactivity change rate associated with boron concentration changes will be within the capability for operator recognition and control.
3 4.1.1. 4 MODERATOR TEMPERATURE COEFFICIENT MTC Amendment No.
g7, PPg//
B 3/4 1-1 The limiting values assumed for the MTC used in the accident and transient analyses were
+ 0.7 x 10 " bk/k/'F for THERMAL POWER levels I
( 70% of RATED THERMAL POWER,
+ 0.2 x 10 4 ak/k/'F for THERMAL POWER Tevels
> 70% of RATED THERMAL and - 2.S x 10
'> ak/k/'F at RATED THERMAL POWER.
Therefore, these limiting values are included in this specification.
Determination of MTC at the specified conditions ensures that the maximum
'ositive and/or negative values of the MTC will not exceed the limiting values.
ST.
LUCIE - UNIT 1
REACTIVITY CONTROL'YSTEMS BASES 3/4.1.1. 5 MINIMUM TEMPERATURE FOR CRITICALITY The MTC is expected to be slightly negative at operating conditions.
However, at the beginning of the fuel cycle, the MTC may be slightly positive at operating conditions and since it will become more positive at lower temperatures, this specification is provided to restrict reactor operation when T is significantly below the normal operating temperature.
avg 3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation.
The components required to perform this function include 1) borated water sources, 2) charging
- pumps,
- 3) separate flow paths,
- 4) boric acid pumps,
- 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.
With the RCS average temperature above 200'F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable.
Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injecti'on system failures during the repair period.
The boration capability of either system is sufficient to provide a
SHUTOOWN MARGIN from all operating conditions of 2.0% ak/k after xenon decay and cooldown to 200'F.
The maximum boration capability requirement occurs at EOL from full power equi librium xenon conditions and requires 7,925 gallons of 8.0% boric acid solution from the boric acid tanks or 13,700 gallons of 1720 ppm borated water from the refueling water tank.
The requirements for a minimum contained volume of 401,800 gallons of borated water in the refueling water tank ensures the capability for borating the RCS to the desired level.
The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4.
Therefore,,the larger volume of borated water is specified here too.
With the RCS temperature below 200'F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restric-tions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.
ST.
LUCIE - UNIT 1
B 3/4 1-2 Amendment No.
g7, g6, <;;
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 Excore Detector Monitoring System and the Incore Detector Monitoring System, provides adequate monitoring of the core power distribution and is-capable of verifying that the linear heat rate does not exceed its limits.
The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2.
In conjunction with the use of the excore monitori ng system and in establishing the AXIAL SHAPE INDEX limits, the following assump-tions are made:
1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied,
- 2) the AZIMUTHAL POWER TILT restrictions of Specifica-tion 3.2.4 are satisfied, and 3) the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.2.
The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable 1 imits of Figure 3.2-1.
The setpoints for these alarms include allowances, set in the conservative directions, for 1) a measurement-calculational uncertainty factor of 1.07,*
2) an engineering uncertainty factor of 1.03, 3) an allowance of 1.01 for axial fuel densifica-ti'on and therm'al expans'io'n, and 4) a THERMAL POWER" measurement uncertainty factor of 1.02.
3 4.2.2 3 4.2.3 and 3 4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING FACTORS -
F AND F AND AZIMUTHAL POWER TILT - T T
The limitations on F
and T
are provided to ensure that the assump-xy q
tions used in the analysis for establishing the Linear Heat Rate and Local Power Density-High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits.
The limitations on F
and T
are provided to ensure that the assumptions q
- An uncertainty factor of 1.10 applies when in LOAD FOLLOW OPERATION.
ST.
LUCI E - UNIT 1
B 3/4 2-1 Amendment No. g7, gg,
POWER DISTRIBUTION LIMITS BASES used in the analysis establishing the DNB Margin LCO, and Thermal Margin/Low Pressure LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. If F F
or T exceed T
T xy' their basic limitations, operation may continue under the additional restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat Rate, Thermal Margin/Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid.
An AZIMUTHAL POWER TILT ) 0.10 is not expected and if it should occur, sub-sequent operati'on would be restricted to only those operations required to identify the cause of this unexpected tilt.
The value of Tq that must be used in the equation F
=
F (1
+ T
)
T xy xy q
and F
=
F (1+T
) is the measured tilt.
l r
q The surveillance requirements for verifying that F
F and T
are T
T within their limits provide assurance that the actual values of Fx, F
T T
and T
do not exceed the assumed values.
Verifying F and F
after T
T q
xy r
each fuel loading prior to exceeding 75K of RATED THERMAL POWER provides additional assurance that the core was properly loaded.
3/4. 2, 5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of ) 1.22 throughout each analyzed transient.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a
12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
ST.
LUCIE - UNIT 1
B 3/4 2-2 Amendment No. p/,.gg 6"
3 4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above the DNBR limit during all normal operations and anticipated transients.
In MODES 1
and 2 with one reactor coolant loop not in operation, this specifica-tion requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In MODE 3, a single reactor coolant loopprovides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.
In MODE 4, and in NODE 5 with reactor coolant loops filled, a single reac-tor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either shutdown cooling or RCS) be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling loops be OPERABLE.
The operation of one Reactor Coolant Pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity, change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump in MODE 5 with one or more RCS cold legs less than or equal to 165'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary
- system, which could exceed the limits of Appendix G to 10 CFR 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either
- 1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or 2) by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 45 F above each of the Reactor Coolant System cold leg temperatures.
3/4.4 '
and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.
Each safety valve is designed to relieve 2 x 10s lbs per hour of saturated steam at the valve setpoint.
The relief capacity of a single safety val've is adequate to relieve any over-pressur e condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS over pressuriza-tion.
ST.
LUCIE - UNIT 1
B 3/4 4-1 Amendment No. gg, 5'g
~ 63
REACTOR COOLANT SYSTEM BASES 4
3/4.4.2 and 3/4.4.3 SAFETY VALVES (Continued)
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.
The combined relief capacity of these valves is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a
. complete loss of turb'ine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pr essure-High) is reached (i.e.,
no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.
Demonstration of the safety valves'ift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure
- Code, 1974 Edition.
3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydrauli-cally solid system and is capable of accommodating pressure sur ges during opera-tion.
The steam bubble also protects the pressurizer code safety valves and power operated relief valve against water relief.
The power operated relief valve and steam bubble function to relieve RCS pressure during all design transients.
Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer-Pressure-High signal minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
The required pressurizer heater capacity is capable of maintaining natural circulation sub-cooling*.
Oper'ability of the" heaters, which are powered. by a diesel generator
- bus, ensures ability to maintain pressure control even with loss of offsite power.
3 4.4.5 STEAM GENERATORS One OPERABLE steam generator provides sufficient heat removal capability to remove decay heat after a reactor shutdown.
The requirement for two steam generators capable of removing decay heat, combined with the requirements of Specifications 3.7.1.1, 3.7.1.2 and 3.7.1.3 ensures adequate decay heat removal capabilities for RCS temperatures greater than 325'F if one steam generator becomes inoperable due to single failure considerations.
Below 325'F, decay heat is removed by the shutdown cooling system.
The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision l.
Inservice inspec-tion of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that ther e is evidence of mechanical demage or progressive degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of character izing the nature and cause of any tube degradation so that corrective measures can be taken.
ST.
LUCIE - UNIT 1
8-3/4 4-2 Amendment No. gg, g7, 5 6