ML17213B033
| ML17213B033 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/08/1983 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17213B032 | List: |
| References | |
| NUDOCS 8302150008 | |
| Download: ML17213B033 (17) | |
Text
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POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS continued c.Verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2, where 100 percent of maximum allowable power represents the maximum THERMAL POWER allowed by the following expression:
MxN 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 Fx curve of Figure 3.2-3..4.2.1.4 Incore Detector Monitoring Systems-The incore detector monitor system may be used for monitoring the core power distribution by verifyi ng that the i ncore detector Local Power Density alarms: a.Are 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 1.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.*An uncertainty factor of 1.10 applies when in LOAD FOLLOW OPERATION.
g If the core system becomes inoperable, reduce power to M x N within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sand monitor linear heat rate in accordance with Specification 4.2.1.ST.LUCIE-UNIT 1 3/4 2-2 Amendment No.
/~~~'3/4.2:~PONE R 0 IS TR I B UT I ON L IN I T S BASES 3/4.2.1 LINEAR HEAT RATE The limitation of 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 thonitoring System, provide adequate monitoring of the core power distribution and are 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 conjuction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made: 1)the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2)the AZIMUTHAL POllER TILT restrictions of Speci fication 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 tlonitoring System continuously provides a direct measure of the peaking factors and the alarms which nave been established for the individual incore detector segments ensure that the peak linear heat rates will b.maintained within the allowable limits of Figure 3.2-1.The setpoints for these a'farms include allowances, set in the conservative directions, for 1)a measurement-cal culational uncertainty factor of 1.07*, 2)an engineering uncei t.ainty factor of 1.03, 3)an allowance of 1.01 for axial fuel densification and thermal expansion, 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 y AND Fr AND AZIMUTHAL POWER TILT Tq The limitation on FT and T are provided to ensure that tne assumptions used in the analysis for Btablisfling 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 FTr and Tq are provided to ensure that the assumptions
~An uncertainty factor of 1.10 applies when in LOAD FOLLOW OPERATION.
ST.LUCIE-UNIT 1 8 3/4 2-1 O.
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 withi'n 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 frcm nucleate boiling (DNB)and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measur able parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperatur e 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 ration, DNBR, defined as the ratio of the heat flux that would cause DNB 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 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 margi n to DNB for all operating conditions.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperatur e'ith four Reactor Coolant Pumps operating for which the minimum DNBR is no less than the DNBR limit for the family of axial shapes and cor responding 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 of 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 tempratur e.Reactor operation at THERMAL POWER levels higher than 112$of RATED THERMAL POWER is prohibited by the high power level trip setpoi nt specified in Table 2.1-1.The area of safe operation is below and to the left of these li nes.ST.LUCIE UNIT 1 B 2-1 SAFETY LIMITS BASES The conditions for the Thermal Margi'n Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.The reactor protective system in ccmbination 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 the DNBR limit and preclude the)exi stence of fl ow instabilities.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System for 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 11(g (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 lit@(2750 psia)of ccmponent design pressure.The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requiranents.
The entire Reactor Cool.ant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
ST.LUCIE-UNIT 1 B 2-3 Amendment No.
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 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 frcm going below the DNBR limit during normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating.
Press uri zer Pressure-High 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 nani nal 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.Contairment Pressure-High The Contairment Pressure-High trip provides assurance that a reactor trip is initiated 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.The setting of 600 psia is sufficiently below the full-load operating point of 800 psig so as not ST.LUCIE-,UNIT 1 B 2-.5 Anendment No.
LIMITING SAFETY SYSTEM SETTINGS BASES Thermal Mar in/Low Pressure The Thermal Margi n/Low Pressure trip is provided to prevent operation when the DNBR is less than the DNBR limit.The trip is initiated whenever the reactor coolant system pressure signal drops below either 1887 psia or a ccmputed value as described below, whichever is'igher.The ccmputed value is a function of the higher of hT 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 AXIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function.'n 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 appr opriate allowances for equipment response time, calculational and measurement uncertainties, and processing error.A further allowance of 30 psia is included to compensate for the time delay associated with providing effective termination of the occur rence that exhib1ts the most rapid decrease in margin to the DNBR limit.Asymmetric Steam Generator Tr ansient Protective Trip 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 steam generators exceeds the trip setpoint.The ASGTPTF is designed to provide a reactor trip for those events associated with secondary systan malfunctions 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 uncertainties, and will ensure a trip at or before reaching the analysis setpoint.ST.LUCIE-UNIT 1 B 2-7
POWER D ISTRIBUT ION 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 al 1 owable CEA group insertion Limits.If F, F or T exceed xy'their basic limitations, operation may continue under the additional r estrictions 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 Pressur e and Local Power Density-High LCOs and LSSS setpoints remain valid.An AZIMUTHAL PO'WER TILT)0.10 is not expected and if it should occur, subsequent oper ation would be restricted to only those operations required to identify the cause of this unexpected tilt.The value of T that must be used in the equation F=F (1+T)q and F=F (1,+T)is the measured tilt.r r'The surveillance requirements for verifying that F, F and T T T are within their limits provide assurance that the actual values of F, F xy'.r and T do not exceed the assumed values.Yeri fying F and F after q xy r each fuel loading prior to exceeding 75$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 minimun 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 sur veillance 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 fl ow'egradation and ensure correl ation of the fl ow indi cation channel s wi th 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 Z-Z Amendment No.
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~~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 NODES 1 and 2 with one reactor coolant loop not in operation, this specification 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 loop provides sufficient heat removal capability for removing decay heat;however, single failure considerations requi re that two loops be OPERABLE.In NODE 4, and in MODE 5 wi th reactor coolant loops filled, a single reactor 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 cooli ng or RCS)be OPERABLE.In NODE 5 with reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat;but si ngle failure considerations and the unavailability of the steam generators as a neat removing component, require that at least two shutdown cooling loops be OPERABLE.r 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 1n 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 pr ovided 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.
3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS frcm being pressurized abgve its Safety Limit of 2750 psia.Each safety valve is designed to relieve 2 x 10~lbs.per hour of saturated steam at the valve setpoint.The relief capacity of a single safety valve is adequate to relieve any overpressure 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 overpressurization.
ST.LUCIE-UNIT 1 B 3/4 4-1 Amendment No.
SAFETY EVALUATION RE: St.Lucie Unit 1 Docket No.50-335 Proposed License Amendment Linear Heat Rate Technical S ecification
-Flux Peakin Au mentation Factors I.Introduction Past practice for ECCS analysis has been to postulate that axial gaps can occur in the fuel rod pellet.stack.Such.gaps could theoretically occur because of fuel column densification in combination with an increase in the cladding ovality.With severe creep ovality, the pellet stack could be gripped by the cladding before densification is'omplete such that a gap would form between pellets as further densification occurs.This gap would lower the fuel density in a horizontal plane, resulting in an increase in thermal neutron flux and higher local rod powers.This possible power increase is used in establishing peaking factor limits.II.Evaluation Exxon Nuclear (ENC)and Florida Power 6 Light (FPL)have made a careful evaluation of the conditions that are necessary to form such gaps and have concluded that for ENC-designed fuel, such gaps will not occur.The justifi-cation for this position has been submitted to the USNRC in support of a revised clad collapse procedure in Reference l.The reasons are: 1.Densification is complete after a few thousand MWD/MT exposure.2.Ovality does not proceed to the point that pellets are gripped by the cladding until after fuel densifi-cation is complete.This conclusion is verified by the calculation of ovality and creepdown with the COLAPX and RODEX2 fuel performance codes.3.The upper plenum spring acts to keep a positive pressure on the pellets to overcome resistance of the pellet stack to downward motion.The spring is fabricated of creep resistant Inconel X-750 to avoid early load relaxation and is designed to provide positive downward pressure over the range of potential densification.
To verify the conclusion that significant gaps are not, formed, ENC has made a number of scans of irradiated rods and confirmed that axial gaps do not existed Because of these results no flux peaking augmentation factors are required for ENC designed fuel.ENC has performed neutronics calculations for Combustion Engineering designed fuel which show that, the peak rod power for CE fuel is at least 10%less than the peak power for ENC fuel during Cycle 6.Because the maximum value of flux.peaking augmentation factors is signifi-cantly less than 10%, no augmentation factor need be applied to CE fuel.Therefore, Florida Power 8 Light Company has concluded that a flux peaking augmentation factor curve need not be contained j.n St.Lucie Unit 1 Technical Specificat.ions.
III.Conclusion Based on the considerations described above, (1)the proposed change does not increase the probability or consequences qf accidents or malfunctions of equipment, important to safety and does not reduce the margin of safety as defined in the basis for any technical specification, therefore, the change does not involve a significant hazards consideration, (2)there is reason-able assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3)such activities wil'1 be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of'he public.
Reference:
(1)XN-NF-82-06,'"Qualification of'Exxon Nuclear Fuel for Extended Burnup", June 1982.
STATE OF FLORiDA)))st Robert E.Uhri being first.duly sworn, deposes and says: That he is Yice President of Florida Power a Light, Company, the L>censee herein;That he has executed the foregoing document,;
that, the state-ments made in this said.document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Robert E.Uhrig Subscribed and sworn to before me this day of 19>10TARY PUBLiC, in and for the County of Dade, State of Florida Ny commission expires: