L-2011-021, St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 4; Technical Specifications Bases Markups
| ML110730298 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/25/2011 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| L-2011-021 | |
| Download: ML110730298 (23) | |
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SECTION NO.:PAGE: 2.0REVISION NO.:3 of 10 4TITLE:TECHNICAL SPECIFICATIONSBASES ATTACHMENT 1 OF ADM-25.04SAFETY LIMITS AND LIMITING SAFETY SETTINGSST. LUCIE UNIT 2BASES FOR SECTION 2.02.1 SAFETY LIMITSBASES2.1.1REACTOR COREThe restrictions of this safety limit prevent overheating of the fuel cladding andpossible 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 resultin 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 or ABB-NV correlation. The CE-1 and ABB-NV DNB correlations have been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, 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, normaloperational transients, and anticipated transients is limited to the appropriate correlation limit for DNB-SAFDL in conjunction with the Extended Statistical Combination of Uncertainties (ESCU) or the revised Thermal Design Procedure (RTDP). This value is derived through a statistical combination of the systemparameter probability distribution functions with the CE-1 or ABB-NV DNB correlation uncertainties. This value corresponds to a 95% probability at a 95%confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
/R4SpecifiedAcceptableFuelDesignLimitfor DNB(DNB-SAFDL) has SECTION NO.:PAGE: 2.0REVISION NO.:4 of 10 4TITLE:TECHNICAL SPECIFICATIONSBASES ATTACHMENT 1 OF ADM-25.04SAFETY LIMITS AND LIMITING SAFETY SETTINGSST. LUCIE UNIT 22.1SAFETY LIMITS (continued)BASES (continued)2.1.1REACTOR CORE (continued)The curves of Figure 2.1-1 show conservative loci of points of THERMALPOWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the DNB-SAFDL is not violated based on the ABB-NV CHF correlation for the reference 1.55 Chopped Cosine Axial Shape and Design Limit shown in Figure B 2.1-1. The dashed line is not a safety limit; however, operation above this line 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 107% of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in Table 2.2-1. The area of safe transient condition is below and to the left of these lines.The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to bevalid are shown on the figure.
The Thermal Margin/Low Pressure and Local Power Density Trip Systems, inconjunction with Limiting Conditions for Operation, the Variable Overpower Trip and the Power Dependent Insertion Limits, assure that the Specified Acceptable Fuel Design Limits on DNB and Fuel Centerline Melt are notexceeded during normal operation and design basis Anticipated Operational Occurrences. Specific verification of the DNB-SAFDL limit using an appropriate DNB correlation ensures that the reactor core safety limit is satisfied.2.1.2REACTOR COOLANT SYSTEM PRESSUREThe restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.The Reactor Coolant System components are designed to Section III, 1971Edition including Addenda to the Summer, 1973, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.The entire Reactor Coolant System was hydrotested at 3125 psia todemonstrate integrity prior to initial operation.
/R4limitF T r vessel5inlet DNB-SAFDL SECTION NO.:PAGE: 2.0REVISION NO.:8 of 10 4TITLE:TECHNICAL SPECIFICATIONSBASES ATTACHMENT 1 OF ADM-25.04SAFETY LIMITS AND LIMITING SAFETY SETTINGSST. LUCIE UNIT 22.2LIMITING SAFETY SYSTEM SETTINGS (continued)BASES (continued)2.2.1REACTOR TRIP SETPOINTS (continued)Containment Pressure-HighThe Containment Pressure-High trip provides assurance that a reactor trip isinitiated prior to or concurrently with a safety injection (SIAS). This alsoprovides assurance that a reactor trip is initiated prior to or concurrently with an
MSIS.Steam Generator Pressure-LowThe Steam Generator Pressure-Low trip provides protection against anexcessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setting of 626 psia is sufficiently below the full load operating point 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 30 psi in the safety analyses.Steam Generator Level-LowThe Steam Generator Level-Low trip provides protection against a loss offeedwater flow incident and assures that the design pressure of the Reactor Coolant System will not be exceeded due to loss of the steam generator heat sink. This specified setpoint provides allowance that there will be sufficientwater inventory in the steam generator at the time of the trip to provide sufficient time for any operator action to initiate auxiliary feedwater beforereactor coolant system subcooling is lost. This trip also protects against violation of the specified acceptable fuel design limits (SAFDL) for DNBR, offsite dose and the loss of shutdown margin for asymmetric steam generator transients such as the opening of a main steam safety valve or atmospheric dump valve.Thetripsetpointisboundingrelativetotheaccidentandtransientanalyseswhichwereperformedusingalower,conservativetripsetpoint.Thetrip setpointandthemethodologyusedtodeterminethetripsetpoint,theas-foundacceptancecriteriaband,andtheas-leftacceptancecriteriaare specifiedintheUFSAR.ThetwofootnotesonthebottomofTSTable2.2-1 areconsistentwiththetworecommendednotesprovidedinNRC'sletterto theNEITechnicalSetpointMethodsTaskForceforSetpointAllowables datedSeptember7,2005.
SECTION NO.:PAGE:3/4.1REVISION NO.:5 of 9 3TITLE:TECHNICAL SPECIFICATIONSBASES ATTACHMENT 3 OF ADM-25.04REACTIVITY CONTROL SYSTEMSST. LUCIE UNIT 23/4.1REACTIVITY CONTROL SYSTEMS (continued)BASES (continued)3/4.1.2BORATION SYSTEMSThe boron injection system ensures that negative reactivity control isavailable 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 makeup pumps, and (5) an emergency power supply from OPERABLE diesel generators.With the RCS average temperature above 200
°°°°F, a minimum of two separateand redundant boron injection systems are provided to ensure singlefunctional 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 injection system failures during the repair period.The boration capability of either system is sufficient to provide a SHUTDOWNMARGIN from expected operating conditions of the limit specified in theCOLR after xenon decay and cooldown to 200
°°°°F. The maximum expectedboration capability requirement occurs at EOL from full power equilibriumxenon conditions. This requirement can be met for a range of boric acid concentrations in the Boric Acid Makeup Tank (BAMT) and Refueling Water Tank (RWT). This range is bounded by 5350 gallons of 3.5 weight percent (6119 ppm boron) from the BAMT and 16,000 gallons of 1720 ppm borated water from the RWT to 8650 gallons of 2.5 weight percent (4371 ppm boron) boric acid from BAMT and 12,000 gallons of 1720 ppm borated water from the RWT. A minimum of 35,000 gallons of 1720 ppm boron is required from the RWT if it is to be used to borate the RCS alone.With the RCS temperature below 200
°°°°F one injection system is acceptablewithout single failure consideration on the basis of the stable reactivitycondition of the reactor and the additional restrictions prohibiting COREALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.
8750 1900 3.5 6119 1900 11,692 7550 10,492 3.1 5420 33 , 000Thisvolumerequirement,however,isexpectedtoalwaysbeboundedbytheECCSRWTvolumerequirementsofSpecification3.5.4.
SECTION NO.:PAGE:3/4.1REVISION NO.:6 of 9 3TITLE:TECHNICAL SPECIFICATIONSBASES ATTACHMENT 3 OF ADM-25.04REACTIVITY CONTROL SYSTEMSST. LUCIE UNIT 23/4.1REACTIVITY CONTROL SYSTEMS (continued)BASES (continued)3/4.1.2BORATION SYSTEMS (continued)Temperature changes in the RCS impose reactivity changes by means of themoderator temperature coefficient. Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SDM.
Small changes in RCS temperature are unavoidable and so long as the required SDM is maintained during these changes, any positive reactivity additions will be limited to acceptable levels. Introduction of temperature changes must be evaluated to ensure they do not result in a loss of required SDM.The boron capability required below 200
°°°°F is based upon providing aSHUTDOWN MARGIN corresponding to its COLR limit after xenon decay andcooldown from 200
°°°°F to 140°°°
°F. This condition requires either 6750 gallons of1720 ppm - 2100 ppm borated water from the refueling water tank or 3550gallons of 2.5 to 3.5 weight percent boric acid solution from the boric acid makeup tanks.The contained water volume limits includes allowance for water not availablebecause of discharge line location and other physical characteristics.The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.The limits on contained water volume and boron concentration of the RWT also ensure a pH value of between 7.0 and 8.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.Ensuring that the BAM pump discharge pressure is met satisfies the periodic surveillance requirement to detect gross degradation caused by impeller structural damage or other hydraulic component problems. Along with this requirement,Section XI of the ASME Code verifies the pump developed head at one point on the pump characteristic curve to verify both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the unit safety analysis. Surveillance Requirements are specified in the In-service Testing Program, which encompassesSection XI of the ASME Code.Section XI of the ASME Code provides the activities and frequencies necessary to satisfy the requirements.
1900 3.1canbesatisfiedbymaintaining SECTION NO.:PAGE:3/4.2REVISION NO.:6 of 6 2TITLE:TECHNICAL SPECIFICATIONSBASES ATTACHMENT 4 OF ADM-25.04POWER DISTRIBUTION LIMITSST. LUCIE UNIT 23/4.2POWER DISTRIBUTION LIMITS (continued)BASES (continued)3/4.2.5DNB PARAMETERSThe limits on the DNB-related parameters assure that each of the parametersare maintained within the normal steady-state envelope of operation assumed in the transient and safety analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the appropriate correlation limit for DNB-SAFDL in conjunction with ESCU or RTDP methodology throughout each analyzed transient.These variables are contained in the COLR to provide operating and analysisflexibility from cycle to cycle. However, the minimum RCS flow based on maximum analyzed steam generator tube plugging, is retained in the TS LCO. Operating within these limits will result in meeting the DNBR criterion in the event of a DNB limited transient.The 12-hour periodic surveillance of these parameters through instrumentreadout is sufficient to ensure that the parameters are restored within theirlimits 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-hour basis.
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ReplacewithREACTORCOOLANTSYSTEMInsert1 The maximum allowable doses to an individual at the exclusion area boundary (EAB) distance for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an accident, or at the low population zone (LPZ) outer boundary distance for the radiological release duration, are specified in 10 CFR 50.67 for design basis accidents using the alternative source term methodology and in Branch Technical Position 11-5 for the waste gas decay tank rupture accident. Dose limits to control room operators are given in 10 CFR 50.67 and in GDC 19.
The RCS specific activity LCO limits the allowable concentration of radionuclides in the reactor coolant to ensure that the dose consequences of limiting accidents do not exceed appropriate regulatory offsite and control room dose acceptance criteria. The LCO contains specific activity limits for both DOSE EQUIVALENT (DE) I-131 and DOSE EQUIVALENT (DE) XE-133.
The radiological dose assessments assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator tube leakage rate at the applicable Technical specification limit. The radiological dose assessments assume the specific activity of the secondary coolant is at its limit as specified in LCO 3.7.1.4, Plant Systems - Activity.
The ACTIONS allow operation when DOSE EQUIVALENT I-131 is greater than 1.0 Ci/gram and less than 60 Ci/gram. The ACTIONS require sampling within four hours and every four hours following to establish a trend.
One surveillance requires the determination of the DE XE-133 specific activity as a measure of noble gas specific activity of the reactor coolant at least once per 7 days.
A second surveillance is performed to ensure that iodine specific activity remains within the LCO limit once per 14 days during normal operation and following rapid power changes when iodine spiking is more apt to occur. The frequency between two and six hours after a power change of greater than 15% RATED THERMAL POWER within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established because the iodine levels peak during this time following iodine spike initiation.
The RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
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154,000 130,500 9203 an 4230 325of307,000gallonsrequiredbytheLCO Thealsoincludes SECTION NO.:PAGE:3/4.11REVISION NO.:3 of 3 0TITLE:TECHNICAL SPECIFICATIONSBASES ATTACHMENT 13 OF ADM-25.04RADIOACTIVE EFFLUENTSST. LUCIE UNIT 2BASES FOR SECTION 3/4.113/4.11RADIOACTIVE EFFLUENTSBASESPages B 3/4 11-2 through B 3/4 11-3 (Amendment No. 61) have beendeleted from the Technical Specifications. The next page is B 3/4 11-4.3/4.11.2.5EXPLOSIVE GAS MIXTUREThis specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdupsystem is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.3/4.11.2.6GAS STORAGE TANKSRestricting the quantity of radioactivity contained in each gas storage tankprovides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, Waste Gas System
Failure.gaseousradioactivewasteinventoryinaeffectivedoseequivalentBranchTechnicalPosition11-5,"PostulatedRadioactiveReleaseDuetoWasteGasSystemLeakorFailure,"ofStandardReviewPlanChapter 11,RadioactiveWasteManagement,"ofNUREG-0800.Thewastegas decaytankinventoryofnoblegasesrequiredtogenerateanexclusion areaboundarydoseof0.1remisthebasisforthelimitof202,500dose equivalentcuriesXe-133,andisderivedbasedonthedefinitiongivenin TechnicalSpecificationTaskForce(TSTF-490),"DeletionofEbar DefinitionandRevisiontoRCSSpecificActivityTechSpec." 0.1