ML22263A035
| ML22263A035 | |
| Person / Time | |
|---|---|
| Site: | Hermes File:Kairos Power icon.png |
| Issue date: | 09/19/2022 |
| From: | Hastings P Kairos Power |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| KP-NRC-2209-015 | |
| Download: ML22263A035 (12) | |
Text
KP-NRC-2209-015 Kairos Power LLC www.kairospower.com 707 W Tower Ave, Suite A 5201 Hawking Dr SE, Unit A 2115 Rexford Rd, Suite 325 Alameda, CA 94501 Albuquerque, NM 87106 Charlotte, NC 28211 September 19, 2022 Docket No. 50-7513 US Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Kairos Power LLC Transmittal of Changes to Construction Permit Application - Exemptions Enclosure, PSAR Chapter 1, PSAR Chapter 3, and PSAR Chapter 14
References:
- 1. Letter, Kairos Power LLC to Document Control Desk, Submittal of the Preliminary Safety Analysis Report for the Kairos Power Fluoride Salt-Cooled, High Temperature Non-Power Reactor (Hermes), September 29, 2021 (ML21272A376)
- 2. Audit Plan, Nuclear Regulatory Commission, Plan for a General Audit of the Hermes Construction Permit Application, February 10, 2022 (ML22039A336)
In September 2021, Kairos Power submitted a Preliminary Safety Analysis Report (PSAR) (Reference 1) as part of the Construction Permit Application (CPA) for the Hermes non-power reactor. That submittal also included a request for exemptions from 10 CFR 50.34(a)(4) and 10 CFR 50.34(b)(4) in Enclosure 3. This letter retracts the requests for exemptions and transmits the applicability of 10 CFR 50.34(a)(4) and 10 CFR 50.34(b) to address NRC feedback in the General Audit (Reference 2). This letter also transmits additional changes to PSAR Chapter 1, PSAR Chapter 3, and PSAR Chapter 14 to remove references to exemptions, pursuant to discussions with the NRC staff during the audit. The applicability of 10 CFR 50.34(a)(4) and 10 CFR 50.34(b)(4) is provided in Enclosure 1 and conforming changes to the pages of the PSAR are provided in Enclosure 2. Kairos Power requests NRC review of these changes as part of the continued review of the Hermes CPA.
If you have any questions or need any additional information, please contact Drew Peebles at peebles@kairospower.com or (704) 275-5388, or Darrell Gardner at gardner@kairospower.com or (704) 769-1226.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on September 19, 2022 Sincerely, Peter Hastings, PE Vice President, Regulatory Affairs and Quality
KP-NRC-2209-015 Page 2
Enclosures:
- 1. Applicability of 10 CFR 50.34(a)(4) and 10 CFR 50.34(b)(4) - Analysis of ECCS Performance
William Kennedy, Acting Chief, NRR Advanced Reactor Licensing Branch Benjamin Beasley, Project Manager, NRR Advanced Reactor Licensing Branch Edward Helvenston, Project Manager, NRR Advanced Reactor Licensing Branch Samuel Cuadrado de Jesus, Project Manager, NRR Advanced Reactor Licensing Branch
KP-NRC-2209-015 Applicability of 10 CFR 50.34(a)(4) and 10 CFR 50.34(b)(4) - Analysis of ECCS Performance (Non-Proprietary)
Applicabilityof10CFR50.34(a)(4)and10CFR50.34(b)(4)AnalysisofECCSPerformance (NonProprietary)
Introduction Thelastsentenceinboth10CFR50.34(a)(4)foraPSARand10CFR50.34(b)(4)foranFSARrequirean analysisofemergencycorecoolingsystem(ECCS)performanceandhighpointventsfollowing postulatedlossofcoolantaccidents.Theunderlyingpurposeoftheserequirementsisbasedonlight waterreactordesignsthatutilizeanemergencysystemtoinjectcoolingfluidsasaresultofcoolantloss duetocoolantpipebreaksandtopreventtheaccumulationofnoncondensiblegasesthatmayaffect thiscoolingfunction.TheKairosPowerfluoridesaltcooled,hightemperaturereactor(KPFHR)design removesheatfromthecorewithoutrelianceonmakeupfluidinjectionfromanemergencycorecooling system.Heatisremovedpassivelyfromthereactorvesselanddoesnotrelyonfluidadditiontoreplace coolantthatlossesfrompipebreaks.Therefore,therequirementforananalysisofECCSperformance andhighpointventsfollowingpostulatedlossofcoolantaccidentsisnottechnicallyrelevanttotheKP FHRdesign.
RegulatoryConformance Theunderlyingpurposeofthesespecificregulationsistodemonstrate,byanalysis,thecapabilityofthe emergencycorecoolingsystemtoremovedecayheatfromthereactorcorebytheinjectionofmakeup coolingfluidtoreplacefluidslostasaresultofcoolantpipingbreaks.TheKPFHRdesignsupports passivedecayheatremovaldirectlyfromthereactorvesselandreactorcorewithoutrelianceon makeupfluidinjectionfromanemergencycorecoolingsystem.TheKPFHRdesignutilizesapassive decayheatremovalsystemthatreliesondirectheatrejectiontoanexternalsetofthimblestoabsorb heatasdescribedinSection6.3ofthesafetyanalysisreport.Naturalcirculationoffluidinthereactor coretosupporttheheatremovaloccursentirelywithinthereactorvesselanddoesnotrelyon externallyconnectedcoolantsystempiping.
Thelossofcoolantfluidsinconnectedpiping(pipebreaks)arenotconsideredaccidentsandare postulatedtooccurduringplantoperations.Thedesignofthereactorvesselandconnectedpiping includessiphonbreakswhichprecludealossofvesselinventoryshouldtherebeabreakinconnected coolantpipingasdescribedinSection4.3ofthesafetyanalysisreport.Alossofthevesselintegrityto maintainfluidovertheactivecoreisnotacredibleeventbydesign.Asaresult,thesafetystrategyfor theKPFHRdesigndoesnotrelyontheinjectionofmakeupcoolantnordoesitneedahighpointthat couldbeopenedinanemergencytoremovegases.PostulatedeventsanalyzedinChapter13ofthe safetyanalysisreportdonotresultintheaccumulationofnoncondensiblegasesinthereactorvessel andcoolantlines.Noncondensiblegasesareminimizedduringnormaloperationviatheinertgas systemdescribedinSection9.1ofthesafetyanalysisreport.
Therefore,therequirementsinthelastsentenceofboth10CFR50.34(a)(4)and10CFR50.34(b)(4)are nottechnicallyrelevanttotheKPFHRdesignandtherequestedevaluationsarenotprovidedinthe PSARorFSAR.Theunderlyingpurposeoftheseregulationsissatisfiedbythedesignandanalysis describedinthesafetyanalysisreports.Becausetheregulationsarenottechnicallyrelevant,an exemptionfromtheseregulationsisnotrequired.
KP-NRC-2209-015 Changes to PSAR Chapter 1, PSAR Chapter 3, and PSAR Chapter 14 (Non-Proprietary)
PreliminarySafetyAnalysisReport
TheFacility KairosPowerHermesReactor 12 Revision0 1.2
SUMMARY
ANDCONCLUSIONSONPRINCIPALSAFETYCONSIDERATIONS TheKPFHRisanadvancedreactortechnologydevelopedintheUnitedStatesoverthelastdecade.The technologyfollowsfromDepartmentofEnergy(DOE)sponsoredresearchanddevelopmentat universitiesandnationallaboratories.ThefundamentalconceptisthecombinationofTristructural Isotropic(TRISO)particlefuelcoupledwithamoltenfluoridesaltcoolant.Thiscombinationresultsina hightemperature,lowpressurereactorsystemwithrobustinherentsafetycharacteristics.The combinationofextremelyhightemperaturetolerantfuelandlowpressure,singlephase,chemically stablereactorcoolantremovesentireclassesofpotentialfueldamagescenarios,greatlysimplifyingthe designandreducingthenumberofsafetysystems.Theintrinsiclowpressureofthereactorand associatedpiping,alongwiththefissionproductretentionprovidedbytheTRISOfuel,enhancessafety andeliminatestheneedforlowleakage,pressureretainingcontainmentstructures.Additionally,the designreliesonpassivedecayheatremovalanddoesnotneedanemergencycorecoolingsystem (ECCS)fordecayheatremovalorreplacementofcoolantinventory.Assuch,theapplicationrequiresa partialexemptionfromtherequirementsin10CFR50.34(a)and10CFR50.34(b)regardingtheanalysis ofECCSperformance.
Themajorplantsystemsarethereactorsystem(RS),theprimaryheattransportsystem(PHTS),andthe decayheatremovalsystem(DHRS).TheRSisdescribedinChapter4,thePHTSisdescribedinChapter5, andtheDHRSisdescribedinChapter6.Otherassociatedplantsupportsystemsaredescribedin Chapter7(instrumentationandcontrol),Chapter8(electrical)andChapter9(auxiliarysystems).
1.2.1 ConsequencesfromtheOperationandUseoftheFacility Akeymeasureofsafetyandconsequencefromtheoperationofthefacilityisthemagnitudeofthe potentialsourcetermassociatedwithoffnormalevents.Thesourcetermrepresentstheamount, timingandnatureofradioactivematerialreleasedandavailableforreleasetotheenvironment followingapostulatedevent.TheKPFHRdesignreliesonafunctionalcontainmentapproachtomeet thesitingregulationsin10CFR100.11(a)fordoselimits.Thefunctionalcontainmentrepresentsan engineeredsafetyfeatureofthereactorandisimplementedprincipallybythehightemperatureTRISO particlefuel.Thefuelutilizesacarbonmatrixcoatedparticlefuel,similartothatdevelopedforhigh temperaturegascooledreactors,inapebblebasedfuelelement.Coatingsontheparticlefuelhave beendemonstratedtoprovideretentionoffissionproductstodesigntemperaturesinexcessof1600°C.
ThefueldesignandperformancearediscussedfurtherinSection4.2.
Thereactorcoolantalsoprovidesasecondaryfunctionalcontainmentroleandisachemicallystable, lowpressuremoltenfluoridesaltcoolant.Themixtureconsistsofanenrichedlithiumfluoride(LiF)and berylliumfluoride(BeF2)saltsinaratioofapproximately2:1.TheMoltenSaltReactorExperiment (MSRE)programandthesubsequentoperationoftheMSREnuclearreactorutilizedthisFluoride LithiumBerylliumbasedsaltasaneffectivenuclearcoolantforboththeprimarycoolant(whichhad dissolvedfuel)andtheintermediatecoolant(whichwascleancoolant)(Reference1).Furthermore, therehasbeensignificantresearchintothestabilityandcompatibilityofthiscoolantinfissionand fusionenergyapplicationssincetheoperationoftheMSRE(Reference2).TheHermesreactoroperates withalow(nearatmospheric)overpressureinthereactorvesselheadspace.Thereactorcoolantis furtherdescribedinSection5.1.
Fissionproductretentionandcontrolinthereactorfacilityreliesonafunctionalcontainmentstrategy (comprisedoftheTRISOfuelandFlibecoolant)asameansofpreventingsignificantradionucliderelease totheenvironmentduringnormaloperationsandpostulatedevents.Thefunctionalcontainmentis describedinSection6.2.Theanalysisofpostulatedeventswhichaddressthesitinglimitsin10CFR 100.11(a)isdescribedinChapter13.
PreliminarySafetyAnalysisReport
TheFacility KairosPowerHermesReactor 14 Revision0 1.2.3 DesignFeaturesandDesignBases Theprincipaldesigncriteria(PDC)forthefacilitySSCsaredescribedinSection3.1andarebasedon thosespecifiedintheNRCapprovedKairosPowerTopicalReport,KPTR003NPA(Reference3).The systemrelatedsectionsthroughoutthisSARdescribehowthedesignbases,includingthePDC,are satisfied.
Asnotedabove,thereactordesignreliesonafunctionalcontainmentapproach,ratherthanalow leakage,pressureretainingcontainmentstructureandreactorcoolantpressureboundarythatis typicallyusedforlightwaterreactors(LWRs)tocontrolthereleaseoffissionproducts.Thefunctional containmentapproachistocontrolradionuclidesprimarilyattheirsourcewithintheTRISOcoatedfuel particleundernormaloperationsandpostulatedevents,withoutrelianceonactivesafetyfeaturesoron operatoractions.ThefunctionalcontainmentreliesprimarilyonthemultiplebarrierswithintheTRISO fuelparticlelayerstoensurethatthedoseatthesiteboundary(frompostulatedaccidents)meets regulatorylimits.Additionally,forthefuelinthereactorcore,thereactorcoolantservesasanadditional barrierprovidingretentionofmostfissionproductsthatcouldescapethefuelparticlebarriers.This additionalretentionbarrierisakeyfeatureoftheenhancedsafetyandreducedsourceterm.Toenable fissionproductretentioninthefuelparticleandthereactorcoolant,thereactorvesselmaintainsthe fuelpebblesinthereactorcoresubmergedinthecoolant.
TheSSCsinthefacilityareassignedanuclearsafetyclassification,asfollows:
SafetyrelatedSSCs:ThoseSSCsthatarereliedupontoremainfunctionalduringnormaloperating conditionsandduringandfollowingdesignbasiseventstoassure:
Theintegrityoftheportionsofthereactorcoolantboundaryreliedupontomaintaincoolant levelabovetheactivecore(seebelow);
Thecapabilitytoshutdownthereactorandmaintainitinasafeshutdowncondition;or Thecapabilitytopreventormitigatetheconsequencesofaccidentswhichcouldresultin potentialexposuresexceedingthelimitssetforthin10CFR100.11.
Nonsafetyrelated:ThoseSSCsthatarenotintheabovesafetyclassification.
Notethatthedefinitionofsafetyrelateddescribedaboveisdifferentfromthatspecifiedin10CFR50.2, Definitions.Thedefinitionin10CFR50.2isbasedonLWRtechnologieswhichrelyonareactorcoolant pressureboundaryasoneofthethreefissionproductretentionbarriers.Asdiscussedabove,theFHR technologydoesnotcreditacoolantpressureboundaryforfissionproductretention,butratherrelies onafunctionalcontainmentasdescribedinSection6.2.However,thereactorvesseliscreditedfor retainingthefuelpebblesinthereactorcoreinaFlibewettedenvironmentforheatremovalunderall postulatedevents,asdescribedinSection4.3.Notethatnootherportionofthereactorcoolant boundaryiscreditedinthesafetyanalysisforfissionproductretentionortoensuredecayheatremoval.
Therefore,anexemptionadeparturefromtothedefinitionofsafetyrelatedin10CFR50.2isrequired necessaryfortheFHRtechnologyasdiscussedinKairosPowerTopicalReportKPTR004NPA (Reference4).NotethatthetermsafetysignificantusedinReference4isnotusedinthisdefinition becausethetermisnotapplicabletotheHermesreactor,consistentwiththediscussioninSection3.1.
TheexemptionrequestwillbeprovidedintheapplicationforanOperatingLicense.
AsummaryofSSCsafetyclassificationsisprovidedinSection3.5.
Preliminary Safety Analysis Report Design of Structures, Systems, and Components Kairos Power Hermes Reactor 3-1 Revision 0 CHAPTER 3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS
3.1 INTRODUCTION
This chapter identifies and describes the principal architectural and engineering design criteria for the structures, systems, and components (SSC) that are required to ensure reactor facility safety and protection of the public. The primary safety feature of the Hermes design is the unique combination of TRISO fuel and Flibe reactor coolant. Other safety-related systems support maintaining the fuel and coolant configuration within acceptable limits. These SSCs include the safety-related portion of the Reactor Building structure, the reactor vessel and internals, the reactor control and shutdown system, and the decay heat removal system.
3.1.1 Design Criteria Kairos Power is pursuing a construction permit and subsequent operating license for the Hermes reactor under 10 CFR 50. The NRC regulations in Title 10 to the CFR have been evaluated for applicability to this facility and the results are contained in the Regulatory Analysis for the Kairos Power Fluoride Salt-Cooled, High Temperature Reactor topical report (Reference 1). The design related regulations that that are addressed by this preliminary safety evaluation report (PSAR) are summarized in Table 3.1-1 and addressed throughout this safety analysis report. In addition, this topical report identified regulations for which exemption were needed. These exemptions are identified in this safety analysis report in their applicable sections and are summarized in Table 3.1-2.
Kairos Power has also developed a set of principal design criteria (PDC) applicable for the KP-FHR technology which has been reviewed and approved by the NRC in Principal Design Criteria for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (Reference 2). The application of these criteria to the SSCs of the test reactor are shown in Table 3.1-2. The site contains only one reactor, with no SSCs shared with another reactor unit, which satisfies PDC 5. Specific details regarding how the other PDC are met by the design are described in the individual sections throughout this safety analysis report and summarized in Table 3.1-2.
Note that several of the PDCs in KP-TR-003 contain the terms safety significant, anticipated operational occurrences, and accidents. These terms are not applicable to the Hermes reactor and are not used in this safety analysis report, which represents a departure from the approved topical report. These terms are relevant to power reactors, which use frequency to bin postulated events. In the non-power reactor licensing framework, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors (NUREG-1537), the postulated events in the design basis are treated the same, regardless of frequency. Consistent with 10 CFR 50.2 (as modified - See Section 1.2.3), SSCs that are relied upon to mitigate the postulated events are classified as safety-related and a significance determination is not made in this framework. There are only two SSC classifications used in this safety analysis report for the Hermes reactor: safety-related and non-safety related. PDCs 1, 2, 3, 4, 5, 13, 14, 15, 16, 17, 18, 20, 28, 31, 32, 33, 34, 44, 61, 71, 73, 75, and 77 use the term safety significant. For these PDCs, the term safety significant is replaced in this safety analysis report with safety-related.
Additionally, PDCs 10, 13, 15, 17, 20, 26, 29, 34, 60, 64, and 73 use the term Anticipated Operational Occurrences. Since there is no distinction between AOOs and accidents in the non-power reactor licensing framework (NUREG-1537), the AOO terminology (including language that differentiates between AOOs and accidents) is replaced by "postulated events in this safety analysis report for the Hermes reactor. PDCs 2, 4, 5, 13, 16, 17, 19, 20, 22, 26, 28, 31, 35, 37, 44, 46, 61, 64, 73, and 75 use the
Preliminary Safety Analysis Report Design of Structures, Systems, and Components Kairos Power Hermes Reactor 3-2 Revision 0 term accidents, and in these instances accident is replaced with postulated events in this safety analysis report.
Note that an exemption toa departure from the 10 CFR 50.2 definition of safety-related is discussed in Section 1.2.3 with respect to the replacement of the words: integrity of the reactor coolant pressure boundary with integrity of the portions of the reactor coolant boundary relied upon to maintain coolant level above the active core. This is a departure from the proposed exemption described in Reference 1. However, as discussed above, the term safety-significant does not apply to Hermes. For Hermes, the safety-related portions of the reactor coolant boundary for the reactor are limited to portions of the reactor vessel (see Section 4.3). Failures of other SSCs containing reactor coolant (e.g.,
pipe breaks within the reactor coolant boundary) do not result in unacceptable consequences as described in Section 13.1.3. A failure of the reactor vessel is a beyond the design basis event as the vessel is designed against such failure consistent with PDC 14. Thus, the makeup inventory of reactor coolant to the reactor vessel is not relied on to mitigate the consequences of a postulated event and the requirements of PDC 33 have been addressed.
3.1.2 NRC Guidance Documents The NRC guidance documents considered in the design of the reactor are identified within this safety analysis report and are listed in Table 3.1-4. The sections cited in this table describe the extent of usage of these guidance documents. Note that Division 1 regulatory guides are not applicable to non-power test reactors and are not included in this table. In some cases, portions of the Division 1 regulatory guides were utilized and are identified in sections throughout this safety analysis report. Codes and standards used in the design of the reactor structures, systems, and components that contain radioactivity are provided in Section 3.6. Other codes and standards are also identified throughout the report.
3.1.3 References
- 1. Kairos Power, LLC, Regulatory Analysis for the Kairos Power Salt-Cooled, High Temperature Reactor, KP-TR-004-P, Revision 3. August 2020.
- 2. Kairos Power, LLC, Principal Design Criteria for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor, KP-TR-003-P-A. July 2019.
Preliminary Safety Analysis Report Design of Structures, Systems, and Components Kairos Power Hermes Reactor 3-5 Revision 0 Table 3.1-2: Regulations Requiring Exemptions for the Hermes ReactorNot Used Regulation Title (or subject of regulation)
SAR Section 10 CFR 50.2 (Definitions
- Reactor Coolant Pressure Boundary)
Reactor coolant pressure boundary 4.3 10 CFR 50.2 (Definitions -
Safety-related structures, systems, and components)
Reactor coolant pressure boundary 1.2.3, 3.6 10 CFR 50.34(a)(4) and (b)(4)
Analysis of SSCs and ECCS 1.2 10 CFR 50.34(b)(9)
Pressurized Thermal Shock This exemption will be provided as part of the application for an operating license.
Reactor coolant pressure boundary 14.1
Preliminary Safety Analysis Report Design of Structures, Systems, and Components Kairos Power Hermes Reactor 3-32 Revision 0 3.6.2 Classification of Structures, Systems, and Components SSCs are assigned safety, seismic, and quality classifications consistent with their safety functions. These classifications are described below. Table 3.6-1 provides a summary of these classifications for all SSCs.
3.6.2.1 Safety Classification SSCs have two possible safety classifications: safety-related or non-safety related. An SSC is classified as safety-related if it meets the definition of safety-related from 10 CFR 50.2 (with exceptions as described in Section 1.2.3). For the KP-FHR technology, the definition of safety-related is modified from 10 CFR 50.2, to be:
Safety-related structures, systems, and components means those structures, systems, and components that are relied upon to remain functional during and following design basis events to assure:
(1) The integrity of the portions of the reactor coolant boundary relied upon to maintain coolant level above the active core; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1) or 10 CFR 100.11 Note that for the KP-FHR technology, the definition above reflects an exemptiona departure from the definitions in 10 CFR 50.2 for light water reactors that include the terminology integrity of the reactor coolant pressure boundary. As described in Section 1.2.3 and the Regulatory Analysis for the Kairos Power Salt-Cooled, High Temperature Reactor Topical Report (Reference 1), this exemption departure is necessary because the technology associated with the KP-FHR is based on a near atmospheric pressure design and the reactor coolant boundary does not provide a similar pressure related or fission product retention function as light-water reactors for which these definitions were based.
SSCs that do not meet the definition, as modified above, are classified as non-safety related.
3.6.2.2 Seismic Classification SSCs are classified in one of two Seismic Design Categories (SDC) consistent with ASCE 43-19 (Reference 2). Safety-related SSCs are classified as SDC-3. Section 3.4 discusses the SDC-3 classification and Section 3.5 discusses requirements for SSCs that are required to maintain their function in the event of a design basis earthquake. All safety-related SSCs are located in the safety-related portion of the Reactor Building, which is discussed in Section 3.5.1.
The credited safety systems designed to function in a postulated event are described in Chapter 13. For a design basis earthquake, the SDC-3 SSCs that are relied upon to perform a specific credited safety function are listed in Table 3.6-1.
Safety-related systems and components are qualified to maintain their safety function during a design basis earthquake, after a design basis earthquake, or both, depending on the function performed. For example, the reactor vessel is required to perform its safety function (i.e., maintain structural integrity) both during and after a design basis earthquake, whereas the decay heat removal system is required to perform its safety function only after the event, and not during. The specific safety function, therefore, is used to define the ASCE 43-19 Limit State that is used to qualify the SDC-3 SSCs.
Seismic qualification is accomplished through analysis, testing or a combination of those methods.
Acceptance criteria is defined in accordance with ASCE 43-19, Chapter 8, and/or its references.
PreliminarySafetyAnalysisReport
TechnicalSpecifications
KairosPowerHermesReactor 141 Revision0 CHAPTER14 TECHNICALSPECIFICATIONS
14.1 INTRODUCTION
Inaccordancewith10CFR50.34(a)(5),thevariablesandconditionsthatareexpectedtobesubjectto technicalspecificationcontrolforthetestreactorfacilityareprovidedinTable14.11.Thesevariables andconditionsaretheresultofthepreliminarysafetyanalysesdescribedelsewhereinthisreport.
ThetechnicalspecificationsandparameterlimitswillbesubmittedwiththeapplicationforanOperating License,consistentwith10CFR50.34(b)(6)(vi)andaddresstherequirementsin10CFR50.36.Notethat inaKairosPowerFluorideSaltCooledHighTemperatureReactor(KPFHR),thereactorcoolant boundarydoesnotserveafissionproductbarrierfunction.Fissionproductretentionisprovidedbythe functionalcontainmentdescribedinSection6.2.Therefore,thelanguagein10CFR50.36(c)(2)(ii),
significantabnormaldegradationofthereactorcoolantpressureboundary,isnotapplicableandwill bereplacedbysignificantabnormaldegradationofthefunctionalcontainment.withanexemption requestasdescribedintheRegulatoryAnalysisTopicalReport(Reference14.31).Theexemption requestwillbeprovidedwiththeapplicationforanOperatingLicense.
Theformatandcontentofthetechnicalspecificationsareconsistentwiththeguidanceprovidedin AmericanNationalStandardsInstitute(ANSI)/AmericanNuclearSociety(ANS)15.1,TheDevelopment ofTechnicalSpecificationsforResearchReactors(ANSI/ANS,2007)andinclude:
SafetyLimitsandLimitingSafetySystemSettings LimitingConditionsforOperation SurveillanceRequirements DesignFeatures AdministrativeControls
14.2 OPERATINGMODES TheoperationalmodesforthereactoraresummarizedinTable14.21.Eachoperationalmodeis definedintermsofcombinationsofcorereactivity,reactorpower,andnominaloutletreactorcoolant temperature.Thesemodesaredescribedindividuallyinthefollowingsubsections.
14.2.1 MODE1:FullPower Inthismode,thereactoriscriticalandthethermaloutputrangesbetween20%and100%ofrated power.Inthismodereactortemperature(outlet)isbetween550°C650°C.Thispowerlevelmaybe desiredfortestingpurposessuchasirradiationdatacollection,transientmaneuvers,andothersystem testingatelevatedpowerlevels.Higherpowersmaybedesiredfortestingpurposessuchasthe collectionofirradiationdataandsystemperformanceevaluations.Thismodeisachievedduringa controlledpowerascensionfromMODE2andisdeclaredwhenthereactorpowerreaches20%or higher.MODE1maybeexitedaspartofacontrolledpowerreductiontoMODE2orautomaticallyasa resultofareactortriptoMODE3.
14.2.2 MODE2:LowPower Inthismode,thereactorrangesfromthestatepointofhotzeropower(alsoknownaszeropower critical)uptolessthan20%ofratedpower.Thisisthepointwherethereactoriscriticalbutwithvery lowneutronfluxandverylownuclearheatgeneration(maintainedattemperaturewiththereactor auxiliaryheatingsystem(seeSection9.1.5.1)asneeded).Inthismodethereactortemperature(outlet)