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RBSUSARRevision14 12.1-1September2001CHAPTER12RADIATIONPROTECTIONThischapterprovidesinformationrelevanttotheradiationprotectionfeaturesofthefacilityandequipmentdesign,plans andpracticestoaccomplishradiologicalcontrol,andestimates ofoccupationalradiationexposuretooperatingandcontractual personnelduringnormaloperationandanticipatedoperational
occurrences.12.1.ENSURINGTHATOCCUPATIONALRADIATIONEXPOSURESAREASLOWASISREASONABLYACHIEVABLE(ALARA)12.1.1PolicyConsiderations97ItisthepolicyofRiverBendStationtokeepallradiationexposureofpersonnelwithinthelimitsestablishedbythe federalgovernmentassetforthinTitle10oftheCodeof FederalRegulations.Administrativeproceduresandpractices maintainingradiationexposureofpersonnelaslowasis reasonableachievable(ALARA)aredelineatedherein.The specificRiverBendStationpositionsonDivision8Regulatory Guides,includingRegulatoryGuides,8.8and8.10,areprovided inSection1.8.Thebasesforthesepositionsarecontainedin thediscussionsfoundinthissectionandthroughoutChapter12.
TheRiverBendStationALARApolicyisconsistentwiththe guidelinesinSectionC.1ofRegulatoryGuide8.8andRegulatory Guide8.10inestablishing,organizing,andoperatingan effectiveALARAprogram.
79ThepurposeoftheALARAprogramistomaintaintheradiationexposureofplantpersonnelasfarbelowtheregulatorylimitsas isreasonablyachievable.14137TheSiteVicePresidentisresponsibleforimplementingtheALARAprogramby:1)ensuringthattheresourcesneededtoachieve thegoalsandobjectivesoftheALARAprogramareavailable,and 2)ensuringthattheauthorityfordevelopingtheproceduresand practicesbywhichthegoalsandobjectivesareachieved, measured,andevaluatedisdelegatedtotheappropriatepersonnel (Section13.1).TheRiverBendStationGeneralManagerreportstotheSiteVicePresidentandisresponsibleforthedevelopmentofplansfor reviewandapprovaloftheradiationprotectionprogramsto ensurethatradiationexposureofpersonnelismaintainedALARA.
71314 RBS USAR Revision 20 12.1-2137The River Bend Station General Manager also assumes responsibility of the overall effectiveness of the ALARA program at the time of fuel loading. This responsibility is carried out
through the Manager-Radiation Protection.The Manager-Radiation Protection is responsible for thedevelopment of the required radiation protection programs, including the ALARA program. He reports directlyto theGeneral Manager. He reports to management on the effectivenessof the ALARA program. A radiation protection supervisor reports to the Manager-Radiation Protection. This supervisor is responsible for the development of the ALARA program. He provides technical assistance in conducting the ALARA program
and in improving the effectiveness of the ALARA program.
13The ALARA Coordinator reports to a radiation protection supervisor and is responsible for coordinating and implementing
the established radiation protection ALARA program. He is
responsible for maintaing occupational radiation exposure of personnel ALARA. He is responsible for conducting design reviews for plant modifications, plant procedures, and work
practices to assure that ALARA objectives are being met.
7Further, it is the responsibility of each supervisor to enforcethe requirements for keeping radiation exposure ALARA and the responsibility of each individual to comply with these
requirements.12.1.2 Design Considerations
This section discusses the general methods and features that implement the policy considerations of Section 12.1.1. Detailed provisions for maintaining personnel exposures ALARA are
presented in Sections 12.3.2 and 12.5.3.8The ALARA philosophy was used during the design and constructionstages of the River Bend Station by General Electric and SWEC engineers in their continuous reviews of plant design and equipment selection for River Bend Station. Operating experience at BWR's has been incorporated into the River Bend
Station design for equipment selection and plant layout to help maintain occupational radiation exposure as low as reasonable
achievable.
87During the design and construction stages, River Bend Station performed an independent ALARA design review of River Bend
Station. The design review of plant systems and facilities was 7
RBSUSARRevision7 12.1-3January19957organizedanddevelopedwithintheRadiationProtectionSection,thenunderthedirectionoftheSupervisor-Radiological Programs.AnALARACoordinatorwasassignedtheresponsibility todevelop,performanddocumenttheALARAdesignreviews.The ALARACoordinatordirectedtheactualdesignreviewsthroughthe ALARAReviewGroup,whichconsistedofGSURadiationProtection personnelandoutsideconsultantsexperiencedinradiation protection,plantdesign,operations,andmaintenance.TheALARA ReviewGroupperformedanddocumentedtheirreviewsinareport totheALARACommittee.TheALARACommitteewaschairedbythe Supervisor-RadiologicalPrograms,orasappointedbythePlant Manager,andconsistsofGSUrepresentativesfromtheOperations, Maintenance,Chemistry,RadiationProtection,andEngineering Sections.SWECandGeneralElectricengineerswereconsultedby theALARACommitteefordesigninformationandforimplementing designchanges.TheALARAReviewGroupbasedontheirreportsof theplantsystemsandfacilityreviews.TheactionsoftheALARA CommitteewereapprovedbytheRiverBendStationPlantManager foralldesignmodificationsandadministrativecontrols recommendedbytheCommitteetomaintainoccupationalradiation exposureALARA.TheALARACommitteeactionsaredocumentedwith ALARACommitteeActionItems,whichprovideahistoryforevery ALARAconcernidentifiedbytheALARAReviewGroup.
712.1.2.1GeneralDesignConsiderationsforALARAExposuresThegeneraldesignconsiderationsandmethodsemployedtomaintainin-plantradiationexposureALARAhavetwoobjectives:1.Minimizingtheamountoftimeplantpersonnelspendinradiationareas2.Minimizingradiationlevelsinroutinelyoccupiedplantareasandinthevicinityofplantequipmentexpectedtorequiretheattentionofplantpersonnel.BothequipmentdesignandarrangementareconsideredinmaintainingexposuresALARAduringplantoperation,including:
normaloperation,radwastehandling,normalmaintenance, correctivemaintenance,refueling,inserviceinspection,other eventsofmoderatefrequency,andcertaininfrequentlyoccurring
events.
RBSUSAR 12.1-4August1987Theseconsiderationsareimplementedduringthedevelopmentofdesigndrawingsinthefollowingmanner:1.EngineeringprovidesDesignwithgeneralguidanceconcerningALARAconsiderations.2.EquipmentlayoutandpipingdrawingsarethenreviewedandapprovedbyresponsibleengineeringpersonneltoensurethatALARAconsiderationshavebeenincorporated intothedesignpriortobeingissuedbythem.3.Equipmentlayoutdrawingsarereviewedandrevisedasnecessarybypersonneltrainedinradiationprotectionto ensurethatlayoutandshieldingprovidedisadequateto meetestablishedradiationzones.12.1.2.2EquipmentGeneralDesignConsiderationsforALARAEquipmentgeneraldesignconsiderationstominimizetheamountoftimeplantpersonnelspendinaradiationareainclude:1.Reliability,durability,construction,anddesignfeaturesofequipment,components,andmaterialstoreduceoreliminatetheneedforrepairorpreventative
maintenance.2.Servicingconvenienceforanticipatedmaintenanceorpotentialrepair.3.Provisions,wherepracticable,toremotelyormechanicallyoperate,repair,service,monitor,or inspectequipment(includinginserviceinspectionin accordancewithASMECode,SectionXI).4.Redundancyofequipmentorcomponentstoreducetheneedforimmediaterepairwhenradiationlevelsmaybehigh.Equipmentgeneraldesignconsiderationsdirectedtowardminimizingradiationlevelsinproximitytoequipmentorcomponentsrequiringtheattentionofpersonnelinclude:1.Provisionsfordraining,flushing,orifnecessaryremotecleaningofequipmentandpipingcontainingradioactive material.
RBSUSAR 12.1-5August19872.Designofequipment,piping,andvalvestominimizethebuildupofradioactivematerialandtofacilitateflushingofcrudtraps.3.Utilizationofweldedconnections,wherefeasible,insteadofflangedorthreadedconnections,valves,valve packings,andgasketstominimizeleakageandspillageof radioactivematerials.Insomecases,radwasteequipment isflangedinordertominimizeremovaltime.4.Provisionsforisolatingequipmentfromradioactiveprocessfluids.12.1.2.3PlantLayoutGeneralDesignConsiderationsforALARAPlantgeneraldesignconsiderationstominimizetheamountoftimepersonnelspendinradiationareasinclude:1.Locatingequipment,instruments,andsamplingstations,whichrequireroutinemaintenance,calibration, operation,orinspection,foreaseofaccessandminimum requiredoccupancytimeinradiationarea.2.Layingoutplantareastoallowremoteormechanicaloperation,service,monitoring,orinspectionofhighly radioactiveequipment.3.Providing,wherepracticable,fortransportationofequipmentorcomponentsrequiringservicetoalower radiationarea.Plantgeneraldesignconsiderationsdirectedtowardminimizing radiationlevelsinplantareasandinthevicinityofequipment requiringtheattentionofpersonnelinclude:1.Separatingradiationsourcesandoccupiedareaswherepracticable(e.g.,pipesorductscontainingpotentially high-radioactivefluidsnotpassingthroughoccupied
areas).2.Providingadequateshieldingbetweenradiationsourcesandaccessandserviceareas.3.Locatingequipment,instruments,andsamplingsitesinthelowestpracticableradiationzone.
RBSUSAR 12.1-6August19874.Providingcentralcontrolpanelstopermitremoteoperationofallessentialinstrumentationandcontrolsfromthelowestradiationzonepracticable.5.Separatinghighlyradioactiveequipment,wherepracticableforpackageunits,fromlessradioactive equipment,instruments,andcontrols.6.Providingmeansandadequatespaceforutilizingmovablesheildingforsourceswithintheserviceareawhen
required.7.Providingmeanstocontrolcontaminationandtofacilitatedecontaminationofpotentiallycontaminated areaswherepracticable.8.Providingmeansfordecontaminationofserviceareas.9.Providingremotelyoperatedbackflushablefiltersystemsforhighlyradioactiveradwasteandcleanupsystems.10.Providinglabyrinthentrancestoradioactivepump,equipment,andvalverooms,asrequired.11.Providingadequatespaceinlabyrinthentrancesforeasy access.12.Maintainingventilationairflowpatternsfromareasoflowerradioactivitytoareasofhigherradioactivity.12.1.2.4DecommissioningDesignConsiderationsforALARATheradiationprotectionaspectsoftheplantdesignalsoaidintheALARAaspectsofdecommissioning.Theseincludethe
following:1.Accessibilityformaintenanceorremovalofequipment.2.Sheildingtoprovideprotectionduringmaintenanceorduringstorageafterterminationofplantoperations.3.Provisionsfordraining,flushing,ordecontaminatingequipmentorpiping.
RBSUSAR 12.1-7August19874.Separationofmoreradioactiveequipmentfromlessradioactiveequipment.5.Featurestominimizecrudbuildup.6.Coatingsappliedtosurfaceslikelytobecomecontaminatedtofacilitatecleanup.12.1.2.5IllustrativeExamplesofALARAImprovements Designimprovements,whichindicateman-remreductionsduringoperationandmaintenance,havebeenimplementedfollowingALARA reviews.Someexamplesfollow.ImprovementBasedonOperationalExperienceOperationalexperiencewithradwastefilters,sludgeprocessing,andradwastesolidificationsystemswasutilizedinthedesign andselectionofcomponentsforsystems.Usewasalsomadeof studiesofoperatingplantexperiencewithsimilarsystemsto determinesystemfailurerates,downtimes,numberofpersonnel andman-hoursrequiredtorepairthesystems,andtheman-rem associatedwithrepairs.OperatingexperiencefromotherunitsandtheALARAdesignfeaturesofRiverBendStationareutilizedinthedevelopment andrevisionofstationoperatingandmaintenanceproceduresand instructionstoensurethatoccupationalradiationexposureis maintainedALARA.ImprovementBasedonALARADesignReviewGSUhasperformedanddocumentedanALARAdesignreviewofRiverBendStationduringthedesignandconstructionstagesas describedinsection12.1.2.Duringthereview,theRiverBend Stationplantfacilitiesandsystemswerestudiedindetailto documentthedesignfeaturesandanypotentialdesign improvementsthatcouldminimizetheoccupationalradiation exposuretopersonnel.Theimprovementsmadeasaresultofthe ALARAdesignreviewcanbecategoriziedaseitheradministrative controlsordesignchanges.TheadministrativecontrolsthatwereimplementedtominimizetheoccupationalradiationexposureatRiverBendStationincludethe
following:1.TheidentificationofpotentialcrudtrapstotheRadiationProtectionSectionforincorporationintotheirradiationsurveyspriortomaintenanceor RBSUSAR 12.1-8August19872.operationactivitiesinareasnearthesepotentialhot spots.3.Theidentificationoftheneedtoemphasizetrainingofmaintenancepersonnel,includinghands-onexperienceinworkinginthecontrolroddrive(CRD)maintenancearea, priortohandlingofradioactiveCRD's.Thisisessential duetothehighradiationlevelsassociatedwithCRD
maintenance.4.TheidentificationoftrainingneedsforRadiationProtectiontechniciansintheareasofcontaminationcontrol andradiologicalhazardswiththeCRDchangeout, transportation,andrebuildactivities.5.Theuseofcondensatewaterthroughouttheplantforasourceofflushinganddecontaminatingwatertominimizethe amountofradwastewatergenerated.6.TheidentificationoftrainingneedsforOperationsandMaintenancepersonnelonthecomponentidentificationsystem atRiverBendStationtominimizeunnecessaryradiation exposureduetoidentificationerrors.7.Operatingproceduresforthereactorwatercleanupsystemtominimizedumpingofreactorcoolantwatertothecondenser tocontrolbuildupofcorrosionproductsinthecondenser andthecondensatesystem.8.TheidentificationoftheplantmodeltotheTrainingDepartmentasausefulALARAtoolforpre-jobplanning.9.Theuseofadedicatedsetofoperatorsforradwasteoperationswillimprovetheefficiencyofradwaste processingandthusminimizethemaintenancerequired.10.Incorporationofadequateflushinginoperatingproceduresforreactorcoolant,radwasteliquid,andresintransfersto minimizeradioactivecorrosionproductbuildup.11.TheidentificationoftrainingneedsforMaintenancepersonnelonthemaintenanceactivitiesassociatedwithmajorcomponentsinthedrywell, RBSUSAR 12.1-9August198712.suchasthemainsteamisolationvalvesandrecirculationpumps,inordertominimizeradiationexposureinpotentiallyhighradiationareas.13.Thecontrolofpersonnelaccesstoareasnearthefueltransfertubeduringfueltransferoperations,usinglockedgates,interlocks,andradiationmonitors.Thedesignchangesthatwereimplementedtominimizethe occupationalradiationexposureatRiverBendStationincluded thefollowing:1.InstallationofflushconnectionsontheCRDscramdischargevolumeheaderstominimizeradiationfieldsdue tocorrosionproductbuildup.2.Modificationofthecondensersheildingtominimizeradiationfieldsinhightrafficareasnearthereactor feedwaterpumpsandinstrumentrackslocatedinthis
area.3.Installationoffloorcurbsthroughouttheplanttocontrolthespreadofliquidcontaminationduetoleakage frompumps,valves,etc.4.Installationofflushconnectionsontheliquidradwasteinletheaderstominimizecorrosionproductbuildup.5.Improvedaccessibilitytoplantsystemcomponentsformaintenanceandoperationactivitiesviaplatforms, verticalladders,catwalks,etc.6.Paintingofwallsandfloorsinpotentiallycontaminatedareasforeaseofdecontamination.7.Useofcondensatewaterforroutinesourcesofflushanddecontaminationwaterthroughouttheplant.8.Installationofjunctionboxesforusingportablecontinuousairmonitorsconnectedtothedigital radiationmonitoringsystem.9.Designimprovementstothereactorwatercleanuppumpsealdesigntoimprovepumpreliability.
RBS USAR Revision 17 12.1-10 10. Modification of sheilding around reactor water cleanup backwash tank to minimize radiation exposure in adjacent pump cubicle. 11. Relocation of solenoid operators, instruments, condensate and service air isolation valves outside high radiation
areas.12. Installation of remote monitoring equipment for the condensate demineralizer regeneration process equipment. 12.1.3 Operational Considerations 137 In accordance with company policy, the radiation exposure of plant personnel is maintained ALARA by means of the radiation
protection program presented in Section 12.5.
The radiation protection training provided at River Bend Station is a primary means of instituting the operational ALARA program.
The extent of this training provided each person is at least commensurate with his job responsibilities and plant areas frequented. The radiation protection training program is
approved by the Manager-Radiation Protection and conducted by either the training department or radiation protection personnel.
13The Radiation Protection Plan is another implementation element of the operational ALARA program. The plan contains the operational ALARA philosophy and is made available to plant
personnel. The plan defines management's commitment to ALARA and also designates those individuals who have the responsibility and
authority to implement the ALARA program.
7The radiation work permit (RWP) system assists in man-rem tracking and maintaining radiation exposure of personnel ALARA.
The RWP system is described in the Radiation Protection Plan and
may include the following: 1. Designation of services to be performed on specific components, equipment, or systems. 2. Anticipated radiation, airborne radioactive material, and contamination levels based on current surveys of the work areas and date of survey. 3. Monitoring requirements such as continuous air monitoring, sampling equipment, and radiation surveys.
RBS USAR Revision 20 12.1-11 4. Requirements for adherence to specific radiological precautions, procedures, or practices and special instructions and equipment to minimize exposures to
radiation and contamination. 5. Protective clothing and equipment requirements. 96. Personnel dosimetry requirements including D L R s, or direct-reading dosimeters, or alarming dosimeters, as appropriate.
- 97. Estimated exposure time required to complete the tasks and estimated doses anticipated from the exposure. 8. Identification of personnel working on the tasks. 9. Authorization to perform the tasks. 10. Actual exposure time, doses, and other information obtained during the operation to assist with post-job dose evaluation and personnel job and dose tracking.
This information can be used to identify deficiencies of problems and may provide the basis for revising procedures, and modifying features or other adjustments that may reduce the man-rems during subsequent similar
operations.82 Operations, maintenance, refueling, inservice inspections, and radwaste system operating procedures are reviewed for compliance with ALARA guidelines outlined in Regulatory Guide 8.8, the Radiation Protection Plan, and Section 12 during initial preparation; subsequent revision will be monitored through the
pre-job ALARA review and/or information gained from actual post-
job evaluations.
2 8Information gained from actual job exposure evaluations is incorporated into the training programs for the radiation worker, radiation protection technician, and supervisory personnel. Exposure reviews are performed at least annually by the ALARA Committee and plant management and are used to identify groups
with the highest exposure in order to assure that doses are ALARA. Upper management can compare the incremental reduction of
doses with the cost of station modifications that could be made.
RBSUSARRevision7 12.1-12January1995TheradiationworkpermitsystemisimplementedtohelpmaintainradiationexposureofpersonnelALARAby:1.Requiringspecificradiologicalprecautions,procedures,or practices.2.Providingprotectiveclothing,equipment,anddosimetryrequirementsforpersonnel.3.Settingdoselimitsasrequired.4.Providingformonitoringrequirementsasneeded.5.Providingarecordofdosesreceivedbypersonneltocompleteatask.6.Providinginformationthatcanbeusedinapost-operationreviewofthetasktoidentifydeficienciesorradiologicalproblems.Thatreviewmayprovidethebasesforrevising procedures,modifyingfeatures,ormakingotheradjustments thatmayreduceexposuresduringsubsequentsimilar
operations.2 2Inadditiontotheprecedingefforts,allemployeeareencouragedtosubmitsuggestionsonmethodsofreducingexposureof personneland/orimprovingtheALARAprogram.7Toensurethatpersonnelcomplywithestablishedradiologicalpolicies,proceduresandpractices,radiationprotection managementpersonnelarechargedwiththeresponsibilityto promptlyadvisehighermanagementofanyunsafepracticeswhich exceedtheirauthoritytocorrect.Theyhavetheauthorityto haltanyoperationwhichintheirjudgmentisunsafe.Radiation ProtectionTechniciansareresponsiblefornotifyingthe operationsshiftsupervisionorradiationprotectionmanagement immediatelyinordertostopworkonanyoperationdeemedtobe radiologicallyunsafe.
7Theradiationprotectiontraining,RadiationProtectionPlan,radiationworkpermitsystemandproceduresreviews RBS USAR Revision 20 12.1-13 all help to ensure that radiation exposure of personnel is maintained ALARA. There are also several work practices or techniques that are used to maintain personnel exposure ALARA.
Some of the ALARA techniques include the following: 971. Personnel required to be monitored for radiation exposure in accordance with 10CFR20.1502 are assigned D LR s to establish exposure history and direct-reading dosimeters to allow determination of accumulated exposure at any time during the job. Dose rate meters are also used, as needed, to provide early evaluation of possible doses to
individuals in specific areas.
7 92. Work involving high dose rates, high levels of removable contamination, or high levels of airborne radioactivity are preplanned. The purpose of preplanning is to prepare for the job carefully so it can be performed in a proper
and safe manner with a minimum exposure of personnel. 3. On jobs involving high dose rates (> 100 mrem/hr), high levels of removable contamination, or high levels of
airborne radioactivity, radiation protection coverage is
provided as required. 4. On complex jobs or jobs with exceptionally high radiation levels, job briefings and/or dry-run training is utilized and, in some cases, mockups are used to familiarize the workers with the operations that they are to perform.
These techniques are beneficial to improving worker efficiency and minimizing the amount of time spent in the radiation field. These efforts are documented, and the
experience is used to improve future efforts. 5. On jobs where general area radiation levels are unusually high, stay times are established. This ensures that
personnel do not exceed applicable dose limits. 6. On jobs that are known or expected to accumulate significant exposures, the job preplanning includes
estimates of the man-rem needed to complete the job. At the completion of the work, a debriefing session is held with the personnel who performed the work (when practical) in an effort to determine how the work could
be completed more efficiently, resulting in less
accumulated exposure. This RBSUSARRevision7 12.1-14January1995information,togetherwiththeproceduresused,actualman-remaccumulated,andpertinentradiologicalinformation, thatmayincluderadiologicalsurveys,contamination reports,ALARAconcerns,etc.,iscompiledandfiledfor referencetoprovideguidanceduringpreplanningofsimilar worksituations.Thisincorporatesexperiencegainedin performingthesetasksintofutureworkefforts.7.Aspractical,entryandexitpointsforworkareasareestablishedinlow-levelrdiationareas.Thisisdoneto minimizedoseaccumulatedwhilechangingprotectiveclothing andrespiratoryequipment.Accesspointsarealso establishedtominimizethespreadofremovable contaminationfromthejobsite.8.Asmuchaspracticable,jobsareperformedoutsideradiationareas.Thisincludesitemssuchasreadinginstruction manualsormaintenanceprocedures,adjustingtoolsorjigs, repairingvalveinternals,andprefabricatingcomponents.9.Individualsworkinginradiologicallycontrolledareasaretrainedtobeawareofthevaryingintensitiesofradiation fieldswithinthegeneralvicinityoftheirjoblocations andinstructedtoremainintheareasoflowerradiation levelsasmuchaspossible,consistentwithperformingtheir assignedtasks.Forcertainhighexposurejobs,maps, postings,and/ordetailedinstructionsareprovidedto clearlydelineatethesourceofradiationortoalert personnelconcerningthelocationofhotspotsandgenerally higherdoserateareas.Providedwiththisinformationand adheringtogoodradiologicalworkpractices,workerswill becognizantoftheirimmediateradiologicalenvironment, thusmaintainingALARAexposures.710.Protectiveclothingandrespiratoryequipmentprescribedbyradiationprotectionarecommensuratewiththeradiologicalhazardsinvolved,andsuchrequirementscannotbedecreased byanyotherpersonnel.Considerationisgiventothe discomfortofworkerstominimizetheeffectofprotective effortsonefficiencyandthetimespentinaradiation
area.7 RBSUSAR 12.1-15August198711.Contaminationcontainments,i.e.,glovebags,plasticbottles,tents,etc.,andspecialventilationsystemsareusedwherepracticablewhenpersonnelareworkingonhighly contaminatedequipment.12.Onsomejobs,specialtoolsorjigsareusedwhentheirusepermitsthejobtobeperformedmoreefficientlyorprevents errors,thusreducingthetimespentinaradiationarea.
Specialtoolsarealsousediftheiruseincreasesthe distancefromthesourcetotheworker,therebyreducingthe exposurereceived.13.Forlong-termrepairjobs,considerationisgiventosettingupacommunicationsnetwork,suchassoundpowered telephonesorclosedcircuittelevision,toassist supervisingpersonnelincheckingonworkprogressfroma lowerradiationarea.14.Whenpractical,systemsandotherpiecesofequipmentwhicharesubjecttocrudbuildup,suchasreactorwatercleanup systemresidualheatremovalsystem,liquidradwastesystem, andvariouspumps,filters,anddemineralizers,areequipped withconnectionswhichcanbeusedforflushingthesystem toeliminatepotentialhot-spotbuildup.Priortoperformingmaintenancework,considerationisgiventoflushingand/orchemicallydecontaminatingthesystemor pieceofequipmentinordertoreducethecrudlevelsand henceexposureofpersonnel.15.Permanentshieldingisused,wherepossible,toreduceradiationexposureattheworksiteandindesignated "waitingareas"topersonnelduringperiodswhentheyare notactivelyinvolvedinthework.Onsomejobs,temporary shielding,suchasleadsheetsdrapedorstrappedovera pipeorconcreteblocksstackedaroundapieceofequipment, isused.Temporaryshieldingisusedonlyifthetotal exposure,whichincludesexposurereceivedduring installationandremoval,isreduced.Experiencewithsuch operationsisusedindevelopingguidelinesinthisarea.
RBS USAR Revision 23 12.1-16 15 Operation of the hydrogen water chemistry system results in several changes to the station chemistry balance with impacts to
- 1. Injection of excess free hydrogen into reactor feedwater shifts the stoichiometric oxygen concentration in the
reactor vessel. The lower oxygen levels can result in free nitrogen combining with the excess hydrogen resulting in generation of volatile NH 3 and NH 4 compounds versus less volatile NOH compounds. These compounds are
removed from the vessel in the main steam and result in increased main steam line dose rates from the additional
nitrogen 16. The magnitude of this N-16 shift to steam
is based on the feedwater hydrogen concentration.
- 2. Switching to a hydrogen water chemistry environment from a normal water chemistry environment changes the chemical environment for the BWR from an oxidizing regime to a reducing regime. As such, the iron-based oxide films
existing in the primary system are gradually transformed
from a hematite structure to a substituted spinel structure. This conversion process increases the
transport of insoluble corrosion products, which subsequently deposit on the recirculation system piping
and component surfaces, causing increased dose rates and exposures. The intensity of this process appears to be dependant on the amount of hematite-type iron in the plant, the amount of hydrogen injected, and the operation of the hydrogen injection system. The injection of depleted zinc oxide (DZO) into the feedwater is a means of controlling Co-60 transport by retaining Co-60 on the fuel surfaces in a tenacious spinel form. However, high crud loading from the feedwater input will lead to higher dose rates and increase the likelihood of hot spots.
Laboratory studies and BWR HWC operating data has shown
that shutdown does rates can be 25-50% higher with frequent cycling of the hydrogen system compared to
steady-state moderate HWC.
15 RBSUSAR 12.2-1August198712.2RADIATIONSOURCES12.2.1ContainedSources 12.2.1.1General Withtheexceptionofthevesselanddrywellshields,shieldingdesignsarebasedonfissionproductandactivationproduct sourcesconsistentwithSection11.1.Forshielding,itis conservativetodesignforfissionproductsourcesatpeakvalues ratherthananannualaverage,eventhoughexperiencesupportsa lowerannualaveragethanthedesignaverage.Itshouldbenoted thatactivationproducts,principallyN-16,controlshielding calculationsinmostoftheprimarysystem.Inareaswhere fissionproductsaresignificant,conservativeallowanceismade fortransitdecay,whileatthesametimeprovidingfortransient increaseofthenoblegassource,daughterproductformation,and energylevelofemission.Areaswherefissionproductsare significantrelativetoN-16include:thecondenseroffgas systemdownstreamofthesteamjetairejector,liquidandsolid radwasteequipment,portionsofthereactorwatercleanupsystem, andportionsofthefeedwatersystemdownstreamofthehotwell includingcondensatetreatmentequipment.Forapplication,thedesignsourcesaregroupedfirstbylocationandthenbyequipmenttype(e.g.,containment,coresources).
Thefollowingparagraphsrepresentthesourcedatainvarious piecesofequipmentthroughouttheplant.Generallocationsof equipmentareshowninthegeneralplantarrangementdrawingsof Section1.2.12.2.1.2Containment 12.2.1.2.1ReactorVesselSources 12.2.1.2.1.1General Theinformationinthissectiondefinesareactorvesselmodelandtheassociatedgammaandneutronradiationsources.This sectionisdesignedtoprovidethedatarequiredforcalculations beyondthevessel.Thedataselectedwerenotchosenforany givenprogram,butwerechosentoprovideinformationforanyof severalshieldprogramtypes.Inadditiontothesourcedata, calculatedradiationdoselevelsareprovidedatlocations surroundingthevessel.Thesedataaregivenaspotential checkpointsforcalculationsbyshielddesigners.
RBSUSARRevision8 12.2-2August199612.2.1.2.1.2PhysicalDataTable12.2-1presentsthephysicaldatarequiredtoformthemodelonFig.12.2-1.Thismodelwasselectedtocontainasfew separateregionsaspossibletoportrayadequatelythereactor.
Table12.2-1providesnominaldimensionsandmaterialvolume fractionsforeachboundaryandregioninthereactormodel.To describethereactorcore,Table12.2-1providesthermalpower, powerdensity,coredimensions,coreaveragematerialvolume fractions,andreactorpowerdistributions.Thereactorpower distributionsaregivenforbothradialandaxialdistributions.
Thisdatacontainsuncertaintiesinthevolumeregionsnearthe edgeofthecore.Thelevelofuncertaintiesfortheseregions isestimatedat20percent.12.2.1.2.1.3CoreBoundaryNeutronFluxes Table12.2-2presentspeakaxialneutronmultigroupfluxesatthecoreequivalentradius.Thecoreequivalentradiusisa hypotheticalboundaryenclosinganareaequaltotheareaofthe fuelbundlesandthecoolantspacebetweenthem.Thepeakaxial fluxoccursadjacenttotheportionofthecorewiththegreatest power.AsshownbythedatainTable12.2-1,thispointisbelow thecoremidplane.Sincethedataarecalculatedwithacore equivalentradius,thefluxrepresentsameanfluxinthe azimuthalanglearoundthecore.Whilethefluxwithinanygiven energygroupisnotknownwithinafactorof2,thetotal calculatedcoreboundaryfluxisestimatedtobewithin+50 percent.12.2.1.2.1.4GammaRaySourceEnergySpectra CoreSpectraTables12.2-3athrough12.2-3dpresentaveragegammarayenergyspectraperwattofreactorpowerinbothcoreandnoncore regions.InTable12.2-3atheenergyspectrainthecoreare presented.Theenergyspectrainthecorerepresenttheaverage gammarayenergyreleasedbyenergygroupperwattofthecore thermalpower.TheenergyspectrainMeVpersecperwattcanbe usedwiththetotalcorepowerandpowerdistributionstoobtain thesourceinanypartofthecore.Thegammarayenergyspectraincludethefissiongammarays,thefissionproductgammaray,andthegammaraysresultingfrom inelasticneutronscatteringandthermalneutroncapture.The totalgammarayenergyreleasedinthecoreisestimatedtobe accuratetowithin+10percent.Theenergyreleaserateabove6MeVmaybeinerrorbyasmuchasafactorof+2.
RBSUSARRevision8 12.2-3August1996PostOperationGammaRayEnergySpectraTable12.2-3bgivesagammarayenergyspectruminMeVpersecperwattinspentfuelasafunctionoftimeafteroperation.
Thedatawerepreparedfromtablesoffissionproductdecaygamma fittedtointegralmeasurementsforoperationtimesof10 8secorapproximately3.2yr.Toobtainshutdownsourcesinthecore, thegammarayenergyspectraarecombinedwiththecorethermal powerandpowerdistributions.Shutdownsourcesinasinglefuel elementcanbeobtainedbyusingthegammarayenergyspectraand thethermalpowertheelementcontainedduringoperation.NoncoreGammaRayEnergySpectraInTable12.2-3c,thegammarayenergyspectrainthecylindricalregionsofthereactorfromthecorethroughthevesselare given.TheenergyspectrumisgivenintermsofMeVpercm 3 persec-wattattheinsidesurfaceandoutsidesurfacesofthe region.Thisenergyspectrummultipliedbythecorethermal poweristhegammaraysource.Thepointontheinsidesurface oftheregionisthemaximumpointwiththeregion.Inthe radialdirection,thevariationinsourceintensitymaybe approximatedbyanexponentialfittothedataontheinsideand outsidesurfacesoftheregion.Theaxialvariationinaregion canbeestimatedbyusingthecoreaxialvariation.The uncertaintyinthegammarayenergyspectraisdueprimarilyto theuncertaintyintheneutronfluxintheseregions.The uncertaintyintheneutronfluxisestimatedtovaryfrom approximately+50percentatthecoreboundarytoafactorof+
3attheoutsideofthevessel.Thecalculationswerecarriedout withvoidsbeyondthevessel.Thepresenceofshieldmaterials beyondthevesselcausesanincreaseinthegammasourceonthe outsideofthevessel.12.2.1.2.1.5GammaRayandNeutronFluxesOutsideVessel Table12.2-4presentsthemaximumaxialneutronandgammarayfluxesoutsidethevessel.Themaximumaxialfluxoccursonthe vesseloppositetheportionofthecorewiththemaximumpower level.ThiselevationcanbelocatedusingthedatafromTable 12.2-1.Thefluxesatthiselevationarebasedonameanradius coreanddonotshowazimuthanglevariations.12.2.1.2.1.6GammaRayDoseRatesandFastNeutronFluxesatthe VesselTable12.2-5presentsthecalculatedfastneutronfluxeswithenergygreaterthan1MeVandgammaraydoseratesinR/hrat threepointsoutsideofthereactorvessel.
RBSUSARRevision8 12.2-4August199612.2.1.2.2RadioactiveSourcesintheReactorWater,Steam,andOffGasTheradioactivesourcesinthereactorwater,steam,andoffgasarediscussedinChapter11.Thismaterialprovidesthe concentrationsduringnormaloperationoftheradioisotopes leavingthereactorvessel.12.2.1.2.3ReactorWaterCleanupSystem Theradioactivesourcesinthereactorwatercleanup(RWCU)systemaretheresultoftheactivityinthereactorwaterin transitthroughthesystemoraccumulationofradioisotopes removedfromthewater.Componentsforthissysteminclude regenerativeandnonregenerativeheatexchangers,pumps,valves, filterdemineralizers,andthebackwashreceivingtank.Most majorcomponentsofthecleanupsystemarelocatedinthe containment;however,theRWCUprocesspumpsarelocatedinthe auxiliarybuilding.ThesystemisdescribedinSection5.4.8.
ThegammasourcesforthecleanupsystemaregiveninTable 12.2-6.Thesourcesshownforthefiltersandreceivingtankare presentduringallmodesofoperation.Therefore,backwashing capabilityisprovidedtoremovetheresidualactivityfor effectiveradwastehandling.12.2.1.2.4MainSteamSystem Allradioactivematerialsinthemainsteamsystemresultfromradioactivesourcescarriedoverfromthereactorduringplant operation.Inmostofthecomponentscarryingmainsteam,the sourceisdominatedbyN-16.IncomponentswheretheN-16has decayed,theotheractivitiescarriedbythesteambecome
significant.12.2.1.2.5RadiationSourcesInTheTransversingIncoreProbe SystemTheradiationsourceforthetransversingincoreprobe(TIP)systemisprovidedinTables12.2-13through12.2-16.The radiationsourceisbaseduponlocationwithinthecoreand residencetime.Asindicatedinthetables,theTIPsystemis dividedintothreecomponentsforshieldingcalculations:
fissionablematerial,nonfissionablematerial,andcable.
Sourcesareprovidedforeachcomponentfor100-secirradiation andzero-secdecay.Thereactorstartupsourceisshippedtothesiteinaspecialcaskdesignedforshielding.Thesourceistransferred underwaterwhileinthecaskandloadedintoBeryliumcontainers.
Thisisthenloadedintothereactorwhileremainingunderwater.
Thesourceremainswithinthereactorforitslifetime.Thus,no uniqueshieldingrequirementsafterreactoroperationare
required.
RBSUSARRevision8 12.2-5August199612.2.1.3AuxiliaryBuildingThebasicsourcesintheengineeredsafeguardsystemsduringnormalplantoperationaretheresultoftheradioactive materialsinthereactorwaterorsteambeingtransportedtothe systems.Thedesignbasissourcetermsforthisequipmentassume thetotalactivityistheconcentrationofreactorwaterorsteam decayedfortheappropriatetimeinterval.Thedesigngammasourcestrengthsintheresidualheatremoval(RHR)systemwereevaluatedforthesystemoperatinginthe reactorshutdownmode.Inthismode,thesystemrecirculates reactorcoolanttoremovereactordecayheat.Thesystemis operatedfromapproximately4hraftershutdownuntiltheendof theshutdownperiod.ThesourceintheRHRsystemisthe activityinthevolumeofreactorwatercontainedinthesystem.
ThesourcestrengthsareshowninTable12.2-7.Thedesigngammasourcestrengthsinthereactorcoreisolationcooling(RCIC)systemwereevaluatedforthesystemoperating undertestconditions.Duringroutinetestingofthesystem,the sourceintheequipmentistheactivityofthesteamdrivingthe systemturbine.ThisactivityisdominatedbyN-16.The radiationsourcedatausedintheshielddesignofthissystem areshowninTable12.2-7.12.2.1.4FuelBuilding 12.2.1.4.1RadioactiveSourcesintheSpentFuel TheradiationsourceforspentfuelisgiveninSection12.2.1.2.1.4,Post-Operation(Table12.2-3b),intermsof MeV/sec-watt.Thedesigncalculationiscarriedoutforamean elementatvarioustimesaftershutdown.12.2.1.4.2FuelPoolCoolingandCleanupSystem Thedominantspentfuelpoolliquidsourceisthatwhichresultsfromcrudwhichmaybeshakenloosefromthefuelrodswhenthey aremovedinthepool.MeasurementsatNineMilePointIhave beenpresentedinapaperentitled"CorrosionProductDepositson FuelattheNineMilePointBoilingWaterReactor"byJ.Blok, S.G.Sawochka,andD.T.SnyderofGeneralElectricCompanyatthe 19thAnnualMeetingofANS,June1973.Thesemeasurementshave beenusedtoestimatethesourcesexpectedatRiverBendStation.
Theexpectedsourcestrengthsforthefuelpoolheatexchanger, thecleanupfilter,andthecleanupdemineralizerarelistedin Table12.2-8.
RBSUSARRevision8 12.2-6August199612.2.1.5TurbineBuilding12.2.1.5.1TurbineSystem PipingandequipmentwhichcontainsteamaresourcesofradiationduetothepresenceofN-16,thepredominantsourceofactivity duringoperation.Fissionproductgases(xenonandkrypton),
wateractivationproducts(O-19andN-17),andthecarryoverof iodineandotherfissionproductspresentinsteamandcondensate areconsideredasadditionalactivitysources.Thecarryoveris conservativelyassumedtobe1.5percentbyweightforhalogens and0.1percentbyweightforotherfissionproducts.TheN-16 concentrationsinequipmentarelistedinTable12.2-9.Credit fortransitdecayfortheshort-livedN-16hasbeenfactoredinto thesourceterms.12.2.1.5.2CondensateandFeedwaterSystems Thesourcesinthecondensatesystemarebasedondecayedmainsteamactivities.Almostallnoncondensablesenteringthecondenserareremovedbytheairejectors.Becauseofthisandtherelativelylongholdup time,theN-16andothergaseousactivityisverylowinthe hotwellandnegligibleintheremainderofthecondensate system.Theactivitiescausedbyactivatedcorrosionandfission productsareshowninTable12.2-10forthecondensateand feedwatersystems.12.2.1.5.3OffGasSystem Theoffgassystemcharcoaldelaybedslocatedintheturbinebuildingcontainsourcesofradiationbasedontheholdupof noncondensablegasesfromthemaincondenserbythemain condenserairremovalsystem.TheN-16activityintheair removalsystemisshowninTable12.2-9.TheN-16activityin theoffgassystem(downstreamoftheholduppipe)isnegligible duetodecay.Therefore,thepredominantradiationsourcesare thefissionproductgases,xenon,andkrypton,andtheirdaughter products.Thesourcetermsusedforshieldingaregivenin Tables12.2-11aand12.2-11b.12.2.1.6RadwasteBuilding Theradwastesystemsourcesareradioisotopes,includingfissionandactivationproducts,presentinthereactorcoolant.The componentsoftheradwastesystemscontainvaryingdegreesof activitydependingonthedetailedsystemandequipmentdesign.Theradionuclidesourcesintheliquidradwastesystemsuchaspipes,tanks,filters,demineralizers,andevaporatorsusedin shieldingcalculationsarelistedinTable12.2-12.Information onthesolidradwastesystemisprovidedinSection11.4.
RBS USAR Revision 24 12.2-7 8 12.2.1.7 Turbine Rotor Storage Enclosures The two turbine rotor storage enclosures are located onsite approximately 800 ft southwest of the Turbine Building. The sources are fixed low level radioactive beta and gamma contamination contained on the two decontaminated stored turbine
rotors and associated diaphragms. The contact dose rate for the stored
components is a maximum of 3 millirem per hour. 8 12.2.1.8 Low Level Radwaste Storage Facility The radiation protection design of the LLRWSF, in terms of shielding and dose estimates is based upon the dose rates of the waste containers in the facility when the facility is filled to capacity.
Table 11.4-5 includes the approximate waste volumes by type over a four year period. The design basis radiation levels for the wastes to be stored in the facility are based on River Bend Station's historical source terms.
The radiation shielding configuration of the LLRWSF is designed in accordance with the dose rate criteria per 10 CFR 20, USAR and site procedures for both the site boundary and for restricted areas. The nearest site boundary to the LLRWSF is greater than 1350 ft. away. The facility is designed so that the maximum offsite dose received at the site boundary from the waste stored in the LLRWSF will be maintained less than or equal to 1 mRem/yr (see Table 12.4-2) and that the onsite dose due to temporary storage will be maintained within 10 CFR 20. The restricted area is the area surrounding the LLRWSF which will be controlled by the Radiation Protection Group for purposes of protection of individuals from exposure to radiation and radioactive materials. The facility is designed so that the maximum dose when the facility is fully loaded at fence line is less than 0.6 mrem per hour.
The projected waste containers curie content is listed in Table 11.4-5.The location of the LLRWSF is shown in Figure 1.2-2.
12.2.1.9 Independent Spent Fuel Storage Installation (ISFSI)
The Holtec International's HI-STORM system radiation protection design, in terms of shielding, is based upon dose rates calculated using the computer code MCNP4A, which is a state of the art Monte Carlo code that offers coupled neutron-gamma transport using continuous energy cross sections in a full three-dimensional geometry. A full analysis was completed and can be referenced with Holtec's Dose to Distance (D2D) document, HI-2043196. As stated in Chapter 9.1.2.5, the ISFSI is located within the protected area of the plant and was designed to contain 40 casks. As stated in D2D, the maximum dose rate at the site boundary would be 3.39E-05 mRem/hr. This dose rate will maintain a less than 1mRem/yr to site boundary and is within 10 CFR 20.Other documents to reference with regard to the ISFSI are theHoltec Certificate of Compliance (CoC), document HI-20443196, Holtec Final Safety Analysis Report (FSAR), NRC Safety Evaluation Report (SER) and the Entergy Nuclear South 10 CFR 72.212 Evaluation Report.
RBS USAR Revision 24 12.2-7a 12.2.2 Airborne Radioactive Material Sources
Design efforts are directed toward keeping all radioactive material contained. Leaks from process systems, refueling, and decontamination may lead to airborne radioactivity. Equipment cubicles, corridors, and areas routinely occupied by operating personnel do not contain significant airborne radioactivity sources. Radioactive equipment which has the potential for leakage is installed in separate shielded compartments not routinely occupied. In general, the direction of airflow within
the building ventilation systems is from areas of low potential for airborne contamination to areas of increasing potential. In this manner, routinely occupied areas are maintained at low levels of airborne radioactivity. Data from operating BWRs corroborate the general lack of airborne activity in corridors and routinely occupied operating areas (2). Air samples and surface contamination swipe samples are performed to verify the
absence of airborne and surface contamination. 8 12.2.2.1 Airborne Sources During Normal Operation
The expected airborne radioactivity concentrations in various plant areas for normal plant operation are shown in Table 12.2-17
and are based on the data given in NUREG-0016 (1)¸ and EPRI-495 (2). The ventilation system design is evaluated in relation to the
location of expected airborne sources as identified in References 1 and 2 to determine the distribution of the radioactivity throughout the plant. These radioactivity concentrations result in the expected annual releases specified in Table 11.3-1, after credit is taken for filtration of radioiodine and particulates. However, determination of the
airborne concentrations within the containment is based on plant specific analyses that account for the cleanup provided by the
containment purge system.
8 The source of airborne radioactivity within the containment during normal plant operation is derived from the assumed discharge of 2,000 pounds per hour of main steam from the safety
relief valves to the suppression pool as well as the drywell leakage of radioactivity evolved from the sumps within the drywell. Cleanup of the containment atmosphere is provided by
the 7,000 cfm exhaust from the containment purge system.
RBSUSARRevision8 12.2-8August1996Corridorsandroutineaccessoperatingareaswithintheradwastebuildingarenotexpectedtohavesignificantairborne radioactivitylevels.Equipmentcubiclesareinfrequently accessedandmaycontainlowlevelsofairborneradioactivity, butdesignprovisionsareprovidedtominimizethereleaseof
radioactivity.Radwastebuildingtanksarefilledfromthetopandasthewatersplashesintothetanks,dissolvedandentrainedradioactivity maybecomeairborne.Thisactivityisnotreleasedintothe atmosphereintheroomsbecausethetankventsareconnected directlytothebuildingventilationsystem.Pumpsandvalves forradioactivesystemsintheradwastebuildingarelocatedin separatecompartmentsthatarenotnormallyoccupied.The radwastebuildingventilationdesignprovidesairflowfromareas oflowpotentialforairbornecontaminationtoareasof increasingpotential.Thisinsuresthatanyleakagefrom radwastepumpsandvalvesisnotdirectedintonormallyoccupied areasofthebuilding,butisexhaustedfromthebuilding.An estimateoftheairbornelevelsexpectedwithinradwastebuilding pumproomsisprovidedinTable12.2-17.Thisisbasedona distributionoftheairborneradioactivitywithinradwaste buildingareasasspecifiedinReferences1and2andrepresents amaximumleveltobeexpectedwithinradwastebuildingequipment
cubicles.Themainpotentialsourceofairborneradioactivitywithintheturbinebuildingisleakagefromvalvesonlargelinescarrying high-pressuresteam.TheRiverBendStationdesignprovidesfor collectionofthisleakageanditstransportbacktothe condenser.Therefore,noblegasairborneconcentrationsare expectedtobenegligiblethroughouttheturbinebuildingexcept forwithinthesteamjetairejector(SJAE)cubicles.These areasnormallyarenotoccupiedduringoperation,andtheexhaust fromthesecubiclesisexhaustedtotheenvironmentafter filtrationtoeliminatethepossibilityofcontaminationof adjoiningareas.Airborneactivitiesintheauxiliarybuildingareexpectedtobelowexceptforwithinthereactorwatercleanup(RWCU)pump cubicle.Thiscubicleisnotnormallyoccupiedduetoradiation levels.However,theexhaustfromthiscubicleisexhaustedto theenvironmentduringnormaloperationandthroughthestandby gastreatmentsystempostaccident.Radioactiveequipmentinthefuelbuildingwhichhasthepotentialforleakageisinstalledinseparateshielded compartmentswhicharenotnormallyoccupiedduetoradiation levels.Theventilationinthefuelbuildingdirectstheair fromthesecubiclestotheenvironmentsothatanairborne probleminthenormallyoccupiedareasisprevented.The ventilationsystemisalsodesignedtosweepairfromthespent fuelpoolsurface,therebyremovingthemajorportionof potentialairbornecontamination.Inaddition,evaporationfrom thespentfuelpoolisminimizedbycoolingofthepool.
RBSUSARRevision8 12.2-9August199612.2.2.2AirborneSourcesDuringRefuelingExperienceatoperatingBWRshasshownthatairborneradioactivitycanresultfromthereactorvesseldryerand separatoriftheirsurfacesareallowedtodry.Otherpotential airbornesourcescouldoccurduringvesselheadventingandfuel movement.Theairborneradioactivematerialsourcesresulting fromreactorvesselheadremovalareminimizedbyventingprior toremovaleithertothedrywellpurgeexhaustsystemortothe maincondenser,withvacuumsuppliedbythemechanicalvacuum pump.Thecontributiontotheairborneradioactivityduetothe reactorvesselinternalsisminimizedbykeepingthemwetor submerged.Expectedairborneconcentrationswithinthe containmentduringrefuelingarepresentedinTable12.2-17and arebasedondatagiveninNUREG-0016.12.2.2.3AirborneSourcesforReliefValveVenting Aspecialconsiderationforairborneconcentrationswithinthecontainmentistheradioactivityreleaseviareliefvalve dischargetothesuppressionpool.Thedifferentmodesofthis typeofdischargeandthefrequencyofoccurrencearediscussed inChapter15.Thelimitingcaseisthemainsteamisolation valveclosure.Thiseventresultsinthereleaseof834,000lbm ofmainsteamtothesuppressionpooloveraperiodof approximatelythreehours.Usingtheexpectedsourcetermsfor mainsteamfromTable11.1-1andaccountingforthepartitioning oftheradioactivitybetweentheairandwater,theairborne radioactivityconcentrationswithinthecontainmentare calculatedasshowninTable12.2-18.
RBSUSARRevision8 12.2-10August1996References-12.2
- 1. CalculationofReleasesofRadioactiveMaterialsinGaseousandLiquidEffluentsfromBoilingWaterReactors(BWR-Gale Code),NUREG-0016,Rev.1,January1979.
- 2. SourcesofRadioiodineatBoilerWaterReactors,EPRI NP-495,ResearchProject274-1,FinalReport,February1978.
RBS USAR 12.3-1 August 1987 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 Facility Design Features
In this section, specific design features to maintain personnel exposures ALARA are discussed. An assessment of compliance of River Bend Station with Regulatory Guide 8.8 is provided in Section 1.8. Applicable discussions can be found in
Sections 11.1 and 12.3 through 12.5.
12.3.1.1 Plant Design Description The equipment and plant design features employed to maintain radiation exposures ALARA are based upon the design considerations of Section 12.1.2 and are outlined in this section for several general classes of equipment (Section 12.3.1.1.1) and
several typical layout situations (Section 12.3.1.1.2).
12.3.1.1.1 Common Equipment and Component Designs
This section describes the design features utilized for several general classes of equipment or components. These classes of
equipment are common to many of the plant systems; thus, the
features employed for each system to maintain minimum exposures
are similar and are discussed by equipment class in the following
paragraphs.
FiltersFilters which accumulate radioactivity carried in liquids are supplied with the means to remotely backflush the filter or
discharge the filter cake.
DemineralizersDemineralizers for radioactive systems are designed so that spent resins can be remotely and hydraulically transferred to the radwaste system for solidification. Fresh resin can be remotely loaded into the demineralizer. Underdrains and downstream strainers are designed for full system pressure drop. The
demineralizers and piping are designed with provisions to flush
with condensate. Combination vent overflow lines prevent entry
of spent resin into the exhaust duct.
EvaporatorsEvaporators are provided with chemical addition connections through normal process piping for the use of chemicals for RBS USAR 12.3-2 August 1987 descaling operations. Space is provided to allow uncomplicated removal of heating tube bundles. The more radioactive components
are separated from those that are less radioactive by a shield wall. Wherever practicable, instruments and controls are located
on the accessible side of the shield wall. Wherever practicable, valves in radioactive lines are located on the accessible side of
the shield wall.
Pumps Where practicable, pumps are purchased with mechanical seals to reduce seal-servicing time. Pumps and associated piping are
arranged to provide adequate spaces for access to the pumps for
servicing. Small pumps are installed in a manner which allows easy removal, if necessary. All pumps in radioactive waste systems are provided with flanged connections for ease in removal. Pump casings are provided with drain connections for
draining the pump for maintenance.
Tanks Whenever practicable, tanks are provided with sloped or conical bottoms and bottom outlet connections. Overflow lines are
directed to the floor and equipment drain system in order to
control any contamination within plant structures. Provisions such as compartmentation of tanks and compartment floor drains
are made to contain overflows and accidental spills.
InstrumentsInstrument devices are located in low-radiation zones and away from radiation sources whenever practical. Primary instrument devices which, for functional reasons, are located in high-radiation zones are designed for easy removal to a lower radiation zone for calibration, if possible. Transmitters and readout devices, whenever practical, are located for servicing in low-radiation zones, such as corridors and the main control room.
Some instruments (such as thermocouples) in high-radiation zones
are provided in duplicate to reduce the access and service time
required.Instrument sensing lines on process piping which may contain highly radioactive liquids and solids are provided with remote backflushing capability to reduce the servicing time required to
keep the lines free of solids. Instrument and sensing line
connections are located in such a way as to RBS USAR Revision 22 12.3-3 avoid corrosion product and radioactive gas buildup. Compression-type fittings are used between instruments and their
process lines to allow for easy removal.
Valves To minimize personnel exposures from valve operations, motor-operated or other remotely actuated valves are used to the maximum extent practicable. Solenoids for air-operated valves in radwaste systems are located on remote racks in low-radiation
zones.
Valves are located in valve galleries so that they are shielded separately from major components. Long runs of exposed piping
are minimized in valve galleries. In areas where manual valves are used on frequently operated process lines, either valve stem extenders or shielding is provided so that personnel need not
enter the radiation area for valve operation.
When equipment in Zone VI is operated infrequently, only those manual valves associated with safe operation, shutdown, and draining of the equipment are provided with remote-manual operators or reach rods. All other valve operations are
performed with equipment in the shutdown mode. Simple reach rods are used to allow operators to retain the feel of whether the
valves are tightly closed or not.
For valves located in radiation areas, provisions are made to
drain adjacent radioactive components when maintenance is
required.
Wherever practicable, valves for clean, nonradioactive systems are separated from radioactive sources and are located in readily
accessible areas.
All manually operated valves in the filter and demineralizer valve compartments required for normal operation and shutdown are remotely operated. Personnel do not enter the valve gallery during flushing operations. The valve gallery shield walls are designed for maximum expected filter backflush activities during
flushing operations.
Full-ported valves are used in systems expected to contain radioactive solids.
RBS USAR Revision 6 12.3-4 August 1993 Valve designs with minimum internal crevices are used where crud trapping could become a problem, especially for piping carrying
spent resin or evaporator bottoms. Valves are flanged in
radwaste systems for ease of removal from piping systems.
PipingThe piping in pipe chases is designed for the lifetime of the unit. There are no valves or instrumentation in the pipe chase.
Wherever radioactive piping is routed through areas where routine maintenance is required, pipe chases are provided to reduce the radiation contribution from these pipes to levels appropriate for
the inspection requirements. Wherever practicable, piping containing radioactive material is routed to minimize radiation
exposure to personnel. 6The majority of the carbon steel Reactor Water Cleanup piping located in the Drywell has been replaced with stainless steel piping to decrease future radioactive crud buildup on the
interior surface. In addition, the piping has been electropolished and preoxidized to further enhance the corrosion resistance of the material. Also, the existing drain "Dead
Legs", which were crud traps and significant dose contributors, were removed. Snubber reduction was incorporated into this modification to reduce the number of maintenance man-hours spent
in the Drywell. These modifications were made to help reduce the area dose rates in the Drywell in accordance with the River Bend
Source Term Reduction program.
6Floor Drains Floor drains and properly sloped floors are provided in each room or cubicle containing serviceable components containing radioactive liquids. When practicable, shielded pipe chases are
used for both major and field-run radioactive pipes. If a
radioactive drain line must pass through a zone lower than that
at which it terminates, proper shielding is provided. Local gas
traps or porous seals are not used on radwaste floor drains.
LightingMultiple electric lights are used to provide a sufficient margin of illumination for each cell or room containing highly
radioactive components so that entry is not required each time a single lamp fails. Incandescent lights, which require less time for servicing than fluorescent lights, are used wherever possible. The fluorescent lights which are RBS USAR Revision 6 12.3-4a August 1993 used in some areas do not require frequent service due to the increased life of the tubes. In order to further minimize the exposure of personnel, all burned-out lamps are normally replaced
when the system is secured and in the flushed-out condition.
HVACThe HVAC system design provides for rapid replacement of filter elements and housings.
RBS USAR Revision 6 12.3-4b August 1993 THIS PAGE LEFT INTENTIONALLY BLANK RBS USAR Revision 20 12.3-5 Sample StationsSample stations for routine sampling of process fluids are located in accessible areas. Shielding is provided at the local sample stations as required to maintain radiation levels. The counting room and laboratory facilities are described in Section
12.5.Clean ServicesWhenever practicable, clean services and equipment such as compressed air piping, clean water piping, ventilation ducts, and
cable trays are not routed through radioactive pipechases.
12.3.1.1.2 Common Facility and Layout Designs This section describes the design features utilized for River Bend Station processes and layout arrangements. These features are employed in conjunction with the general equipment designs described in Section 12.3.1.1.1 and are discussed in detail in
the following paragraphs.
Valve GalleriesValve galleries are provided with labyrinths for personnel protection. Floor drains are provided to control radioactive leakage. To facilitate decontamination in valve galleries, concrete floors are covered with a smooth surface coating which
allows easy decontamination.
PipingPipes carrying radioactive materials are routed through controlledareas properly zoned for that level of activity. Each piping run is individually analyzed to determine the potential
radioactivity level and surface dose rate. Where it is necessary that radioactive piping be routed through corridors or other low-radiation zone areas, appropriate shielding is provided.
Whenever practicable, valves and instruments are not placed in
radioactive pipeways. Whenever practicable, equipment compartments are used as pipeways only for those pipes associated
with equipment in that compartment. When possible and practicable, radioactive and nonradioactive piping are separated to minimize personnel exposure. Should maintenance be required, provision is made to isolate and drain
radioactive piping and associated equipment.
RBS USAR 12.3-6 August 1987 Potentially radioactive piping is always located in appropriately zoned and restricted areas. Process piping is monitored to
ensure that access is controlled to limit exposure (Section
12.5).Piping is designed to minimize low points and dead legs. Drains are provided on piping where low points and dead legs cannot be
eliminated. Thermal expansion loops are raised rather than dropped, where possible. In radioactive systems, the use of nonremovable backing rings in the pipe weld joints is prohibited to eliminate a potential crud trap for radioactive materials.
Piping carrying resin slurries or evaporator bottoms is run vertically whenever possible, and large radius bends are utilized
instead of elbows. Whenever possible, branch lines having little or no flow during normal operation are connected above the horizontal midplane of
the main pipe.
PenetrationsTo minimize radiation streaming through penetrations, as many penetrations as practicable are located with an offset between
the source and the accessible areas. If offsets are not practicable, penetrations are located as far as possible above
the floor elevation to reduce the exposure to personnel. If these two methods are not used, then alternate means are
employed, such as baffle shield walls or grouting the area around
the penetration.
Contamination ControlAccess control and traffic patterns are considered in basic plant layout to minimize the spread of contamination. Equipment vents and drains from highly radioactive systems are piped directly to the collection system instead of allowing any contaminated fluid to flow across to the floor drain. All-welded piping systems are employed on contaminated systems to the maximum extent practicable to reduce system leakage and crud buildup at joints.
The valves in some radioactive systems are provided with leakoff
connections piped directly to the floor and equipment drains.
Decontamination of potentially contaminated areas within the plant is facilitated by the application of suitable smooth
surface coatings to the concrete floors and walls. Floor drains with properly sloping floors are provided in all potentially contaminated areas of the plant. In RBS USAR Revision 20 12.3-7 addition, radioactive and potentially radioactive drains are separated from nonradioactive drains. In uncontrolled access areas where contamination is expected, radiation monitoring equipment is provided (Section 12.3.4).
Those systems which become highly radioactive, such as radwaste slurry piping systems, are provided with flush and drain connections. Certain systems have provisions for chemical and
mechanical cleaning prior to maintenance. Concrete shield wall thicknesses are dimensioned for the auxiliary, turbine, fuel, reactor, and radwaste buildings on Fig. 12.3-6 through 12.3-9. These figures also show radiologically controlled areas, personnel and equipment decontamination areas, contamination control areas, location of the health physics facilities, location of area radiation
monitors, location of control panels for radwaste equipment and components, location of the onsite laboratory, and location of
the counting room.
Equipment LayoutIn those systems where process equipment is a major radiation source (e.g., fuel pool cleanup, radwaste, condensate demineralizer), pumps, valves, and instruments are separated from the process component. This allows servicing and maintenance of these items in reduced radiation zones. Control panels are
located in low-radiation areas. Major components such as tanks, demineralizers, and filters in radioactive systems are isolated in individual shielded
compartments insofar as practicable.
Provision is made on some major plant components for removal of these components to lower radiation zones for maintenance. Labyrinth entranceway shields or shielding doors are provided for each compartment from which radiation could stream to access
areas and exceed the radiation zone dose limits for those areas. Exposure from routine inplant inspection is controlled by locating, whenever possible, inspection points in properly shielded low background radiation areas. Radioactive and nonradioactive systems are separated as far as practicable to
limit radiation exposure from routine inspection of nonradioactive systems. For radioactive systems, emphasis RBS USAR 12.3-8 August 1987 is placed on adequate space and ease of motion in a properly shielded inspection area. Where longer times for routine inspection are required, and permanent shielding is not feasible, sufficient space for portable shielding is provided. In high-radiation areas where routine inspection is required, remote
viewing devices are provided as needed. When this is not
practicable, written procedures which reduce radiation exposure
by reducing the total time exposed to the radiation field are used, and access to high-radiation areas is under direct
supervision of radiation protection personnel.
Corrosion Product Control Control and minimization of the buildup, transport, and deposition of activated corrosion products in the reactor coolant and auxiliary systems are important to maintaining the occupational radiation exposure ALARA. Studies at operating BWR
plants have shown that activated corrosion products (primarily Co60) are the most significant contributors to occupational radiation exposures. Several design features are utilized at River Bend Station to minimize the buildup and control the
transportation of corrosion products in the reactor coolant and related auxiliary systems. The condensate demineralizers and the reactor water cleanup filter/demineralizers are designed to maintain the required reactor water quality by the removal of corrosion products from the condensate/feedwater and from the reactor recirculation water, respectively. The selection of
system components (valves and pumps) and the layout of the system piping are designed to minimize crud trap locations, such as
instrument taps off the pipe bottoms. Major components and pipe sections in reactor coolant systems can be isolated and flushed
with high quality water to decontaminate the corrosion products built up on internal surfaces prior to maintenance or surveillance operations. Material selection for piping and components in contact with the reactor coolant and feedwater is
based on maintaining the amount of cobalt content to a minimum and on minimizing the material erosion and corrosion rates.
Reactor vessel internals, feedwater heater tubes, and reactor
recirculation piping are all stainless steel material that will not significantly contribute to the activated corrosion products in the reactor coolant. System layup practices and startup operating procedures are established to flush out reactor coolant related auxiliary systems and feedwater system prior to operation
so that a minimum amount of corrosion products will enter the
reactor water.
RBS USAR Revision 20 12.3-9 12.3.1.2 Radiation Zoning and Access Control Access to areas inside the plant structures and plant yards is regulated and controlled by radiation zoning and access control (Section 12.5.3.3). Each radiation zone defines the radiation level range to which the aggregate of all contributing sources
must be attenuated by shielding. All plant areas are categorized into radiation zones according to expected radiation levels and anticipated personnel occupancy
with consideration given toward maintaining personnel exposures
as low as is reasonably achievable and within the standards of
10CFR20. Each room, corridor, and pipeway of every plant building is evaluated for potential radiation sources during normal operation and shutdown; for maintenance occupancy requirements; for general access requirements; and for material exposure limits to determine appropriate zoning. Specific zoning for each plant area is shown on Fig. 12.3-1 through 12.3-5b. All frequently accessed areas (e.g., corridors) are shielded for Zone
II or III access. Whenever practicable, the measured radiation level and the location of the source are posted at the entry to any radiation
or high-radiation area.
The control of ingress or egress of plant operating personnel to radiologicallycontrolled areas and procedures employed to assure that radiation levels and allowable working time are within the
limits prescribed by 10CFR20 are described in Section 12.5. 7Any area having a radiation level which could cause a deep dose equivalent exposure in any 1 hr in excess of 5 mrem is conspicuously posted with a sign or signs bearing the radiation
symbol and the words "CAUTION - RADIATION AREA." In lieu of the control device or alarm signal required by paragraph 20.20.1601 of 10 CFR Part 20, any area in which the intensity of radiation is such that a major portion of the body could receive in any 1 hr a deep dose equivalent greater than 7
RBS USAR Revision 7 12.3-10 January 1995 100 mrem* but less than 1,000 mrem shall be barricaded and conspicuously posted as a high radiation area, and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)**. Any individual or group of individuals permitted to enter such areas shall be provided with or
accompanied by one or more of the following:
1.A radiation monitoring device which continuously indicates
the radiation dose rate in the area.
2.A radiation monitoring device which continuously integrates
the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas
with this monitoring device may be made after the dose rate levels in the area have been established and personnel have
been made knowledgeable of them. 7 3.An individual qualified in radiation protection procedures, with a radiation dose rate monitoring device, who is
responsible for providing positive control over the activities within the area, and who shall perform periodic radiation surveillance at the frequency specified on the
RWP by radiation protection. In addition to the requirements for high radiation areas, accessible areas with radiation levels such that a major portion of the body could receive in 1 hr a deep dose equivalent greater than 1,000 mrem shall be posted at each entrance with a sign bearing the radiation symbol and the words "CAUTION" or "DANGER" and "LOCKED HIGH RADIATION AREA" and provided with locked doors to prevent unauthorized entry. The keys shall be maintained under the administrative control of the Operations Supervision on duty and/or radiation protection. Doors shall remain locked except during periods of access under an approved RWP that specifies the dose rate levels in the immediate work area. For
accessible areas that are located within large areas, such as the
containment, where
- Measurement made 12 in from the source of radioactivity. ** Radiation protection personnel or personnel escorted by radiation protection personnel shall be exempt from the RWP
issuance requirement during performance of their assigned duties, provided they are otherwise following plant radiation
protection procedures for entry into high radiation areas.
7 RBS USAR Revision 16 12.3-11 March 2003 7 no enclosure exists for purposes of locking and no enclosure can be reasonably constructed around the individual areas, and within which radiation levels are such that a major portion of the body could receive in 1 hr a deep dose equivalent in excess of 1,000 mrem*, then that area shall be roped off and conspicuously
posted, and a flashing light shall be activated as a warning
device. In lieu of the stay-time specification of the RWP, continuous surveillance, direct or remote (such as use of closed-circuit TV cameras), may be made by personnel qualified in
radiation protection procedures to provide positive exposure
control over the activities within the area.
716Special design features are provided to control access to areas adjacent to the fuel transfer tube to prohibit inadvertant exposure of personnel during fuel transfer. These areas include
the containment fuel transfer isolation valve room, the annulus area, and the fuel building midsupport room, areas shown on Fig. 12.3-9 and 12.3-11. A system of administrative controls plus door interlocks and annunciators in the main control room
assures that fuel transfer is not conducted when these areas are
occupied. The entrance to these areas and the removable access plugs are clearly marked with signs stating that potentially lethal radiation fields are possible during fuel transfer. The removable plugs are locked. These areas are inaccessible once the fuel transfer system is placed in an operational mode.
Permanent shielding in these areas is designed to assure that the shield contact radiation levels are no greater than 100 rads per hour. Refer to Figures 12.3-1 through 12.3-9 for the calculated
radiation levels.
1612.3.2 Shielding 12.3.2.1 Design Objectives The objective of the shielding is to protect the operating personnel and the general public from radiation emanating from the reactor, power-conversion process, and auxiliary systems, including equipment and piping. Shielding requirements in the
plant are determined so as to perform the following functions: 7
- Measurement made 12 in from the source of radioactivity.
7 RBS USAR Revision 22 12.3-12 1. Ensure that exposure to radiation of plant personnel, contractors, and visitors is ALARA and within 10CFR20 limits.
- 2. Limit exposure to radiation of plant personnel in the main control room to within the limits of 10CFR50.67, to ensure that the plant can be maintained in a safe
condition under accident conditions. Amendment 132 revised the design basis accident main control room dose
limit requirements to incorporate the limits of 10CFR50.67. The limits of 10CFR50 Appendix A
, also remain applicable to the RBS design basis.
- 3. Limit exposure to radiation of certain components with specified radiation tolerances if they are in
high-radiation areas.
- 4. Limit exposure to persons at the restricted area boundary to a small fraction of 10CFR20 as a result of direct
radiation during normal operation.
12.3.2.2 Design Description
12.3.2.2.1 General Design Guides
In order to meet the design objectives, the following general design guides are used in the shielding analysis of the plant.
An assessment of compliance with Regulatory Guides 1.69 and 8.8
in the design of shielding is provided in Section 1.8.
All systems containing radioactivity are identified and shielded, based on the access requirements of adjacent areas. The radiation zone designation for each area and the amounts of shielding needed to ensure these zones are determined. Radiation
zones are shown on Fig. 12.3-1 through 12.3-5.
An effort is made to locate process systems in such a manner as to minimize shielding. Use of labyrinths is made in order to eliminate any streaming radiation from equipment. Segregation of radioactive equipment is provided, wherever practicable, such that shielding reduces the dose rate from adjacent components to
less than 20 mrem/hr.
Penetrations are placed so that they do not pass through the shield wall in a direct line with the radiation source in order to prevent streaming. If this is not feasible, adequate
shielding is provided.
Wherever possible, radioactive piping runs in such a manner as to minimize radiation exposure to plant personnel. This involves
minimizing radioactive pipe routing in corridors,
RBS USAR 12.3-13 August 1987 avoiding the running of high-activity pipes through low-radiation zones, using shielded pipe chases where routing of high-activity pipes in low-level areas cannot be avoided, and separating
radioactive and nonradioactive pipes for maintenance purposes. To maintain acceptable levels at valve stations, motor-operated or full-ported valves are used where practical. For valve maintenance, provision is made for drainage of associated equipment so that radiation exposure is minimized. If manual
valves are used, provision is made to shield the operator, if
practicable, from the valve by the use of temporary shielding or
valve stem extensions. Provisions are made to shield major sources during inservice inspection to allow sufficient access. The principal shield material is concrete of density 135 lb/cu ft. Alternatively, water, steel, high-density grout, or lead is used as needed.
12.3.2.2.2 Plant Shielding Description Plant building layouts, which provide locations of equipment containing radioactive fluids and indicate shield wall thickness, radiation zone designations, access control, and radiation
monitor locations, are shown on Fig. 12.3-6 through 12.3-9. The general description of plant shielding in the different plant buildings is as follows.
Reactor Building Shielding for the reactor building includes the primary shield, drywell, and shield building walls. The primary shield wall surrounds the reactor and reduces gamma heating in the drywell concrete wall; reduces activation of, and radiation effects on, materials and equipment in the drywell; and provides limited access in the drywell for shutdown periodic inspection and maintenance. The drywell is Zone VI during normal
operation.The drywell wall provides additional shielding in order to permit access to the containment during normal operation. Within the containment, there are several shielded rooms that enclose reactor water cleanup system equipment and RBS USAR Revision 16 12.3-14 March 2003 piping, and safeguard and process equipment. In addition, the main steam lines are within a shielded tunnel. The shield building wall is a 2 1/2-ft-thick concrete structure that completely surrounds the nuclear steam supply system. This wall attenuates the system radiation to ensure that levels outside the building are less than 0.2 mrem/hr. In addition, in the unlikely event of an accident, the shield building wall shields personnel and the public from radiation sources inside
the containment. Details on the construction of these walls are presented in Section 3.8. RBS provides shielding designs for the fuel transfer tube and canal commensurate with the guidance of Regulatory Guide 8.8 that
results in radiation doses within the limits of 10CFR20. 16Access to areas where contact with the fuel transfer tube may occur is administratively controlled. Further protection against
inadvertent personnel exposures is assured through system
interlocks that prevent fuel transfer tube operation when these accessible areas are unsecured. Signs are posted stating that potentially lethal radiation fields are possible during fuel
transfer.16Access to the fuel transfer tube area within the reactor building is through a 2 1/2 x 2 1/2 ft opening that is located in the
annulus and through an entrance to the isolation valve room in
the containment. The access opening in the annulus is plugged with a 5 1/2-ft-thick, solid concrete block to maintain the
radiation levels in the area within the zoning requirements. The seismic shake spaces between the steel containment and the structures inside and outside the containment are shielded by steel plates. Fig. 12.3-11 shows the reactor building areas
through which the spent fule transfer tube passes.
RBS radiation zone maps identify the maximum expected total radiation levels during operations and refueling with consideration given toward maintaining personnel exposure as low as is reasonably achievable and within the standards of 10CFR20.
The zone maps show all the routinely visited areas such as reactor water cleanup (Figure 12.3-4, el 162'-3"), standby liquid control areas (Figure 12.3-3, el 141-0"), TIP Station (Figure 12.3-2, el 98'-0"), CRD hydraulic control unit (Figure 12.3-4, el 114'-0"), and containment RBS USAR Revision 14 12.3-15 September 2001 personnel airlocks (Figure 12.3-4, el 114'-0" and el 162'- 3"). These zone radiation levels include contributions from any
potential streaming through the drywell shield wall penetration.
Turbine BuildingThe anticipated major radiation source in the turbine building is the primary steam containing activation gases, principally N-16, and fission products. Radiation shielding is provided around the following equipment in order to ensure that the required access
zone radiation limits are met around the shielded areas: 1. Main condenser 2. Feedwater heaters and drain receiver tanks
- 3. Steam air ejectors
- 4. Steam extraction piping
- 5. Off gas equipment and piping
- 6. Condensate demineralizers and regeneration facilities
- 7. Steam seal evaporator
- 8. Turbines
- 9. Radwaste steam reboiler
- 10. Moisture separator/heaters. Areas within these shields are high-radiation zones and have limited access.
Dose rates to areas north of hydrogen analyzers and off gas sample rack cubicles located at el 123'-6" of the off gas area are estimated to be less than 100 mRem/hr during normal plant operation. Dose rates are gradually decreased to less than 5 mrem/hr to areas north of the glycol cooler/tank/pumps and west
of the local panels. A detailed radiation zone description is
provided on Fig. 12.3-3. 14In the vicinity of the west side of the condenser there are localized areas where the dose rate is estimated to be greater
than 100 mrem/hr.
14 RBS USAR Revision 22 12.3-16 14 If during operation the dose rate is found to be greater than 100 mrem/hr, then additional shielding will be installed or
administrative access control will be implemented.
14 Auxiliary and Radwaste Buildings Concrete walls, removable blocks, labyrinths, and pipe chases are used to shield the safety feature equipment in the auxiliary building and process equipment in the radwaste building, including valves and piping, in accordance with general guides of Section 12.3.2.2.1. Radiation limits are specified for all access zones, and shielding is provided as necessary to maintain the
zones within these limits.
Fuel Building
The shielding in the fuel building consists mainly of water and
concrete. The water levels are maintained to keep maximum dose rates to 2 mrem/hr when the fuel elements are in the spent fuel
racks, as well as from fuel pool cooling and cleanup equipment
such as the demineralizers and filters. 16 Access to the fuel transfer tube area within the fuel building is through an opening that is 3 ft-0 in by 3 ft-6 in. This opening is blocked with removable solid concrete that is 3 ft-0 in thick
to maintain the radiation levels in the area within the zoning
requirements. Administrative procedures, which include a locked gate, are employed to control this access. Fig. 12.3-11 shows the fuel building areas through which the spent fuel transfer
tube passes.
16 Main Control Room The main control room shielding design is based on the requirements of 10CFR50.67, which requires that personnel can
occupy and have access to the main control room following a
hypothetical maximum accident, maintain full control, and shut
down the plant. Amendment 132 revised the design basis accident main control room dose limit requirements to incorporate the limits of 10CFR50.67.
The limits of 10CFR50 Appendix A, General Design Criteria 19
, also remain applicable to the RBS design basis.
Direct shielding of the main control room from the fission product inventory in the containment is provided by the concrete
walls between them. Emergency air-conditioning and filtration systems are provided for post-LOCA conditions in the main control
room, as described in Section 9.4.1.
RBS USAR Revision 14 12.3-17 September 2001 General Plant Yard AccessAreas adjacent to the off gas treatment facilities, turbine, building, reactor building, and radwaste building that are accessible on an unlimited basis receive less than 0.2 mrem/hr.
Protection for these areas is afforded by concrete building walls and by access-controlling fences. Unlimited access areas adjacent
to the unshielded condensate storage tanks are protected by fencing and by controlling the level of radioactivity in the stored condensate by passing the stored water through either the
condensate demineralizers or the liquid radwaste system before
storage.8Unlimited access areas next to the two stored onsite turbine rotors and diaphragms located approximately 800 feet southwest of the Turbine Building are protected by two modular enclosures that
surround the two turbine rotors and diaphragms.
812.3.2.3 Method of Shielding Design 14The shielding approach and methods are consistent with those described in RP-8A (1). Where shielded entry ways to compartments containing high-radiation sources are necessary, labyrinths are designed using a general purpose Monte Carlo radiation transport
code. The shielding provides a total dose rate contribution at the entry way to below the upper limit of the radiation zone
specified for the area.
1412.3.2.4 Post-Accident Access to Vital Areas NUREG-0737 Item II.B.2 requires that following a design basis accident (DBA) there exist adequate protection of safety-related equipment and the capability for safe and adequate access to vital plant areas consistent with the dose guidelines of GDC 19.
At RBS these requirements are met by providing sufficient shielding of components containing post-accident sources.
Protection of safety-related equipment is addressed in Section 3.11. This section addresses the capability of safely and adequately reaching and working in designated vital areas following a DBA at RBS. Assessment of this capability is based
upon the following considerations: 1. NUREG-0737 source term assumptions. 2. Identification of vital areas and routes to and from these areas. 3. Earliest time post-accident when access to each of these areas is required. 4. Occupancy time required in each of these areas.
RBS USAR Revision 17 12.3-18 5. Walking speeds and driving times along identified access/egress routes. 6. Hoisting times required for moving heavy grab samples. Based upon the above, dose rates along the designated access/egress routes and at manned locations within the designated vital areas are calculated in order to ensure that anticipated doses to essential personnel from post-accident radiation sources meet the GDC-19 limits of 5 rem whole body, or
its equivalent to any part of the body. River Bend has implemented Alternate Source Term (AST), per Regulatory Guide 1.183, which revised the design basis source term from TID-14844 to NUREG-1465. As allowed per Regulatory Guide 1.183, a scoping/sensitivity evaluation demonstrated that the current vital area access doses using NUREG-0737 bound those using Regulatory Guide 1.183.
12.3.2.4.1 Liquid and Gaseous Source Terms The post-accident doses within the designated vital areas and enroute to and from these vital areas are determined based upon
liquid and gaseous source terms defined in NUREG-0737 as follows: 1. Containment/Drywell Gaseous Source Terms - 100 per-cent of the core equilibrium noble gas inventory and 50 percent of the core equilibrium iodine activity are
assumed to be released from the reactor vessel following a DBA LOCA. Of this release, 100 percent of the core noble gases occupy the drywell and containment volumes.
Of the 50 percent of the core halogens released, 50 percent plates out in the drywell; the remaining 25 percent of core halogens are distributed in the
drywell and containment atmospheres. 100 percent of the core noble gases and 25 percent of the core halogens are available for release from
containment following a DBA LOCA. This source term is the basis for cloud shine doses, ventilation intake doses, shine from ventilation filtration systems, auxiliary building airborne doses, containment shine, and post-accident sampling system (PASS) gaseous piping
doses. For evaluation of the shielding design and handling of grab samples for plant effluents, an additional source term of 100 uCi/cc of gaseous radioiodines and particulates is assumed for buildup on the filters, in accordance with Regulatory Guide 1.97 and NUREG-0737
Item II.F.1. Source Terms for Liquid-Containing Systems - The liquid source term is assumed to consist of RBS USAR Revision 19 12.3-19 50 percent of the core equilibrium halogen inventory and 1 percent of the remaining core fission products for
systems recirculating post-LOCA suppression pool fluids.For evaluation of the shielding design and handling of PASS liquid samples, an additional source term of 100 percent core noble gases, 50 percent core iodines, and 1 percent remaining core fission products mixed in the reactor pressure vessel coolant volume is assumed.
This source term is considered to demonstrate compliance
with the Regulatory Guide 1.97 and NUREG-0737 Item II.B.3
requirement for sampling capabilities at 10 Ci/cc.12.3.2.4.2 Identification of Vital Areas and Access/Egress RoutesThe following onsite areas/facilities are identified as requiring access following an accident in order for operators to perform
vital accident mitigation and/or recovery tasks: 71. Operational Support Center (OSC) - The OSC is a staging area for emergency response personnel. The OSC is located in the services building on el 1 23 ft. Should the OSC become uninhabitable during any portion of the
post-accident period, then the Control Room will serve as backup to the OSC. The location of the OSC is shown on
Fig. 12.3-15, Sheet 1.
- 72. Main Plant Ventilation Exhaust Effluent Monitor Grab Sample Area - Located on the auxiliary building roof (el 171-ft 0 in) near the personnel hatch (as shown on Fig. 12.3-15, Sheet 4), this area is considered vital in order to obtain post-accident grab samples of main plant
ventilation exhaust particulate and radioiodine
effluents.3. Post-Accident Sample Station (PASS) Control Panel and Sample Panel - Located on the east side of the auxiliary building on el 114 ft 0-in (as shown on Fig. 12.3-16, Sheet 2, and 12.3-17), this area is considered vital in order to obtain reactor coolant and containment gas samples for radiological and chemical analyses following
a LOCA.
RBS USAR Revision 22 12.3-204. Chemistry Laboratory - This facility (shown onFig. 12.3-15, Sheet 7) is located on el 109 ft-3 in of the services building. Following an accident it is the principal laboratory where grab samples are taken to be
analyzed. This area is served by a filtered ventilation
system. In the event that post-accident radiation
levels become restrictive in this laboratory due to large radioactivereleases and/or unfavorable meteorology, then the Environmental Services Group Laboratory will serve as a backup facility.115. Primary Access Point - The primary access point (PAP),located on the north side of the services building, is the security center for RBS and serves as the facility
through which personnel enter and exit the protected
area. Should post-accident radiation levels in this
facility become prohibitively large, by which time all
nonessential personnel will have been evacuated from the
site in accordance with the emergency response plan, then the security supervisor will lock all access gates to the protected area. The guard force will move from the PAP to the emergency operations facility (EOF) in the training building and maintain control of unauthorized access to the site. A backup security terminal for protected area access and egress and for
personnel accountability is located in the TSC.
116. Main Control Room - Following an accident, the maincontrol room (located on el 136 ft-1 5/8 in of the control building) serves as the primary emergency
response facility. As such, this facility is designed for continuous occupancy. Post-LOCA habitability of the main control room is discussed in Section 6.4, and
post-LOCA doses in this facility are presented in
Table 15.6-7.7. Technical Support Center (TSC) - Following an accident, the TSC serves as a backup emergency response facility to the main control room. The TSC is designed for
continuous occupancy to serve the needs of emergency
operations management and support. The TSC is located on el 123 ft-6 in of the services building and is shown
on Fig. 12.3-15, Sheet 1, and Fig. 12.3-16, Sheet 1.It has been determined that the post-accident access to areas addressed in NUREG-0737, Item II.B.2, which have not RBS USAR 12.3-21 August 1987 been identified above are not required at RBS. The plant design is such that the accident mitigation functions associated with these additional areas can be performed from the main control
room.Following an accident, essential support plant personnel will be assembled in the OSC and dispatched at appropriate times to the designated vital areas. Since the PAP and the main control room are continuously occupied during normal operations, it is reasonable to expect that essential personnel will be present in
these areas when the accident occurs. The main plant exhaust
duct effluent monitor grab sample area and the PASS room are vital areas for which access and egress routes have been determined. The egress routes from these two areas lead to the chemistry laboratory in the services building before terminating
at the OSC. Obtaining a post-LOCA iodine grab sample at the main plant exhaust duct monitor is postulated in the RBS vital area dose analysis to be a two-phase effort in order to minimize the dose to any one individual. Phase I personnel start from the OSC and
travel to the grab sample area to disconnect the shielded grab sample (weighing approximately 275 lb) and mount it onto a cart.
From there, the grab sample is moved by cart to the west side of the auxiliary building roof, where it is left for Phase II personnel to transport to the chemistry laboratory. Phase I
personnel proceed down the west stair tower to a waiting vehicle, drive to the services building, and return to the OSC. Phase II personnel start from the OSC and travel by truck to the west side of the auxiliary building, up the west stair tower, and over to the waiting cart and grab sample. Phase II personnel rig the shielded grab sample to a jib crane, lower the grab sample onto the waiting truck, walk down the stairs to the truck, and then drive to the services building. Phase II personnel then
transport the grab sample by cart to the chemistry laboratory and finally return to the OSC. The details of these scenarios are presented in Tables 12.3-5 and 12.3-6. Fig. 12.3-15, Sheets 1
through 7, illustrates the routes traveled.
Retrieval of a shielded liquid PASS sample is assumed to be performed in a single phase. The sample retrieval scenario is outlined in Table 12.3-7, and the pathways are illustrated on
Fig. 12.3-16, Sheets 1 through 3.
12.3.2.4.3 Major Dose Contributors in the Vital Areas To determine post-accident doses to personnel going to and working in the previously identified vital areas, the RBS USAR 12.3-22 August 1987 following radiation contributors are considered in the RBS vital areas analysis: 1. Auxiliary Building a. ECCS piping b. Containment shine (including shine through personnel hatch at el 170 ft) c. Building airborne activity
- d. Overhead cloud shine on auxiliary building roof
- e. Reactor vessel liquid and containment airborne sources in PASS room piping and in the grab samples f. Effluent iodine collected on main plant exhaust duct effluent monitor grab sample filters. 2. Services Building - TSC a. Containment shine
- b. Overhead cloud shine
- c. Intake of containment releases
- d. Intake of ESF leakage
- e. TSC intake filter shine. 3. Services Building - Chemistry Laboratory a. Containment shine
- b. Overhead cloud shine
- c. Intake of containment releases
- d. Intake of ESF leakage
- e. Radiation from grab samples during analyses. 4. OSC, Corridors, and General Areas a. Airborne cloud doses RBS USAR Revision 10 12.3-23 April 1998 b. Radiation from grab samples along the pathways to the chemistry laboratory c. Containment shine.
12.3.2.4.4 Vital Areas Access Design Review Results Results of the RBS vital area access design review, together with major assumptions, are presented in Tables 12.3-5, 12.3-6, 12.3-7, 12.3-8, and 12.3-9. These results show the following: 1. No worker will be exposed to gamma doses in excess of GDC-19 limits if the iodine effluent grab sample retrieval operation is begun as early as 7 hr post-LOCA
and this operation is carried out in two phases. 2. No worker will be exposed to gamma doses in excess of GDC-19 limits while retrieving PASS samples if this
effort is initiated as early as 1 hr post-LOCA. 3. The TSC design meets the radiological habitability criteria in GDC-19, SRP 6.4, NUREG-0696, and NUREG-0737. 4. The gamma doses to a laboratory technician in the chemistry laboratory will meet the GDC-19 criteria as early as 1 hr post-LOCA. The shielded hot storage room and the unshielded hot laboratory may each be occupied
for up to 30 min at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> post-LOCA. 5. The use of a pressure demand self-contained breathing apparatus may be required in order for GDC-19 and SRP 6.4 thyroid dose limits to be met along access/egress routes and in vital areas not serviced by filtered ventilation
intakes.212.3.2.4.5 Plant Design Features Necessary to Implement Design Review Results 10 210 RBS USAR Revision 17 12.3-24 2The following design changes were implemented in order to support the vital area access scenarios: 1. A jib crane on the west side of the auxiliary building roof to lower the iodine sample onto a truck waiting
below.2. A moveable hoist above the el 123 ft-6 in floor in the east-west tunnel near the auxiliary building stair landing will facilitate the lifting of the shielded PASS sample and cart from el 114 ft-0 in. in the auxiliary
building to the east-west tunnel at el 123 ft-6 in. 3. Extension of the shield wall near the main plant exhaust duct radiation monitor. 4. Shielding between PASS system components.
212.3.3 Ventilation 12.3.3.1 Design Objectives The primary function of the plant ventilation systems is to provide suitable environmental conditions that are safe and comfortable for operating personnel and adequate for the functioning of equipment, and to provide effective protection for
operating personnel against possible airborne radioactive
contamination. The systems are designed to operate such that the in-plant airborne radioactivity levels for normal operation (including anticipated abnormal operational occurrences) for normally occupied areas within plant structures containing radioactive components are within the limits of 10CFR20, Appendix B, Table I. For other buildings, including all
administration areas, airborne radioactivity levels are maintained within the limits of 10CFR20, Appendix B, Table II.
The maximum values are not expected to be greater than maximum permissible concentration (MPC) during normal operation.
Section 12.2.2 discusses airborne radionuclide sources for the fuel building, the reactor building, turbine building, auxiliary building, and radwaste building. Exposures within the plant are
discussed in Section 12.4.1.
The main control room ventilation system is designed to provide a suitable environment for equipment and continuous personnel
occupancy in the main control room under all modes of plant operation in accordance with 10CFR50, Appendix A, Criterion 19
and 10CFR50.67
.
RBS USAR Revision 22 12.3-25 12.3.3.2 Design Guidelines
In order to limit and reduce the airborne radioactivity in accordance with the design objectives, the following general
design guidelines are employed to the maximum extent practicable.
Guidelines to Minimize Airborne Radioactivity
- 1. For radioactive systems, equipment vents and drains are piped directly to a collection device connected to the
collection system instead of allowing any contaminated
fluid to flow across the floor to a floor drain.
- 2. All-welded piping systems are employed on contaminated systems to the maximum extent practicable. Gasketed flanged connections are used for equipment where welded
connections are impractical.
- 3. The valves in systems containing radioactive fluids are provided with leakoff connections piped directly to the
condenser.
- 5. Contaminated equipment is designed to minimize the potential for airborne contamination during maintenance
operations. These features may include flush connections on pump casings for draining and flushing the pump prior to maintenance or flush connections on piping systems
that could become highly radioactive.
- 6. All sinks and chemical laboratory work areas where radioactive samples or material are handled are provided with exhaust hoods to protect operating personnel from
airborne contaminants.
- 7. Transient airborne contamination may result due to maintenance. Special procedures such as system
RBS USAR 12.3-26 August 1987 isolation and glove boxes are instituted to minimize the contamination on a case-by-case basis. 8. Filters in all systems are changed based upon the airflow and the pressure drop across the filter bank. In the
case of prefilters, a pressure drop equivalent to 1 in of
water across the bank is cause for changeout. The HEPA filters are changed when the pressure drop reaches 2 in
of water equivalent across filter bank. The charcoal adsorbers are changed based on the residual adsorption capacity of the bed as measured by test samples or
canisters removed and analyzed at intervals. 9. The majority of the accumulated radioactivity in the filtration units is removed by replacing the contaminated
filters.Guidelines to Control Airborne Radioactivity1. Air movement patterns are provided from areas of low potential radioactivity to areas of progressively higher
potential radioactivity prior to final exhaust. 2. Areas with radioactive contaminants are kept under negative pressure to minimize exfiltrationof contamination. Positive pressure is maintained in the main control room during normal operation and after
postulated accidents to minimize infiltration of
potential contaminants. 3. HEPA filters and charcoal filters are provided on exhausts from the fuel, auxiliary, containment (including drywell), and portions of radwaste and turbine buildings
to remove airborne radioactivity and to reduce onsite and
offsite radiation levels. The Engineered Safety Features (ESF) filter units for the fuel, auxiliary, containment, and control buildings are designed and constructed in
accordance with Regulatory Guide 1.52 with the exceptions
listed in Table 6.5-1. The charcoal filter units for the radwaste, turbine, and containment purge systems are designed and constructed in accordance with the requirements of Regulatory Guide 1.140 with the
exceptions listed in Table 9.4-6 RBS USAR 12.3-27 August 1987 4. The fresh air supply to the main control room is from remote air intake and is designed to be operable during loss of offsite power and LOCA. The air is filtered and passed through HEPA filters and charcoal adsorbers to prevent contamination of the control room by excessive
radioactivity.5. Containment isolation valves are installed in various systems in accordance with General Design Criteria 54 and 56, including valve controls, to assure that the
containment integrity is maintained (Section 6.2.4). 6. The main control room emergency filtration system, SGTS, and fuel building charcoal filtration systems are provided with redundant Seismic Category I equipment to
control the spread of airborne radioactivity. The extent to which redundant components are provided is discussed
in Sections 9.4 and 6.5. 7. Atmospheric tanks and condensate demineralizer area tanks which contain radioactive materials are exhausted to the atmosphere through charcoal filters of their respective
building ventilation system. 8. The primary drywell/containment purge system reduces airborne radioactivity within the drywell/containment to
acceptable levels prior to release to containment. A second system is provided to operate in the event of either mechanical failure of the normal operating system
or to purge the containment and drywell at a faster rate for personnel entry. The purge air is exhausted through the standby gas treatment system prior to exhaust to
remove airborne iodine and particulates.
Guidelines to Minimize Personnel Exposure from HVAC Equipment1. The ventilation fans and filters are provided with adequate access space to permit servicing with minimum personnel exposure (Fig. 12.3-12 through 12.3-14). The
HVAC system is designed to allow rapid replacement of
components.
RBS USAR 12.3-28 August 1987 2. Ventilation ducts are designed to minimize the buildup of radioactive contamination within the ducts to the extent practicable.3. Air is recirculated in clean areas only. 4. Access and service of ventilation systems in radioactively contaminated areas is expedited by component location to minimize personnel exposure during
maintenance, inspection, and testing as follows: a. The outside air supply units and building exhaust system components are located in ventilation equipment rooms and in general areas. These units are located in radiation Zone I or II, and are accessible during plant operation. Work space is
provided around each unit for anticipated
maintenance, testing, and inspection. b. Local cooling equipment, servicing the normal building requirements, is generally located in areas of low contamination potential, radiation Zones I, II, III, and IV. The drywell coolers are located in Zone VI. However, fully redundant components are
provided, which eliminate the need for servicing
during plant operation, thereby allowing servicing to be deferred until the next shutdown period.
Several turbine building unit coolers are located in Zone VI; however, these are expected to require a minimum of maintenance which can be scheduled during
shutdown.12.3.3.3 Design DescriptionsThe ventilation systems which have been designed in accordance with the guidelines in Section 12.3.3.2 are described below. The expected gaseous effluents for radiologically significant areas such as the auxiliary building, off gas building, radwaste building, fuel building, reactor building, and turbine building
are provided in Table 11.3-1.
RBS USAR Revision 17 12.3-29 12.3.3.3.1 Main Control Room Ventilation 16 During normal plant operation, main control room air is recirculated through an air-conditioning unit to maintain main control room design conditions of 80°F and 70 percent maximum relative humidity. Outdoor air for ventilation and
pressurization is normally provided from a local air intake located on the roof of the control building at el 161 ft-0 in. A radiation monitoring system and a control room charcoal filter recirculation system have been provided to detect and reduce radiation levels in the main control room in the event of a DBA.
Redundant radiation monitors in the local outdoor air intake duct
detect high radiation in the outdoor air supply. A high radiation signal alarm in the main control room automatically diverts the outdoor air to charcoal filtration units, shuts down and isolates the utility exhaust fan, and starts the emergency
charcoal filtration unit trains. Redundant radiation monitors are also provided in the remote air intake duct. The use of remote air intake is by operator action. When the remote air intake damper is opened, the normal intake isolation damper is closed manually and the air is passed through charcoal filtration
units to maintain a positive pressure. An area monitor located
in the main control room verifies that radiation levels are kept
within acceptable limits.
16There are two 100-percent capacity charcoal filter trains to provide clean, filtered air during DBA. The charcoal filters are
activated carbon which have a 98-percent efficiency for elemental iodine removal and 98-percent efficiency for organics at 25°C and 70-percent relative humidity. A complete description of main control room ventilation system and habitability is described in
Sections 9.4.1 and 6.4, respectively.
12.3.3.3.2 Fuel Building Ventilation14The fuel building is ventilated by an air-conditioning unit (with two redundant supply fans) and air distribution ductwork. During normal operation ventilation air is exhausted through an exhaust fan. During a fuel handling accident, the supply air system is shut off and the outside air is supplied through an intake duct.
The building air is exhausted through charcoal filtration units.
The exhaust system is designed to automatically route the exhaust air through charcoal filter units upon detection of high airborne radioactivity in the building exhaust by redundant radiation
monitoring systems.
14 RBS USAR 12.3-30 August 1987 Independent unit coolers are provided in various cubicles to maintain a design ambient temperature. The air is recirculated through chilled water cooling coils. A complete description of
fuel building ventilation is given in Section 9.4.2.
12.3.3.3.3 Containment The containment cooling system consists of recirculating air-conditioning units which maintain the design containment
temperature and relative humidity. During normal system operation, the containment coolers recirculate the air through chilled water cooling coils.
Vaneaxial fans are provided above the crane rail to avoid stagnation in the dome area and thus prevent accumulation of
radioactive gases. During normal operation, the containment purge system is used intermittently in order to maintain airborne concentrations in the containment less than 25 percent of maximum permissible concentrations as given in 10CFR20, paragraph 20.203.D(1)(ii) and Appendix B, Table 1, Column 1. The drywell is purged only during
periods when personnel entry is required. A backup system is
provided to operate in the event of either a mechanical failure
of the normal operating system or to purge the containment and
drywell at a faster rate for personnel entry. During a DBA the
containment and drywell purge system are shutdown.
A complete description of the containment ventilation system is found in Section 9.4.6.
12.3.3.3.4 Drywell The drywell cooling system consists of recirculation type unit coolers to maintain the design drywell temperature and relative
humidity. Six unit coolers are provided to distribute cooled air effectively through ductwork. The cooling water for the drywell
cooling system is provided by the plant service water system. The containment continuous purge system is operated by opening the isolation valves to the drywell. The containment purge system can be used for drywell purge, containment purge, or
recirculation cleanup. In any mode of operation, the activity of
the air is continuously monitored.
RBS USAR Revision 12 12.3-31 December 1999 A complete description of drywell ventilation is found in Section 9.4.7 and containment purge system in Section 9.4.9.
12.3.3.3.5 Annulus Pressure Control System 12The annulus pressure control system is designed to maintain a negative pressure of at least -3 in W.G. in the annulus space during normal plant operation. Two 100-percent capacity centrifugal exhaust fans are provided to exhaust air from the annulus to the plant exhaust duct. The suction of the two
exhaust fans is connected to a common duct which penetrates the
shield building to draw the annulus atmosphere.
12Redundant radiation monitors are provided in the exhaust duct and the air is continuously monitored for radioactivity prior to discharge to atmosphere. Following LOCA and/or detection of high
airborne radioactivity concentration, the annulus exhaust is diverted to the SGTS automatically. A pressure differential switch is provided to alarm the operator in the main control room
when the pressure exceeds -3 in W.G. A complete description of annulus pressure control system is found in section 9.4.6.
12.3.3.3.6 Auxiliary Building The auxiliary building ventilation system is designed to maintain design ambient air temperature and relative humidity. Outside
air is supplied to various areas of the building through ductwork
distribution system. The air is exhausted to the plant main
exhaust duct by an exhaust fan and ductwork. The exhaust air is continuously monitored for radioactivity. Upon detection of high airborne radioactivity, the normal supply
and exhaust fans are manually shut off and air is exhausted through the SGTS. In this mode, negative pressure is maintained within the auxiliary building to prevent outleakage of
radioactive air. Unit coolers are provided in various cubicles to maintain design ambient temperature. The auxiliary building electrical tunnel is ventilated by drawing outdoor air through the outdoor air intake
duct and discharging the air to the plant exhaust duct. A complete description of auxiliary building ventilation system is found in Section 9.4.3.
RBS USAR Revision 16 12.3-32 March 2003 12.3.3.3.7 Radwaste Building The radwaste building is ventilated by six unit coolers connected to a common intake plenum. Exhaust air from the building is
provided by two separate systems. One system serves the general
areas and equipment cubicles and the other exhausts air from the area in which tanks containing potentially radioactive fluids are located. The tanks with potentially high levels of radioactivity are vented through iodine filtration units. The radwaste
building exhaust is continually monitored for high radiation. A greater volumetric flow is exhausted from the radwaste building than is supplied to it to maintain a negative pressure. A complete description of the radwaste building ventilation system
is found in Section 9.4.3.
12.3.3.3.8 Turbine Building The turbine building ventilation air system delivers outdoor air to the areas of low airborne contamination of the turbine building. An exhaust air duct system is connected to each of the potentially contaminated cubicles. All the air is exhausted through contaminated cubicles thus maintaining an air flow pattern from areas of low potential radioactivity to areas of high potential radioactivity. In this way, clean area
passageways are kept free of radioactive contaminants.
The vault area in the off gas building is served by a low temperature recirculating refrigeration system. 16Radiation monitors are provided in the turbine building and off gas building exhaust ducts. Mechanical vacuum pump discharge is
drawn through an iodine filtration unit prior to discharge to the
plant exhaust duct.
16A complete description of the turbine building ventilation system is found in Section 9.4.4.
12.3.3.4 Air Cleaning System Description The air cleaning systems which utilize special filtration equipment to limit airborne radioactive contaminants are: 1. Standby gas treatment system (SGTS) - described in Section 6.5 RBS USAR 12.3-33 August 1987 2. Main control room charcoal filtration system - described in Sections 9.4.1 and 6.4 3. Fuel building charcoal filtration system - described in Section 9.4.2 4. Continuous containment purge system - described in Section 9.4.6 5. Radwaste building exhaust filtration system - described in Section 9.4.3 6. Mechanical vacuum pump system - described in Section 9.4.4. The guidance and recommendations of Regulatory Guide 1.52 concerning maintenance, in-place testing provisions for
atmospheric cleanup systems, and air filtration and adsorption units have been used as a reference in the design of the various safety-related charcoal filter systems. The extent to which Regulatory Guide 1.52 has been followed is discussed in Section 6.5. Nonsafety-related charcoal filters are designed to meet the requirements of Regulatory Guide 1.140 as discussed in
Section 9.4.3. Provisions specifically included to minimize personnel exposures and to facilitate maintenance or in-place testing operations are
as follows: 1. The filter adsorption of radioactive material during normal plant operation is a slow process; therefore, in
addition to monitoring for pressure drop, the filters are
checked for radioactivity buildup on a scheduled
maintenance basis with portable equipment, and the filter elements are replaced before the radioactivity level is of sufficient magnitude to create a personnel hazard.
Filter sections in which the radioactivity level (due to a postulated accident) is such that a change of the filter elements constitutes a personnel hazard are removed and replaced. Therefore, the filters can be maintained as necessary after a DBA, and it is not
necessary for workers to handle filter elements until the
short-lived isotopes have had sufficient time to decay to
minimize exposure during replacement. 2. Active elements of the atmospheric cleanup systems are designed to permit ready removal.
RBS USAR Revision 12 12.3-34 December 1999 83. The filter units are located in separate independent rooms. Adequate access to active elements is provided to simplify element handling. Ample space is provided in the filter rooms for accommodating safe personnel movement during replacement of components, including the
use of necessary material handling facilities, and during
any in-place testing operation. The provision for inservice inspection, testing, and decontamination of
charcoal filter units is conducted in accordance with the
requirements discussed in the Technical Specification/
Requirements Manual.
- 84. Typical layout with minimum distances for access and servicing is shown on Fig. 12.3-12, 12.3-13, and 12.3-14.
No filter bank is more than three filter units high; each
filter unit is 2 ft by 2 ft. 5. The clear space for doors throughout the plant is a minimum of 2 ft-6 in by 7 ft. 6. The filters are designed with replaceable 2 ft by 2 ft units that are clamped in place against compression seals. The filter housing is designed, tested, and
proven to be airtight with marine bulkhead type doors
that are closed against compression gaskets. The charcoal adsorber cells are gasketless type designed for ease
of installation and replacement. The cell module is designed so that charcoal can be added to or removed from the cell as necessary without violating the integrity of the cell.
Externally mounted test canisters are provided for laboratory
testing of charcoal. 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation 12.3.4.1 Area Radiation Monitoring Objectives 12The area radiation monitoring system is provided to supplement the personnel and area radiation survey provisions of the health physics program described in Section 12.5 to ensure compliance with the personnel radiation protection guidelines of 10CFR20 and
10CFR50.12 RBS USAR Revision 12 12.3-35 December 1999 Provisions for post-accident monitoring are also addressed. The specific River Bend Station positions on Regulatory Guides 1.97, 8.2, 8.8, and 8.12 are provided in Section 1.8. Bases for these
positions are contained in the discussions found in this section
and throughout Chapter 12.
Consistent with this purpose, the area radiation monitors function to: 1. Alert plant personnel entering or working in nonradiation or low-radiation areas of increasing or abnormally
high-radiation levels. 2. Inform the main control room operator of the occurrence and the approximate location of an abnormal radiation
increase in nonradiation or low-radiation areas. 3. Comply with the requirements of 10CFR50, Appendix A, GDC 63, for monitoring fuel and waste storage and
handling areas. 4. Comply with the requirements of 10CFR50, Appendix A, GDC 64, to monitor post-accident radiation levels in the
reactor containment and drywell. 5. In general, assist in maintaining personnel exposures ALARA.The area radiation monitoring system has no function related to
the safe shutdown of the plant, or to the quantitative monitoring
of releases of radioactive material to the environment.
12.3.4.1.1 Area Radiation Monitoring System Design Criteria
The following design criteria are applicable to the area radiation monitoring system: 1. Range - To cover the various ranges anticipated in the plant five different models of the basic instrument with
the following scales are provided: 12 10-2 to 10 3 mrem/hr 10-1 to 10 4 mrem/hr 10 0 to 10 5 mrem/hr 10 0 to 10 8rem/hr (for containment anddrywellpost accident monitoring) 10-2 to 10 4 R/hr (fuel transfer area monitors) 12 RBS USAR 12.3-36 August 1988 2. Alarms - Each area radiation monitor is provided with a red beacon to alarm a high radiation condition, an amber beacon to alarm an alert radiation condition, and a horn for an audible alarm on either high or alert radiation condition. A channel failure alarm light, which is on during normal operation, turns off upon detection of channel failure. All alarms are annunciated in the main control room. Alarm set points are adjustable over the
range of the detector. 3. Sensitivity - Area monitors are sensitive to gamma radiation of photon energies 100 keV and above, with the
exception of the containment post-accident monitors which
are sensitive to photon energies of 60 keV and above. 4. Environmental conditions - The area monitors are designed to operate in the normal environmental conditions for the
areas in which they are located for the design life of the plant. The post-accident containment area monitors
are designed to remain functional during a DBA.
12.3.4.1.2 Criteria for Location of Area Monitors Generally, area radiation monitors are provided in areas to which personnel normally have access and for which there is a potential for personnel to receive high-radiation doses (e.g., in excess of 10CFR20 limits) in a short period of time. Plant areas which
meet one or more of the following criteria are monitored: 1. Areas which during normal plant operations, including refueling, could exceed radiation limits due to system
failure or personnel error. 2. Areas which are continuously occupied to perform plant shutdown following an accident. 3. Area monitors are located in accordance with the requirements in GDC 63 of 10CFR50, Appendix A. 14. Post-accident drywell and containment monitors are located in accordance with the requirements of Regulatory Guide 1.97 and with the requirements in Item II.F.1(3) of NUREG-0737.
1 RBS USAR 12.3-37 August 1988 15. Post-accident vital area monitors are located in accordance with Regulatory Guide 1.97.
112.3.4.1.3 System Description (Area Radiation Monitoring) The area radiation monitoring system detects, measures, and records ambient gamma radiation levels at various locations. It also provides audible and visual alarms in monitored areas and annunciation in the main control room if gamma radiation exceeds
a specified limit. It provides visual indication in the area monitored and at a main control room annunciator if there is a
malfunction in any area monitor.
Each area channel consists of a detector assembly, a check source assembly, a data acquisition unit, indicators, and alarms. All monitors are independent, and failure of one monitor has no
effect on any others. The nonsafety area radiation monitors are powered from 120-V ac regulated buses. The redundant post-accident containment monitors and safety-related monitors are powered by 120-V ac
divisional busses.
The location of each area radiation detector is indicated on the radiation zoning and access control drawings, Fig. 12.3-6 through 12.3-10, and is listed in Table 12.3-1. Consistent with the
previous criteria, the following general areas are monitored:
1.Areas required for safe shutdown 2.Sampling rooms and chemistry laboratory 3.Containment and drywell 4.Fuel storage and handling areas 5.Waste storage and handling areas 6.Maintenance areas.
12.3.4.1.4 Safety Evaluation The area radiation monitoring system is not essential for the safe shutdown of the plant, and serves only to warn plant personnel of high-radiation levels in various plant areas.
Except for the two high-range containment monitors, the two
high-range drywell monitors required for post-accident
monitoring, and the two containment purge RBS USAR Revision 22 12.3-38 isolation monitors, the system serves no active emergency function during operation.
The area radiation monitoring system is designed to operate
unattended for extended periods of time. A visual display of ambient radiation dose rate and trend information for any detector is available on demand in the main control room. These monitors provide audible and visual alarms at the detector and annunciate in the main control room if the radiation levels exceed preset limits. A recorder located on the radiation monitor panel in the main control room provides a permanent record of the radiation levels for the four post-accident safety-related containment and drywell monitors and the containment purge isolation monitors. Also, analog indication is
provided on the main benchboards in the main control room for quick operator assessment of post-accident monitoring (PAM)
conditions. 87 12.3.4.1.5 Calibration and Testing
Each of the monitors is calibrated by the instrument manufacturer prior to shipment using sources certified by the National Institute of Standards and Technology (NIST) or traceable to the
NIST. In-plant calibration, using a standard radioactive point
source traceable to NIST, is done at each refueling or whenever maintenance work is performed on the detectors or in accordance
with the Technical Specifications/ Requirements.
8 7 The proper functioning of each monitor is verified periodically by checking the instrument response to the remotely operated radioactive check source provided with each detector. Proper operation of the monitor's electronics is verified periodically
by the use of internal check circuitry provided in each unit.
12.3.4.2 Airborne Radioactivity Monitoring Objectives
Airborne radioactivity monitoring is provided in compliance with 10CFR20 and Regulatory Guide 1.45. The purpose of the airborne radioactivity monitoring system is to monitor the air within an
enclosure by direct measurement of either the enclosure
atmosphere or the exhaust air from the enclosure. The system
indicates and records the levels of airborne radioactivity and, if abnormal levels occur, actuates alarms. Alarms are provided
to alert personnel that airborne radioactivity is at or above the selected set point. The system provides a continuous record of airborne radioactivity levels, which aids operating personnel in
maintaining airborne radioactivity at the lowest practicable
level.
RBS USAR 12.3-39 August 1987 The main objectives of those in-plant airborne radioactivity monitoring systems which are required for safety are to initiate appropriate protective actions to limit the potential release of radioactive materials from the primary and secondary containment
or the intake of radioactivity into the main control room if predetermined radiation levels are exceeded in major HVAC streams. Additional objectives are to have those systems available under all operating conditions including accidents, and
to provide main control room personnel with an indication of the radiation levels in the major HVAC streams plus alarm
annunciation if high-radiation levels are detected. The safety-related radiation monitoring systems provided to meet these objectives are: 1. Fuel building ventilation exhaust 2. Main control room ventilation. The main objective of the in-plant airborne radioactivity monitoring systems that are not required for safety is to monitor major building exhaust airborne radioactivity levels and to alarm
in the main control room when predetermined levels are exceeded.
The nonsafety-related radiation monitoring systems (RMS) include: 1. Auxiliary building ventilation
- 2. Turbine building ventilation (including condensate demineralizer area) 3. Off-gas building ventilation
- 4. Radwaste building ventilation exhaust
- 5. Main plant exhaust duct monitors
- 6. Containment and drywell atmosphere monitors. 12.3.4.2.1 Airborne Radioactivity Monitoring System Design Criteria The criteria for determining the type of airborne radioactivity monitoring systems are based upon the nature and type of
radioactive releases expected and the location being monitored.
RBS USAR 12.3-40 August 1987 The guidance of ANSI N13.1 and Regulatory Guide 1.21 is followed for the airborne radioactivity monitoring system design. In the case of the drywell and containment radioactivity monitoring systems, which are used to detect leakage from the reactor coolant pressure boundary, the guidance of Regulatory
Guide 1.45 is followed. The design criteria for the safety-related in-plant airborne radioactivity monitoring systems are to: 1. Withstand the effect of natural phenomena (e.g., earthquakes) without the loss of capability to perform
their functions 2. Perform their intended safety functions under normal and postulated accident conditions 3. Meet the reliability, testability, independence, and failure mode requirements of engineered safety features 4. Provide continuous display on main control room panels 5. Permit the checking of the operational availability of each channel during reactor operation with provision for calibration function and instrument checks 6. Assure an extremely high probability of accomplishing their safety functions in the event of anticipated
operational occurrences. The design criteria for the nonsafety-related in-plant airborne
radioactivity monitoring systems are to: 1. Provide data output in the main control room on demand 2. Permit checking the operational availability of each channel during reactor operation with provision for calibration function and instrument checks 3. Perform their intended functions under normal operating conditions for the design life of the plant.
RBS USAR 12.3-41 August 1987 Additional design criteria for the main plant exhaust duct extended range monitor and the containment and drywell atmosphere
monitors are detailed in the following sections.
12.3.4.2.2 Criteria for Airborne Radioactivity Monitor Locations The criteria for locating airborne radioactivity monitors are dependent upon the point of leakage, the ability to identify the source of radioactivity so that corrective action may be performed, and the possibility for exposing personnel to airborne
radioactivity.1. Airborne radioactivity monitors sample drywell and containment atmospheres for reactor pressure boundary
leak detection. 2. The outside air intake ducts for the main control room area are monitored to measure the possible introduction of radioactive materials into the main control room to
ensure habitability of those areas requiring personnel
occupancy for safe shutdown. 3. Exhaust ducts servicing an area containing processes which, in the event of a major leakage, could result in
concentrations within the plant approaching the limits
established by 10CFR20 for plant workers are monitored. Airborne process and effluent radiation monitor locations and functions are summarized in Table 12.3-2. ANSI N13.1 was used as a guide in locating monitors and sample points. Monitor
locations are shown on the shielding arrangement and facilities
drawings, Fig. 12.3-6 through 12.3-10.
12.3.4.2.3 System Description (Airborne Radioactivity Monitors)
12.3.4.2.3.1 Drywell and Containment Atmosphere The drywell and containment atmosphere radiation monitors are designed for early reactor coolant pressure boundary leak
detection in accordance with Regulatory Guide 1.45. A radiation monitor located in the containment is dedicated to sampling the drywell atmosphere. Samples are drawn from the drywell general area and pumped through the monitoring system and then returned to the drywell. Each sample is continuously monitored for particulate and gaseous activities. A removable iodine cartridge filter, which may RBS USAR 12.3-42 August 1987 be used for laboratory analysis, is provided between the moving particulate filter and the gas sample chamber. Alarms are provided for high and alert radiation levels for each channel.
Alarms are also provided for channel or sampling system component failure. All alarms are actuated locally at the monitor and in
the main control room. Recorders are provided in the main control
room to maintain a permanent record of drywell radiation levels.
Similarly, a radiation monitor is dedicated to sampling the containment atmosphere. This monitor is designed to obtain a representative sample of the radiation level in the containment.
The monitor and alarms are identical to those described for the
drywell.The drywell and containment radioactivity monitoring systems are designed to function throughout a design basis seismic event and
are powered by 120-V ac divisional buses.
12.3.4.2.3.2 Fuel Building Ventilation Exhaust The fuel building ventilation exhaust radiation monitors are designed to measure radionuclide effluent emissions to be reported and evaluated in accordance with Regulatory Guide 1.21.
The monitoring system is also designed to activate the safety-grade filter systems on detecting high-radiation levels in the effluent and to monitor post-accident effluent gas and particulate levels. During normal operating conditions, the
effluent from the fuel building is not filtered and these
monitors indicate the airborne radiation levels in the fuel
building.The fuel building ventilation exhaust is monitored by one offline gas monitor and one offline gas and particulate monitor as described in Section 11.5.2.1.2.1. Sampling is performed by an automatic isokinetic sampling system with probes and returns located in the fuel building ventilation exhaust duct.
Connections are available for grab and tritium sampling. High and alert radiation levels, channel failure, and sampling system
failure are alarmed locally and in the main control room.
On a high radiation signal from these monitors, the fuel building exhaust air is diverted through the safety-grade filter trains. The fuel building ventilation exhaust monitors are powered by 120-V ac divisional buses.
RBS USAR 12.3-43 August 1987 12.3.4.2.3.3 Main Plant Exhaust DuctThe main plant exhaust duct radiation monitors are designed to measure radionuclide effluent emissions to be reported and evaluated in accordance wtih Regulatory Guide 1.21. The monitoring system is also designed to monitor post-accident
gaseous and particulate effluent levels. During normal operations, these monitors indicate the general unfiltered level
of airborne radioactivity that is being released to the
environment.
The main stack exhaust radiation monitors are similar to the fuel building ventilation exhaust monitors described in Section 12.3.4.2.3.2. Sampling is performed by an automatic isokinetic sampling system with probes located in the main plant exhaust stack above the auxiliary building. High and alert
radiation levels, channel failure, and sampling system failure
are alarmed locally and in the main control room. The main plant exhaust duct offline gas monitor is powered by the 120-Vac divisional buses. The main plant exhaust duct offline
gas/particulate monitor is powered by the 120-Vac regulated bus.
12.3.4.2.3.4 Main Control Room Ventilation The main control room ventilation radiation monitors are designed to measure the radiation levels in the two inlet ducts of the
main control room ventilation system, and automatically divert
the air through the emergency filter system on detection of high
radiation at the local intake.
Redundant offline gas monitors are provided at both the local and the remote outside air intake ducts. Sampling is performed by sampling systems with probes and returns located near the intakes. The monitors for the local intake are located in the
control building. The monitors for the remote intake are located
in the standby service water pumphouse. The main control room radiation monitors provide a single channel for gaseous activity only. Fixed-particulate and iodine filters are located upstream of the gas sample chamber and can be removed
for analysis in the health physics laboratory. High and alert
radiation levels, channel failure, and sampling system failure
are alarmed locally and in the main control room.
RBS USAR Revision 10 12.3-44 April 1998 On a high radiation signal from the local outside air intake monitors, the air is diverted through safety-grade filter trains. The control room ventilation radiation monitoring system is powered by 120-Vac divisional buses.
12.3.4.2.3.5 Nonsafety-Related Process and Effluent Airborne Radiation Monitoring Systems The nonsafety-related airborne radiation monitoring systems are designed to comply with Regulatory Guide 1.21 for monitoring all effluent streams and major contributing process streams for a
more accurate determination of the airborne radiation levels in the final effluent, and for identification of the sources of airborne radiation in the plant. The following nonsafety-related process ventilation systems exhaust through the main plant
exhaust duct: 1. Auxiliary building ventilation 2. Turbine building ventilation
- 3. Condensate/demineralizer and off gas building ventilation.10The effluent from each of the process streams is monitored by a single gaseous and particulate radiation monitor. The monitors are similar to the drywell and containment radiation monitors described in Section 12.3.4.2.3.1. Sampling is performed by an isokinetic sampling system with probes and returns located in the
ductwork. The effluent streams are not filtered except for the
auxiliary building exhaust air which is automatically diverted
through the SGTS upon a high radiation signal from the reactor building annulus ventilation monitors. High and alert radiation levels, channel failure, and sampling system failure are alarmed
locally and in the main control room.
10Power for these monitors is supplied by 120-Vac buses.
The radwaste building ventilation exhaust effluent radiation monitors are designed to measure radionuclide releases to be reported and evaluated in accordance with Regulatory Guide 1.21.
Since the effluent from general areas of the radwaste building is unfiltered, these monitors also indicate airborne levels of
radiation in the building.
RBS USAR Revision 22 12.3-45 Sampling is performed by an automatic isokinetic sampling system with probes and returns located in the exhaust duct. The radwaste building ventilation monitors are similar to the fuel
building ventilation exhaust monitors described in Section 12.3.4.2.3.2. High and alert radiation levels, channel failure, and sampling system failures are alarmed in the
auxiliary and main control rooms.
The radwaste building ventilation exhaust monitors are powered by
120-V ac regulated buses.
12.3.4.2.4 Safety Evaluation
The in-plant HVAC airborne radioactivity monitors have the
safety-related functions of isolating their particular
ventilation systems and actuating the associated filtered emergency systems, as discussed in Sections 12.3.4.2.1, 12.3.4.2.2, 12.3.4.2.3.2, and 12.3.4.2.3.4. These monitors are redundant, Seismic Category I, and powered from the emergency
power system.
The combination of the airborne radioactivity monitoring system in conjunction with the administrative controls restricting and limiting personnel access, standard health physics practices, ventilation flow patterns through the plant, and plant equipment layout is sufficient to guarantee the safety of personnel
throughout all areas of the plant where access is required.
12.3.4.2.5 Airborne Particulate Detector Ranges and Minimum Detectable Concentrations
Monitor particulate channel ranges are tabulated in Table 12.3-2, and corresponding minimum detectable concentrations (MDCs) are tabulated in Table 12.3-3. Calculated MDC values for monitor particulate channels in the turbine building, fuel building, auxiliary building, radwaste building, and offgas building are sufficiently low that 10 MPC-hr* of airborne iodine and particulate radionuclides in any normally occupied compartment in
these five buildings can be detected.
The particulate radionuclide concentrations to be detected by the
particulate channels are developed by normalizing the building
airborne radionuclide inventories (derived from the building effluent inventories tabulated in Table 11.3-1) to
- Maximum Permissible Concentration, as defined in 10CFR20.
RBS USAR Revision 22 12.3-46 1 MPC and multiplying the resultant radioiodine concentrations by the fraction which is presumed (based upon data in NUREG-0016) to
be in particulate form.
A recorder located on the radiation monitor panel in the main control room provides a permanent record of the radiation levels for each safety-related channel. Also, analog indication is provided in the main control room for quick operator assessment
of the following:
- 1. Fuel building ventilation exhaust
- 2. Main plant exhaust
- 3. Control room ventilation intake
- 4. Radwaste building ventilation exhaust. 7 The continuous monitoring system consists of radiation monitors calibrated by the manufacturer. Calibration standards are traceable to the NIST. Secondary standards counted in reproducible geometry during the primary calibration are supplied with each continuous monitor. Each continuous monitor is
calibrated annually using the secondary radionuclide standards.
7 The count rate response of each continuous monitor to remotely positionable check sources supplied with each monitor is recorded by the manufacturer after the primary calibration, again after installation, and together with the instrument background count
rate at intervals during reactor operation to ensure proper
functioning of the monitors.
Decay corrections are provided for the sources to permit correction for source decay. An electronic circuit check is performed through the use of an internal oscillator or
pulse-generating circuit.
12.3.4.3 Accident Consideration
The plant is designed so that only certain vital areas need be occupied during the course of an accident. Should personnel access to specific areas be necessary, portable monitoring
equipment described in Section 12.5 is used.
Information on post-accident radiation levels is available from safety-related monitors specifically designed for this purpose.
In addition, information may be available from other safety-related and nonsafety-related area and airborne radiation
monitors, and no credit is taken for these.
RBS USAR 12.3-47 August 1987 The nonsafety area monitors and nonsafety-related airborne monitors are supplied by 120-V ac buses. The safety-related airborne radioactivity monitors and area monitors are powered by
120-V ac divisional buses. The ranges of the nonsafety-related monitors include the maximum anticipated operational radiation levels. Radiation or radioactivity levels in excess of the maximum range causes the
monitors to continue to read upscale through an anti-saturation circuit. High or upscale radiation or radioactivity is indicated
by a local audible and visible alarm, alerting personnel not to enter the affected area until a survey with portable equipment establishes that personnel can not be subjected to doses in excess of established limits. Safety-related post-accident monitors have extended ranges to cover postulated accident
radiation levels.
12.3.4.4 Portable Monitors Portable, moving, and fixed-filter paper continuous air monitors (PCAMs) are provided and are normally operated at fixed locations in the plant to provide trend indication. However, during a
period of extensive maintenance in an area with a potential for airborne radioactivity, a PCAM is moved to the maintenance area.
The PCAMs have the capability to be connected to the digital radiation monitoring system via plug-in junction boxes. Junction boxes are located throughout the plant, as shown in Table 12.3-4, and provide the capability to remotely monitor airborne activity in the affected area from the control room, auxiliary control
room, and Rad Tech clean work area in the services building. Intermittent sampling using portable air samplers is used to verify and/or supplement continuous air-monitoring equipment.
These portable air samplers are also used in areas with a potential for some airborne radioactivity, but not enough to
justify installation of a PCAM.
Portable instrumentation is described in Section 12.5.
RBS USAR 12.3-48 August 1987 Reference - 12.3 1. Radiation Shielding Design and Analysis Approach for Light Water Reactor Power Plant RP-8A. Topical Report, Stone & Webster Engineering Corporation, May 1975.
RBSUSAR 12.4-1August198712.4DOSEASSESSMENTRadiationexposuresintheplantareprimarilyfromcomponentsandequipmentcontainingradioactivefluids,andtoalesser extentfromthepresenceofairborneradionuclides.In-plant radiationexposuresduringnormaloperationandanticipated operationaloccurrencesarediscussedinSection12.4.1.
Radiationexposuresatotheronsitelocationsoutsidetheplant whicharisefromonsiteradioactivesources,thepresenceofN-16 intheplant,andradioactivegaseouseffluentsarediscussedin Section12.4.2.DoseassessmentistheestimationofoccupationalradiationexposureatRiverBendStationtoverifythattheplantdesign featuresandproposedmethodsofoperationensurethatradiation exposuresareaslowasreasonablyachievable(ALARA).Thedose assessmentinvolvesestimatesofoccupancytimes,doseratesin plantareas,andthenumberofpersonnelinvolvedinthe followinggeneralcategories:reactoroperationandsurveillance; routinemaintenance;wasteprocessing;refueling;in-service inspection;andspecialmaintenance.Radiationexposurestooperatingpersonnelarewithin10CFR20limits.Radiationprotectiondesignfeaturesdescribedin Section12.3andhealthphysicsprogramoutlinedinSection12.5 assurethattheoccupationalradiationexposures(ORE)to operatingandconstructionpersonnelduringoperationand anticipatedoperationaloccurrencesareaslowasisreasonably achievable(ALARA).Thedoseassessmentevaluationprocessisaimedateliminatingunnecessaryexposuresandtoconsidercost-effectivedose reducingmethodstominimizethenecessaryoperationalexposures.12.4.1ExposuresWithinthePlant TheoccupationalradiationdoseassessmentforRiverBendStationwasperformedusingtheguidelinesofRegulatoryGuide8.19 (3).Thebasisfortheannualman-remestimateswasoperatingdata fromsimilarBWRplantstakingintoaccountthedesign improvementsthatimpacttheoccupationalradiationexposureat RBS.Theprojectedradiationdoseratesthroughouttheplant facilitiesarebasedonassumedradiationconditionsafter5yrs ofplantoperationandthedesignradiationdoseratesfromUSAR Section12.3.OperationaldatafromseveralBWRs (4)showingtheaverageannualman-remperunitoverseveraloperatingyearsto
be RBSUSAR 12.4-2August1987948man-remperyearispresentedinTable12.4-1.Thesedataindicatethatinrecentyearsoccupationalradiationexposures havebeenmuchlargerthantheradiationexposuresreportedfor operatingBWRplantsinthemid-1970s.Theprimaryreasonfor theincreaseinradiationexposurehasbeentheincreasein manpowernecessarytosupporttheexpandingspecialmaintenance
activities.DistributionofannualoccupationalradiationexposuressuggestedinRegulatoryGuide8.19 (3)workfunctionsisgiveninTable12.4-3forallBWRsoverseveralyears.TheaveragevaluesindicatethatoperatingBWRplantshaveapproximately76percent annualoccupationalexposureattributedtomaintenance, 40percentofwhichisroutineand36percentofwhichis special.Inrecentyears,plantmodificationsattributedto feedwaterspargerrepairs,inspection,repairandreplacementof recirculationpiping,ThreeMileIslandlesson-learned modifications,andincreasedsnubberandpipehangerinspections havecontributedtothegrowingamountofoccupationalradiation exposuresassociatedwithspecialmaintenanceworkfunctions.
DesignfeaturesdescribedinSections12.1and12.3fortheRBS BWR/6plantshouldminimizethespecialmaintenancework experiencedatearlier-designedoperatingBWRplants.DesignimprovementsforRBSthatareexpectedtoreducetheoccupationalradiationexposuresincludethefollowing:1.IncorporationofflushconnectionsontheCRDscramdischargevolumeheaderpermitscondensateflushingof pipingtominimizecorrosionproductholdupinahigh personnelaccessarea.2.UseoffilteredcondensatewaterforCRDhydraulicfluidandforthereactorrecirculationpumpsealpurge providesacleanwatersourcethatshouldextendpump seallife.3.Installationofpermanenthoistingsystemandaccessplatformsfortherecirculationpumps,mainsteamisolationvalvesandsafety-reliefvalvesminimizes maintenancetimeindrywell.4.Improvedrefuelingplatformmakesfuelhandlingactivitiesmoveefficientlysolesstimeisspentonthe
platform.
RBSUSAR 12.4-3August19875.Amultistudtensionerreducestheamountofman-hoursnecessarytohandlethereactorvesselheadstuds.6.AnewhandlingtoolandplatformfortheremovalofCRDsfrombeneaththereactorvesselreducescrewsizeandtimespentinthehighradiationarea.7.Improvedfueldesignminimizesthebuildupofradiationlevelsnearreactorcoolantsystemsandreducesthe amountoffuelassemblyshippingactivities.8.Improvedpipingmaterialfortherecirculationsystemeliminatesthespecialmaintenancewhichwasrequiredon olderBWRrecirculationpipingduetostresscorrosion
cracking.9.Inserviceinspectionaccessisimprovedbyremoteequipmentdevelopmentandbyaccessdoorsandplugsinthebiologicalshieldforreactorvesselweld
inspection.10.Installationofapositivepressureleakagecontrolsystemforthemainsteamlineisolationvalvesshould reducethesurveillanceandmaintenanceactivities relatedtotechnicalspecificationrequirementsfor
leakage.11.Useofseparateshieldedcubiclesforlocatingredundantcomponentsandhighlyradioactivecomponentsminimizesradiationexposuresduringmaintenanceactivities.12.Useofmechanicalsnubbersshouldreducethefrequencyofinspectionnecessaryincomparisontothehydraulic-operatedsnubbers.13.InstallationofaCRDflushtankandfiltersystemremoveshighlyradioactivecorrosionandfissionproductsfromtheCRDinternalspriortorebuilding.TheoccupationalradiationexposureforRBScanbedeterminedfor eachoftheRegulatoryGuide8.19 (3)workfunctioncategoriesbyestimatingtheoccupancytimeandthemanpowerrequirements involvedforeachofthesixradiationzonesdefinedin Section12.3.Table12.4-4showstheexpectedoccupancytimes andanticipatedmanpowerneedsforreactoroperationsand
surveillance, routine maintenance, RBSUSAR 12.4-4August1987wasteprocessing,refueling,andinserviceinspection.Theannualman-remforeachworkfunctioncanbeestimatedby assuminganaveragedoserateforeachradiationzoneand assumingthatthereare2,080workinghoursperyearforeach personfortheoperationsandsurveillance,routinemaintenance, andwasteprocessingworkactivities.Refuelingactivitiesare assumedtobebasedon160hrperyearforeachpersonand inserviceinspectionisbasedon192hrperyearforeachperson.
Theseman-hourestimatesareconsistentwiththosereportedfor operatingBWRplantsandpredictedbasedonGEexperience (5).Theaveragedoseratesforeachradiationzoneareusedin Table12.4-4asmorerealisticestimatesoftheactualradiation environmentthatworkerswouldexperienceratherthanthemaximum doserateswhichweredeterminedforshieldingdesigncriteria basedonveryconservativesourcetermassumptions.The estimatedoccupationalradiationdoses,summarizedforeach categoryinTable12.4-5,giveanoverallestimateforRBSas 827.5man-remperyear.Thespecialmaintenancecontributionwas estimatedtobe298.7man-rembasedonpreviousoperationalBWR experience(Table12.4-3).Specialmaintenancecomprises approximately36percentofthetotalannualman-rem.Itis anticipatedthatspecialmaintenanceworkactivitieswillbe significantlyreducedforRBSbasedontheBWR/6design improvementsandtheRBSdesignfeatures.Table12.4-5also providesthepercentagedistributionofthetotalannual occupationaldoseforeachworkfunction.Thisisconsistent withtheaveragedistributionsascalculatedfromoperatingBWR datashowninTable12.4-3.Afurtherestimateoftheoccupationalexposureswasmadebyidentifyingspecifictaskswithineachofthesixworkfunction categories.Variousdatafromoperatingplantsandcurrent publications(References2through10)wereusedtoidentify tasksandestimatemanpowerneedsandexpectedradiationlevels.
Tables12.4-6through12.4-11providetheestimatesof occupationalexposuresbasedontheidentificationofspecific taskswithineachofthesixworkfunctioncategories:routine operationsandsurveillance;routinemaintenance;waste processing;refueling;inserviceinspection;andspecial maintenance.Table12.4-12summarizestheoccupationaldose estimatesforthesixworkfunctions.Acomparisonbetween Tables12.4-12and12.4-5showsthatthedoseestimatesbasedon theidentificationofspecifictasksareconsistentbuttendto belessthanthedoseestimatesbasedonoccupancytimesforeach radiationzoneforthegeneralworkfunctionclassification.The onlyinconsistentestimateisthespecialmaintenancework function.Itisdifficulttopredictallthepossible RBSUSAR 12.4-5August1987specialmaintenanceworkatRBSsincethereislittleoperatingexperienceassociatedwithBWR/6plants.Operatingexperienceat olderBWRs(Tables12.4-1,12.4-3and12.4-5)hasshownthatthe specialmaintenancedoseisapproximately36percentofthetotal annualaverageman-rem(948man-rem/year)whichwouldpredicta valueof342man-remforspecialmaintenaceworkonly.Design improvementsandfeaturespreviouslydiscussedshouldreducethe specialmaintenanceman-remestimateforRBStoarangeof80to 290man-remperyear.RBSalsohasevaluatedpersonnelexposureresultingfromtheactuationofSRVsbasedonReference2.TheSRVdischargeevent consideredintheanalysisistheType2isolationeventwhere thereactorpressureisinitiallycontrolledbythecyclic liftingoftheSRVs.AllSRVsareassumedtoopenwiththelow setreliefvalvesremainingopenfollowingtheclosureofthe othervalves.Designbasisradiationsourcesfornormal operationareusedintheanalysis.Normalventilationin containmentisassumedandairborneconcentrationsarenot correctedforplateoutonthewalls.Thisevaluationdetermines thedosetoanoperatorlocatedintheTIPdriveareawholeaves thisareafollowingtheisolationeventandexitsthecontainment atthepersonnelhatchatelevation114ft.Itisassumedthat theoperatortakes4mintoexitthecontainment.A nonhomogeneousdistributionofdesignbasisairbornesources withinthefirst4minfollowingtheeventisassumedthatis consistentwiththereferencestudy.Thewholebodyandlensof eyedosecalculatedforthiseventis140mrem.Thebetaskin andthyroiddosesare390mremand0.80mrem,respectively.12.4.2ExposuresatLocationsOutsidethePlantStructures Radiationexposuresatlocationsoutsidetheplantarisefrom:1)onsiteradioactivesourcesoutsideplantbuildings,2)direct andair-scatter(skyshine)contributionsduetothepresenceof N-16intheplantbuildings,and3)releaseofgaseouseffluents fromtheplant.ThedoseduetoN-16isthepredominant contributor.Estimateddosesfortherestrictedareaboundary aresummarizedinTable12.4-2.Theseestimatesmeetthedose guidelinesof10CFR20and40CFR190.12.4.2.1DoseduetoRadiationSourcesOutsidePlant BuildingsTheonlyonsitesourceswhichexistoutsideofplantbuildingswithpotentialforadirectradiationdose RBS USAR Revision 24 12.4-6 8A 8 contribution to persons outside of plant structures are the condensate storage tank, the temporary dry active waste storage
facilities, Radioactive Material Storage Areas and the two turbine rotor modular enclosures. However, the minimal activity within the tank, temporary dry active waste storage facilities, and the two turbine rotor modular enclosures produces a
negligible dose rate at the RAB.
8 8A 12.4.2.2 N-16 Dose Contributions 14 Radiation dose rates as measured near plants such as River Bend consist of three parts. The parts are direct radiation, radiation from scattered gammas, and background radiation.
Background radiation is a combination of radiation from
terrestrial sources and cosmic radiation. Scattering sources comes from a direction other than the original source.
Scattering can be accomplished by buildings around and near the
source, and by the atmosphere (skyshine).
The Monte Carlo technique was used to evaluate the direct and scatter doses for River Bend Station including the increase N-16
activity as a result of Hydrogen Water Chemistry. The modeling case chosen for River Bend was that for a plant with moisture separators/reheaters (MSRs) on the operating floor, with 3 feet
thick perimeter walls, 1.5 feet thick MSR covers, and 1 foot thick inner walls. All walls were assumed to be concrete by the model. While this model is not identical to River Bend design, it is a reasonable match since River Bend's steel inner walls (about six inches thick) have the same stopping power as one foot of concrete. The correlation developed was validated using
observed data.
The results indicate that the dose rates drop below 1mR/yr at 1900 feet. The minimum distance from the turbine building to the
RAB is 2,760 feet.
14 RBSUSAR 12.4-7August198712.4.2.3ExposuresDuetoAirborneActivityDoseratesresultingfromairborneactivityattheRestrictedAreaBoundary,basedon2,000hr/yroccupancy,arelistedin Table12.4-2.12.4.3ExposuretoConstructionWorkers RBSisasingle-unitplant;therefore,estimatedannualdosestoworkersconstructinganadditionalplantduetoradiationfroman existingoperatingplantarenotapplicable.
RBSUSAR 12.4-8August1987References-12.41.RadiationShieldingDesignandAnalysisApproachforLightWaterReactorPowerPlants,RP-8A.Stone&WebsterEngineeringCorporation,Boston,MA,May
1975.2.MarkIIIContainmentDoseReductionStudy,GE22A5718,Revision1,January29,1980.3.OccupationalRadiationDoseAssessmentinLight-WaterReactorPowerPlantsDesignStage Man-RemEstimates,RegulatoryGuide8.19, Revision1,June1979.4.OccupationalRadiationExposureatCommericalNuclearPowerReactor1982,NUREG-0713,Volume4, December1983.5.OccupationalRadiationExposure,GeneralElectricCompany,NEDO-24606,January1979.6.AnAssessmentofEngineeringTechniquesforReducingOccupationalRadiationExposureat OperatingNuclearPowerPlants,AtomicIndustrial Forum,February1980.7.Pelletier,C.A.,etal.CompilationandAnalysisofDataonOccupationalRadiationExposureExperimentatOperatingNuclearPowerPlants.
AIF/NESP-005,September1979.8.Hall,T.M.DeterminingEffectivenessofALARADesignandOperationalFeatures,NUREG/CR-0446, April1979.9.OccupationalRadiationExposureReductionTechnologyPlanningStudy,Stone&WebsterCorporation,EPRINP-1842,May1981.10.Pettit,P.J.CompendiumofDesignFeaturestoReduceOccupationalRadiationExposureatNuclear PowerPlants,AIF/NESP-20,April1981.
RBS USAR Revision 17 12.5-1 12.5 HEALTH PHYSICS (RADIATION PROTECTION) PROGRAM 12.5.1 Organization 12.5.1.1 Program and Staff Organization
The objective of the radiation protection program is to control the occupational radiation exposure of personnel within the plant and reduce exposure to the general public and offsite environs from ionizing radiation resulting from operation of the plant.
Specifically, the radiation protection program consists of policies, procedures, instructions, rules, and practices to keep
individual radiation exposure within the limits set forth in
10CFR20 and beyond that to maintain total radiation exposure of personnel as low as is reasonably achievable (ALARA)
(Section 12.1). The program is designed to assure that the plant operations meet the radiation protection and training requirements of 10CFR19, 10CFR20, 10CFR50 Appendix I, and the River Bend Station positions on NRC Regulatory Guides 8.2, 8.8, and 8.10, which are provided in Section 1.8. The program assures that radiation protection training is provided, that personnel and in-plant area radiation monitoring is performed, and that records of training, exposure of personnel, and surveys are maintained. The program provides guidelines covering the
handling and monitoring of radioactive materials. 1372The RBS Plant Staff organization, which includes the radiation protection organization, is shown on Fig. 13.1-2. The General Manager is responsible for the overall effectiveness of the radiation protection program. The General Manager delegates the conduct of the radiation protection program to the Manager - Radiation Protection.
Manager - Radiation Protection is equivalent to the Radiation Protection Manager referred to in Regulatory Guide 1.8 (Revision 1-R) and is responsible for the
enforcement of the program. It is the responsibility of all plant supervisory personnel to ensure that plant personnel are made aware of the management commitment to keep all occupational
radiation exposure ALARA.
The Manager - Radiation Protection is responsible for establishing the radiation protection program and provides technical assistance for conducting the program. He assures that
personnel radiation exposure is maintained ALARA.
The Manager - Radiation Protection is required to have a degree in science or engineering or the equivalent with some formal training in health physics and at least 5 yr. responsible professional experience in health physics. A minimum of 3 of the
5 yr. of experience in applied radiation protection is required
at a light-water reactor facility.
2 7 13 RBS USAR Revision 20 12.5-2 1372Radiation protection supervisors and management personnel report to the Manager - Radiation Protection and ensure that the
established radiation protection program is properly implemented.
A radiation protection supervisor implements the operational aspects of the radiation protection and ALARA Programs. This is
accomplished through the use of approved procedures, trained and
qualified radiation protection personnel, properly calibrated equipment, and appropriate ALARA concepts. A radiation protection supervisor is designated backup for the Manager - Radiation Protection and meets the requirements of Regulatory Guide 1.8 (Revision 1-R). A radiation protection supervisor implements the Dosimetry and Radiological Health Programs. This is accomplished through the
use of approved procedures, an accredited personnel dosimetry program, maintenance of exposure histories, maintenance of
properly calibrated equipment, maintenance of whole body, D osimetry and bioassay counting capabilities, and providing respiratory mask fit tests.
28The organization also includes personnel to provide for the radwaste Process Control Program processing solid radwaste, trending, trend analysis, shield calculations, dose assessment, emergency planning, internal audits of radiological work
practices, and computer applications.
8ALARA concepts are implemented by the ALARA coordinator through procedure reviews, Radiation Work Permits, design reviews, system
walkdowns, and preplanning. Corporate support is provided for external audits and assessments.
712.5.1.2 Radiation Protection Program Objectives The objectives of the radiation protection program are:1. To provide administrative control of plant personnel to ensure that radiation exposure is maintained within the limits of 10CFR20 and that such exposure is maintained
ALARA.1372. To provide administrative control over plant effluent releases and to limit these releases below 10CFR20 values
and the values given in the plant procedures.
7 RBS USAR Revision 7 12.5-3 January 1995 3. To establish a radiation protection records systemwhich meets the requirements of applicable federaland state regulations.12.5.1.3 Radiation Protection Program The radiation protection program becomes effective in phases upon implementation of the preoperational environmental radiological program, 2 years prior to scheduled operation start, and continues throughout operation of the station. The program includes policies, procedures, instructions, rules, and practices to ensure that the program objectives are fulfilled in a safe and reasonable manner. These program facets are reviewed periodically, and recognized improvements are incorporated. The
program includes: 1. Necessary measures and guidelines to reduce exposureof personnel and to limit effluent releases. 2. Appropriate radiation protection training of personnel commensurate with their respective responsibilities. 3. Provisions for additional Emergency Plan training of personnel who may be assigned to radiation emergency
teams.4. Respiratory protection training and supplies of equipment where needed to reduce internal exposure to personnel. 5. Appropriate dosimetry of personnel for external dose assessment.76. An internal dose assessment program which includes passive monitoring and bioassays.
- 77. Requirements for segregating and appropriately posting radiologically controlled areas to control exposure
potential.8. Procedures for access to radiologically controlled areas including radiation work permit authorization. 9. Training in duties to be performed in radiation areas. 10. Radiological instrumentation and equipment to assess exposure accurately.
RBS USAR Revision 12 12.5-4 December 1999 11. Procedures for properly handling incoming and outgoing shipments of radioactive materials in accordance with NRC, DOT, and other applicable requirements. 12. Proper protective clothing and equipment to limit the potential hazard of transferable contamination. 13. A radiological environmental monitoring program for evaluating the environmental impact of plant operation. 1214. The radiation protection program coordinates with the emergency operating procedures and servere accident procedures for staff and facility activation during an
emergency. Emergency planning is discussed in
Section 13.3.
1212.5.2 Equipment, Instrumentation, and Facilities 12.5.2.1 Radiation Protection Facilities An assessment of compliance of radiation protection facilities design with Regulatory Guides 1.97, 8.3, 8.4, 8.8, 8.9, 8.12, 8.14, and 8.15 is provided in Section 1.8. The radiation protection facilities are located in the services building and contain:1. Radiation protection offices and work areas 2. Chemistry and radiochemistry laboratories
- 3. Counting rooms for radioactivity analysis
- 4. Equipment and supplies storage spaces
- 5. A first aid facility
- 6. Personnel and small item decontamination facilities
- 7. Instrument calibration, maintenance, and storage facilities8 Locker rooms and toilet facilities 2 2Radiation survey instruments, air monitoring equipment, dosimeters, counting instruments, and other miscellaneous radiation protection supplies are stored in the services building. The radiation protection offices and work areas are equipped with filing cabinets, desks, and tables to RBS USAR 12.5-5 August 1987 provide work space for radiation protection personnel and storage space for records. An adequate supply of clean protective clothing and respiratory equipment is supplied by onsite laundry facilities or vendor supply services. Protective clothing and respiratory protection equipment are available in the services building and at local
contamination area access points set up at job sites. Change facilities are provided for these areas consisting of benches and
clothes racks.
Radiation protection sample counting equipment is located in the radiation protection workrooms. Equipment used for routine counting of smears/swipes and air samples such as pancake probes
or end window G-M counters, alpha and beta scintillation detectors, and/or gas flow proportional counters are located in this area. Samples requiring gamma isotopic analysis can be counted using spectrometric systems located in the chemistry
laboratories.A decontamination facility for personnel is provided in the services building. Decontamination of personnel is conducted under the direction of radiation protection personnel. The decontamination facility consists of a change area, monitoring area, lavatory, and showers. Decontamination of accident victims
on stretchers can also be performed. All personnel exiting contaminated areas are required to monitor/frisk themselves for contamination. In addition to local
frisking, a portal monitor or equivalent is also provided at the exit of the radiological controlled areas and the primary access
point building to alarm in case of contamination on person or
clothing.12.5.2.2 Radiation Protection Instrumentation 12.5.2.2.1 Laboratory Instrumentation Laboratory instrumentation located in the radiation protection and chemistry work areas allows plant personnel to ascertain the
radioactive material present in survey samples. Typical samples would be contamination survey smears, airborne survey filters, and charcoal cartridges. There is also capability to conduct tritium surveys and process various other types of samples.
Counting room instrumentation at least equivalent to that listed in Table 12.5-1 is maintained. The criteria for selection of
these various counters is to obtain instrumentation that can reliably and efficiently count samples, provide the necessary low backgrounds and sensitivities, and analyze the counting data to provide information in a more easily usable RBS USAR Revision 23 12.5-6 8 7 form. Each counting system is checked and calibrated at regular intervals when in use and after undergoing repair with standard radioactive sources traceable to the National Institute of Standards and Technology (NIST). They are also calibrated prior to use if the period since the last calibration is greater than the regular interval. Counting efficiency, background count
rates, and high voltage settings are checked.
7 8 12.5.2.2.2 Portable Survey Instrumentation
Portable survey instrumentation is located in the radiation protection work area and at inplant control points. Portable equipment allows plant personnel to perform surveys for direct
radiation, airborne radioactivity, and surface contamination.
The criteria for selection of these instruments is to obtain accurate and reliable instrumentation that is easily serviced and
covers the entire spectrum of radiation measurements expected at
the station during normal or accident conditions. 8 7 Each portable survey instrument is calibrated annually, when in use, and after undergoing repair. They are also calibrated prior to use if the period since the last calibration is greater than
the regular interval. Calibrations are performed using calibration sources traceable to NIST, and instruments are source checked to verify proper operation in accordance with plant
procedures. The quantities of each type of instrument permit calibration, maintenance, and repair without causing a shortage in operational instrumentation. Portable radiological survey
instrumentation is listed in Table 12.5-2.
7 8 12.5.2.2.3 Monitoring Instruments for Personnel 9 7 4 2 Monitoring of personnel is provided by the use of a DLR {dosimeter of legal record, which may be an optically stimulated luminescent dosimeter (OSLD) or a thermoluminescent dosimeter (TLD)}, or direct-reading dosimeters, or albedo neutron TLDs or the equivalent, and survey instrumentation. Personnel who are required to be monitored for radiation exposure in accordance
with 10CFR20.1502 are issued a DLR. DLRs are analyzed periodically and when special circumstances warrant. DLR readings are normally used as the official record of personnel exposure if required by 10CFR20.1502. Calibrations and quality
control performed on the TLDs and the TLD system are in accordance with NRC requirements, by means of a NAVLAP certified
facility.
2 4 7 9 RBS USAR Revision 23 12.5-7 7 Direct-reading dosimeters are worn by personnel in the controlled area as specified by plant instructions. These dosimeters provide a day-to-day or job-to-job estimate of personnel exposure. Direct-reading dosimeters are provided as shown in
Table 12.5-3.
7 9 If neutron exposure is expected, personnel are issued neutron-sensitive DLRs. These DLRs normally provide the official
record of exposure of personnel to neutron radiation.
In the case of a lost or improper reading on a DLR, or entry into the Controlled Area without the properly issued dosimetry, the individual exposure is estimated using appropriate methods and
documented in accordance with plant procedures.
9 Extremity dosimetry is provided if extremities are expected to receive exposure significantly higher than that to the body. Its use is normally specified on the RWP. The direct-reading dosimeters are calibrated annually, and when damage is suspected, or after repair is performed. 7 Survey instrumentation for the measurement of personnel contamination consists of G-M and/or gas flow proportional count rate meters (contamination friskers), Personnel Contamination
Monitors (PCM's) and portal monitors. These instruments are
calibrated annually or prior to use after undergoing repair. The monitoring instrumentation for personnel listed in Table 12.5-3
is maintained as a minimum.
7 12.5.2.2.4 Radiation Protection Equipment
Portable air samplers are used to collect samples for laboratory determination of airborne radioactive material concentrations.
Air samplers are calibrated for flow semiannually. Samples may be counted for radioactive particulate, radioiodine, and airborne
gaseous activities. Portable continuous air monitors are used to monitor airborne concentrations at specific work or field locations. Local indication is provided as well as trend information. Visual and audible alarms are provided with
variable set points.
The portable air samplers shown in Table 12.5-4 may be used to monitor iodine levels in buildings under accident conditions as required in Item III.D.3.3 of NUREG-0737. Sample analysis is performed in the Services Building laboratory facilities (hot lab and counting rooms) on instruments described in
Section 12.5.2.2.1.
Respiratory protection equipment is available at specific issue
points. Self-contained breathing apparatus for RBS USAR Revision 7 12.5-8 January 1995 emergency use is available in the services building and at emergency equipment storage lockers. 7An instrument calibrator is used for calibrating gamma dose rate instrumentation. This is a self-contained, heavily shielded, multiple source calibrator. Neutron, beta, and alpha radiation sources are also available for instrument calibration and/or
response checks. Sources are traceable to an NIST source.
7Protective clothing is supplied for personnel working in radiologically controlled areas. The clothing required for a
particular instance is prescribed by radiation protection
personnel on a radiation work permit based on actual or potential
radiological conditions.7An adequate inventory of protective clothing is maintained on hand to support plant operations and maintenance activities.
This clothing typically includes lab coats, coveralls, hoods, caps, plastic suits, gloves (plastic, rubber, cloth), shoe
covers, boots, and rubbers as necessary. Tape is provided for
sealing.7Additional contamination control supplies are available. These include vacuum cleaners equipped with HEPA filters, mops, absorbent paper, plastic sheets and bags, barricade ropes, signs, and labels.
Radiation protection equipment includes that given in Table 12.5-4 as a minimum. 12.5.2.2.5 Other Radiation Protection Instrumentation The area radiation monitoring system is installed in areas where it is desirable to have constant dose rate information. Monitors
indicate dose rate locally and/or in the control room and provide local audible and visual alarms upon reaching a preset dose rate.
Fixed continuous airborne radioactivity monitors are also
provided at strategic locations where the potential for exposure
of personnel to airborne contamination is most likely.
12.5.3 Procedures Section 13.5 provides information on the procedures which cover applicable areas of plant operations and maintenance. In addition, this section describes certain methods embodied in procedures and/or instructions to ensure that personnel radiation exposure is maintained ALARA. Assessments of compliance with Regulatory Guides 1.8, 1.16, 1.33, 1.39, 8.2, 8.7, 8.8, 8.9, 8.10, and 8.13 are provided in Section 1.8, and compliance with 10CFR19 is addressed in RBS USAR 12.5-9 August 1987 Section 12.5.1.1. The preceding are incorporated into procedures and/or instructions as appropriate. Strict adherence to the plant radiation protection procedures and instructions ensures that personnel radiation exposures are
within the limits of 10CFR20 and ALARA concepts. 12.5.3.1 Radiation and Contamination Surveys Radiation protection personnel conduct routine surveys throughout the plant for radiation, contamination, and airborne radioactive materials. The techniques used to perform these surveys are specified in radiation protection procedures. The frequency of routine surveys is specified in radiation protection procedures and depends on the location, type of area and how it is used, frequency of use, potential for radiological hazard, and likelihood of changes in radiological conditions. Records of
surveys are retained for reference, and appropriate data are posted for information of plant personnel. Availability of
current survey information aids in keeping exposure of personnel ALARA. Radiation surveys may be performed for alpha, gamma, beta, and/or neutron exposure rates. Contamination surveys are performed to establish gross beta-gamma contamination levels and may be processed for specific types of radiation (alpha-beta-gamma) or specific radionuclide (via gamma spectroscopy). Air samples are taken to establish airborne concentrations of particulates, noble gases, and/or radioiodine, and specific radionuclide information may be obtained. Special surveys related to specific operations or maintenance activities may be performed prior to, during, or after such activities, based on information required to keep radiation exposure ALARA.12.5.3.2 Procedures and Methods to Maintain Exposures ALARA ALARA considerations are incorporated into various types of plant procedures and instructions which cover activities involving exposure. Examples of procedures and methods used to maintain radiation exposure ALARA in the various operational categories
are discussed in this section.
12.5.3.2.1 Refueling
Procedures and methods used to maintain radiation exposure ALARA during refueling outages include the following: 1. An RWP is used to provide positive radiological control over work in progress.
RBS USAR Revision 7 12.5-10 January 1995 2. Training is conducted to familiarize workers with procedures and equipment to be used. 3. Before removing the vessel head, the primary system is degassed and sampled to minimize expected airborne activity when the head is removed. 744. During movement of irradiated fuel assemblies, the active fuel is maintained under a minimum height of water as
required by plant procedures.
- 45. The refueling cavity water is normally filtered to reduce the activity in the water and to lower exposure rates.
- 76. Radiation levels in work areas are monitored and precautions taken as necessary, consistent with ALARA. 7. Filtered or exhaust ventilation is operated as appropriate to minimize airborne radioactive material. Certain components and surfaces are maintained wet to minimize personnel exposure from potentially high airborne radioactivity concentrations during refueling operation, including the short time the steam dryer and part of the steam separator are out of water. In addition, administrative controls are implemented to
minimize personnel exposure using respiratory protection equipment when necessary and containment access control during
the transfer operation. 12.5.3.2.2 Inservice Inspection Protection procedures and methods used during inservice inspections to maintain radiation exposure ALARA are as follows: 1. An RWP is used to provide positive radiological control over work in progress. 2. Training is conducted to familiarize workers with procedures, equipment, radiation and contaminationlevels, and protective clothing requirements appropriate to a
particular job. 3. Insulation is designed, where practical, for ease of removal and replacement where removal is required for
repetitive inspections.
RBS USAR 12.5-11 August 1987 4. Equipment is calibrated and checked for proper operation prior to entry into a radiation area. 5. Temporary shielding is used, where practicable, to reduce radiation exposure. 6. Filtered or exhaust ventilation is provided, as necessary, to minimize airborne radioactive material. 12.5.3.2.3 Radwaste Handling Procedures and methods used to maintain radiation exposure ALARA during radwaste handling include the following: 1. Handling of radwaste by personnel is minimized by plant design and local procedures. 2. The volume of radwaste which requires handling has been reduced by plant design and operating practices. 3. Special tools are used when practical to minimize time spent close to the radioactive source. 4. An overhead crane is provided to move drums or liners as required.5. Adequate storage has been provided to minimize multiple handling of drums. 6. The labeling of drums or liners is completed prior to filling.7. Filtered or exhaust ventilation is provided to minimize the airborne radioactive material from waste handling
operations.12.5.3.2.4 Spent Fuel Handling, Loading, and Shipping Procedures and methods used during spent fuel handling, loading, and shipping to maintain radiation exposure ALARA include the
following:1. An RWP is used to provide positive radiological controls over work in progress. 2. Training is conducted to familiarize workers with procedures and equipment required to complete
assignments.
RBS USAR Revision 7 12.5-12 January 1995 743. Movement of irradiated fuel assemblies and loading into shipping casks is accomplished with the active fuel maintained under a minimum height of water as required by
plant procedures.
4 74. Fuel handling cranes and extension tools are used to handle shipping casks, fuel assemblies, and inserts. 5. The spent fuel pool water is filtered to reduce the radiation exposure of personnel in the area. 6. The spent fuel pool water is cooled, and surface air ventilation is provided to minimize the airborne
radioactive material. 7. After a shipping cask is loaded with spent fuel, it is decontaminated using a pressurized water washing device
prior to shipment. 12.5.3.2.5 Normal Operation Procedures and methods used during normal operation to maintain radiation exposure ALARA include the following: 1. By station design and construction, significant radiation sources are minimized and shielded. 2. The RWP system is used to provide positive radiological control over work in progress. 3. Training and retraining in radiation protection is provided to persons assigned tasks in controlled areas. 4. Controlled areas are conspicuously posted and maintained in accordance with 10CFR20. 5. Special access control procedures are used for entry to areas where significant exposure might be received. 6. Protective clothing and equipment are provided. 7. An area radiation monitoring system is installed and provides indication of radiation levels with local alarms, where appropriate. 8. The ventilation system is designed to minimize the spread of airborne contamination.
RBS USAR Revision 7 12.5-12a January 1995 79. Shielding effectiveness is verified by an initial startup survey for gamma and neutron radiation.
7 RBS USAR Revision 7 12.5-12b January 1995 THIS PAGE INTENTIONALLY LEFT BLANK RBS USAR 12.5-13 August 1987 12.5.3.2.6 Routine Maintenance Examples of procedures and methods used during routine maintenance to maintain personnel radiation exposure ALARA are as
follows:1. Maintenance work involving systems that collect, store, contain, or transport radioactive materials must be
covered by an approved RWP. 2. Training is provided, as required, to accomplish assigned tasks in controlled areas. 3. Routine maintenance procedures incorporate appropriate precautions and radiological considerations. 4. Equipment is moved to areas with lower radiation and contamination levels for maintenance when practical. 5. Special tools are used when practical. 6. Portable shielding is used as practical.
- 7. Periodic monitoring by radiation protection technicians, as radiological conditions warrant, is provided. 8. Post-job debriefings are utilized for high exposure jobs to obtain input from personnel actually performing the work. This is utilized in revising procedures and
instructions as appropriate for ALARA considerations.
12.5.3.2.7 Sampling Procedures and methods of maintaining personnel radiation exposure ALARA during sampling include the following: 1. Sampling of radioactive systems is normally performedinside sample hoods. 2. Procedures specify the protective clothing and sampling techniques to be used. 3. Monitoring of radiation levels in the work area during sampling and of the sample container when sample is
collected.4. Training in proper handling of sample containers after samples are collected.
RBS USAR Revision 7 12.5-14 January 1995 5. Training in proper storage and disposal of radioactive samples.12.5.3.2.8 Calibration Procedures and methods of maintaining radiation exposure of personnel ALARA during calibration include the following: 1. Detailed procedures are followed for each calibrator use. 2. The instrument calibrator is properly shielded.
- 3. An interlock is provided to prevent opening the calibrator door while the source is exposed. 4. Portable sources used to calibrate fixed instruments are transported and maintained in shielded containers as required.5. The RWP system is used to provide positive radiological controls over calibration. 6. Where possible, fixed instruments are situated in a low radiation area so that necessary test signals can be
inserted with the instrument in place. 12.5.3.3 Access Control One of the means available for ensuring that personnel radiation exposure is maintained ALARA is to control access to areas of
potential exposure. Areas of the plant are designated, posted, and controlled according to the degree of radiological hazard or
potential hazard in that area. 7 For radiological control purposes the restricted area is divided into clean and controlled areas. Controlled areas are posted as
required by 10CFR20. Posted areas include the following: 1. Radiation areas 2. High radiation areas
- 4. Very high radiation areas
- 5. Contaminated areas 7
RBS USAR Revision 20 12.5-15 76. Radioactive material areas 7. Airborne radioactivity areas 8 8. Deleted 810 9.RadiologicallyControlled Area (R CA) (Controlled Access Area (CAA) is an equivalent posting for RCA but RCA is preferred) 7Normal access to the RCA is through an established control point.
Regulations and precautions that apply as minimum conditions for entry to a controlled area include:
101. Personnel monitoring equipment 2. Training as required by 10CFR19 73. Other requirements posted at the area entrance or imposed by radiation protection personnel. Positive control is exercised over entries into high radiation areas as specified by 10CFR20 and station procedures. Entrance into radiation areas as well as any work performed in
radiologically posted areas requires an RWP.
712.5.3.4 Contamination Control Controlling the spread of removable contamination is necessary to ensure that radiation exposure of personnel is maintained ALARA.
Contamination limits for personnel, equipment, and surface areas are delineated in radiation protection procedures. Surveys for removable contamination are performed by radiation protection
personnel on a routine basis. Special surveys are performed in
areas whenever a change in contamination levels is likely and may
have radiological importance. Areas found contaminated are posted, isolated (with ropes, barriers, etc.), and decontaminated as practical. The level of contamination and number of such areas are minimized. Step-off pads are used at exits from contaminated
areas to help control the spread of contamination. Tools and equipment used in potentially contaminated areas are surveyed for removable contamination and fixed contamination prior to removal from controlled areas. If tools or equipment do not meet the clean area limits, they must be decontaminated before removal from the controlled area unless specifically authorized and controlled by RBS USAR Revision 8 12.5-16 August 1996 radiation protection personnel. Some tools and equipment are for use only in a contaminated area. They are normally identified with a conspicuous unique marking and stored in designated areas.
These items are also surveyed periodically and decontaminated as
appropriate. They are considered to be contaminated and are used
only by personnel in appropriate protective clothing. Personnel are protected from inhalation, ingestion, or absorption of contamination by the use of protective clothing, respiratory protection equipment, and good work practices. Personnel are responsible for surveying themselves for contamination when exiting contaminated areas. If contamination is found, the
individual notifies radiation protection personnel and is
decontaminated in accordance with approved plant procedures. 12.5.3.5 Respiratory Protection 87When airborne radioactivity is detected in the restricted area at 30 percent of the derived air concentration values in 10CFR20, the area is isolated and posted as an airborne radioactivity area
and access is controlled. Entry into these areas requires the
issuance of an RWP and monitoring of internal exposure. The RWP
system specifies if respiratory protection is required. Air sampling and counting are used to ensure that respiratory protective equipment specified on the RWP remains appropriate.
The respiratory protection program conforms to 10CFR20
requirements.
7 812.5.3.6 Radiation Work Permit The purpose of the RWP is to help control exposure of plant personnel to radiological hazards, thus maintaining radiation exposure of personnel ALARA. The RWP maintains exposure of
personnel ALARA by providing requirements and instructions to: 1. Control access to radiologically hazardous areas 2. Restrict the spread of radioactive contamination 7 73. Assure that proper radiological procedures and precautions are followed4. Provide a history of how work was completed and exposure received by personnel which can be used for review in efforts to reduce exposure.
RBS USAR Revision 20 12.5-17 5. Document exposures of personnel to ensure no one exceeds the limits of 10CFR20. RWPs are issued by the radiation protection department and approved in accordance with radiation protection procedures.
Typical information contained on an RWP may include: 1. Type and location of job 2. Exposure limits of personnel involved
- 3. Radiological conditions of work area (radiation, contamination, airborne activity, etc.) 4. Type and number of dosimetry devices to be used
- 5. Protective clothing and equipment requirements
- 6. Allowable working time (stay time)
- 7. Any special precautions or instructions. Adherence to the RWP requirements ensures that work in posted areas is performed in a radiologically safe manner, and adherence
to the permit is mandatory. Repeated or willful violation of RWPs is cause for administrative action. Reporting and
documentation is specified in administrative procedures. 12.5.3.7 Dosimetry of Personnel 97Plant employees, contractors, and visitors who are required to be monitored for radiation exposure in accordance with 10CFR20.1502 are required to wear a D LR and a direct reading dosimeter when in a radiologically controlled area, in accordance with 10CFR20.
Only those individuals who have completed the required training and qualification program are authorized to enter radiologically controlled areas unescorted. Other personnel must be accompanied
at all times by a properly trained escort. Personnel who are not required to be monitored for radiation exposure in accordance with 10CFR20.1502 are required to wear at least a direct-reading dosimeter when in a radiologically
controlled area. 4 Neutron sensitive D LR s are issued to those individuals who are expected to be subjected to neutron exposure. RWPs specify the need for neutron dosimetry. D LRs are processed periodically and 4 7 9 RBS USAR Revision 25 12.5-18 9 7 4 provide the official exposure record of personnel. Direct-reading dosimeters are read at least daily by each individual who
has entered a radiologically controlled area to provide a daily estimate of exposure. The DLRs are processed more frequently when an individual's exposure status is in doubt. The results of direct-reading dosimeters and DLRs are recorded and maintained by
plant personnel in accordance with approved plant procedures.
4 9 10 The internal monitoring program consists of "passive" monitoring and bio-assays. Portal monitors (or the equivalent) are set up
at major exits to the radiologically controlled area to provide for passive internal and external monitoring. The whole body counters plus in-vitro analysis (as appropriate) are used in the determination of internal dose. Plant procedures provide guidance
with regard to bio-assays.
10 Exposure data of all personnel are collected and recorded on NRC Form 5 or the equivalent. Occupational exposures incurred by individuals prior to working at River Bend Station are summarized
on NRC Form 4, or the equivalent.
7 12.5.3.8 Radiation Protection Training
Plant personnel (permanent and temporary), visitors, and contractors are properly trained in the fundamentals of radiation
protection and certified or qualified prior to being allowed to enter the radiologically controlled area unescorted. Training provided is commensurate with the degree of radiological hazard associated with specific work assignments. Permanent plant
employees are required to attend a retraining program in radiation protection on a periodic basis. Personnel whose duties do not require entry into radiologically controlled areas are trained to recognize radiation and contamination area signs and are made aware of the reasons for keeping out of such areas.
Training must meet the requirements of 10CFR19.
Plant personnel (permanent and temporary) and contractors who enter areas where respiratory protection is needed are trained, certified, and/or qualified in respiratory protection prior to being allowed to enter the area. These RBS USAR Revision 7 12.5-19 January 1995 same people are required to attend a retraining program in respiratory protection at least annually. Personnel whose duties do not require entry into controlled areas are trained to recognize respiratory protection area signs and are made aware of
the reasons for keeping out of such areas. Training is
consistent with NUREG-0041. 12.5.3.9 Radioactive Materials Safety Program
Various types and quantities of radioactive sources are employed to calibrate the process and effluent radiation monitors, the area radiation monitors, and portable and laboratory radiation
detectors. Check sources that are integral to the area, process, and effluent monitors consist of small quantities of byproduct material that do not require special instructions for radiation protection purposes. The same consideration applies to solid and
liquid radionuclide sources of exempt quantities or concentrations which are used to calibrate or check the portable and laboratory radiation measurement instruments. Recognized methods for the safe handling of radioactive materials are incorporated in procedures and instructions to ensure proper handling of radioactive material. External doses are minimized by a combination of time, distance, and shielding considerations.
Internal doses are minimized by the measurement and control of
loose contamination.Sealed radionuclide sources having activities greater than the quantities of radionuclides defined in 10CFR20 and 10CFR30 are subjected to material controls for radiological protection.
These controls include:1. Monitoring of all shipments containing radioactive materials in accordance with 10CFR20. 72. Inventory and leak testing of sealed sources in accordance with approved plant procedures. 3. Labeling of each source or source holder in accordance with 10CFR20.1904.
- 74. Storage in a controlled area of each source that is not installed in an instrument or other equipment. 5. Maintenance of records on the results of inventories, leakage tests, use, location, and the receipt and final
disposition dates for all sources. Additional details of the materials safety program are provided
in the plant instructions.
RBS USAR 12.5-20 August 1987 Radioactive sources that are subject to the material controls described herein are used or handled only as authorized by radiation protection personnel. Each individual using these sources is familiar with the radiological restrictions and limitations placed on their use. These limitations protect both
the user and the source.