ML12045A255

Updated Revision to Nine Mile Point, Units 1 & 2 - Emergency Action Levels, Attachments 4 and 5

ML12045A255
Person / Time
Site:	Nine Mile Point
Issue date:	01/13/2012
From:	Constellation Energy Nuclear Group, EDF Group, Nine Mile Point
To:	Office of Nuclear Reactor Regulation
References
TAC ME6221, TAC ME6222
Download: ML12045A255 (550)

Text

ATTACHMENT 4UPDATED REVISION TO NINE MILE POINT UNIT 1 AND UNIT 2EMERGENCY ACTION LEVELSNine Mile Point Nuclear Station, LLCJanuary 13, 2012 THIS PAGE IS ANOVERSIZED DRAWING ORFIGURE,THAT CAN BE VIEWED AT THERECORD TITLED:UPDATED REVISION TOEPIP-EP-001, "ATTACHMENT 1 EALMATRIX UNIT 1" PAGES 1 & 2.WITHIN THIS PACKAGED-05 THROUGH D-06 THIS PAGE IS ANOVERSIZED DRAWING ORFIGURE,THAT CAN BE VIEWED AT THERECORD TITLED:UPDATED REVISION TOEPIP-EP-002, "ATTACHMENT 1 EALMATRIX UNIT 2" PAGES 1 & 2.WITHIN THIS PACKAGED-07 THROUGH D-08 ATTACHMENT 5UPDATED REVISION TO NINE MILE POINT UNIT 1 AND UNIT 2EMERGENCY ACTION LEVEL BASIS DOCUMENTSNine Mile Point Nuclear Station, LLCJanuary 13, 2012

'NINE MILE POINT NUCLEAR STATIONEMERGENCY PLAN MAINTENANCE PROCEDUREEPMP-EPP-01 01REVISION 00 (Draft RAI 1 12)UNIT 1 EMERGENCY CLASSIFICATION TECHNICAL BASESTECHNICAL SPECIFICATION REQUIREDApproved by:J. KaminskiDirector Emergency PlanningTHIS IS A COMPLETE REVISIONDateEffective Date:PERIODIC REVIEW DUE DATE:*1 -.v Page No.Coversheet.i ..ii .iii .iv ....1 ....2 ....3 ....4 ....5 ....6 ....7....8 ....9 ....10 ....11 ....12 ....13 ....14 ....15 ....16 ....Chan~ge No.LIST OF EFFECTIVE PAGESPawe No. Change No.17 ....18 ....19 ....20 ....21 ....22 ....23 ....24 ....25 ....26 ....27 ....28 ....29 ....30 ....31 ....32 ....33 ....34 ....35 ....36 ....37 ....Pagqe No. Change No.38 ....39 ....40 ....41 ....42 ....43 ....44 ....45 ....46 ....47 ....48 ....49 ....50 ....51 ....52 ....53 ....54 ....55 ....56 ....57 ....58 ....Page iEPMP-EPP-0101Rev 00 Draft A Paqe No.59 ....60 ....61 ....62 ....63 ....64 ....65 ....66 ....67 ....68 ....69 ....70 ....71 ....72 ....73 ....74 ....75 ....76 ....77 ....78 ....79 ....Chanqe No.LIST OF EFFECTIVE PAGES (Cont)Page No. Chan-ge No.80 ....81 ....82 ....83 ....84 ....85 ....86 ....87 ....88 ....89 ....90 ....91 ...:92 ....93 ....94 ....95 ....96 ....97 ....98 ....99 ....100 ....Pa~ge No. Change No.101 ....102 ....103 ....104 ....105 ....106 ....107 ....108 ....109 ....110 ....111 ....112 ....113 ....114 ....115 ....116 ....117 ....118 ....119 ....120 ....121 ....Page iiEPMP-EPP-0101Rev 00 Draft A Page No.122 ....123 ....124 ....125 ....126 ....127 ....128 ....129 ....130 ....131 ....132 ....133 ....134 ....135 ....136 ....137 ....138 ....139 ....140 ....141 ....Change No.LIST OF EFFECTIVE PAGES (Cont)Page No. Change No.142 ....143 ....144 ....Page No. Change No.Page iiiEPMP-EPP-0101Rev 00 Draft A SECTION1.02.03.04.05.06.06.1Table of ContentsTITLE PAGEPURPOSE ..............................................................................................DISCUSSIO N .........................................................................................2.1 Background .................................................................................2.2 Fission Product Barriers .........................................................2.3 Emergency Classification Based on Fission ProductBarrier Degradation .....................................................................2.4 EAL Relationship to EO Ps ...........................................................2.5 Sym ptom-Based vs. Event-Based Approach ...............................2.6 EAL Organization ........................................................................2.7 Technical Bases Inform ation .......................................................2.8 Operating Mode Applicability .......................................................2.9 Validation of Indications, Reports and Conditions .......................2.10 Planned vs. UNPLANNED Events ..............................................2.11 Classifying Transient Events .......................................................2.12 Multiple Simultaneous Events and IMMINENT EAL Thresholds..2.13 Emergency Classification Level DowngradingREFERENCES .......................................................................................3.1 Developm ental .............................................................................3.2 Im plem enting ...............................................................................3.3 Com m itm ents ..............................................................................DEFINITIO NS .........................................................................................NMP1-TO-NEI 99-01 EAL CROSSREFERENCE ..................................ATTACHM ENTS ....................................................................................Attachment I -Emergency Action Level Technical Bases ....................Category R Abnormal Rad Release / Rad Effluent ..............................RG 1.2 ...............................................................................RG 1.3 ...............................................................................R S 1 .1 ................................................................................RS1.2 ................................................................................R S 1 .3 ................................................................................RA1.1 ................................................................................RA1.2 ................................................................................RA1.3 ................................................................................R U 1 .1 ............................................................. ..................RU1.2 ................................................................................R U 1 .3 ................................................................................Page ivEPMP-EPP-01 01Rev xx Draft F Table of ContentsSECTION TITLECategory RCategory HPAGE(cont'd)R A 2 .1 ................................................................................R A 2 .2 ................................................................................R U 2 .1 ................................................................................R U 2 .2 ................................................................................R A 3 ................................................................................Hazards and Other Conditions Affecting Plant Safety ......H A 1 .1 ...............................................................................H A 1 .2 ...............................................................................H A 1 .3 ...............................................................................H A 1 .4 ...............................................................................H A 1 .5 ...............................................................................H A 1 .6 ...............................................................................H U I .1 ...............................................................................H U 1 .2 ...............................................................................H U 1 .3 ...............................................................................H U 1 .4 ..................................................................... .........HU1.5 ..................................................... .........H A 2 .1 ...............................................................................H U 2 .1 ...............................................................................H U 2 .2 ...............................................................................H A 3 ...............................................................................H U 3 ...............................................................................H U 3 .2 ...............................................................................H G 4 .1 ..............................................................................H G 4 .2 ..............................................................................H S 4 .1 ...............................................................................H A 4 .1 ...............................................................................H U 4 .1 ...............................................................................H S 5 .1 ...............................................................................H A 5 .1 ...............................................................................H G 6 .1 ........................................... ..................................H S 6 .1 ...............................................................................H A 6 ...............................................................................H U 6 .1 ...............................................................................Page vEPMP-EPP-0101Rev xx Draft F SECTION TITLECategory ECategory CCategory STable of ContentsPAGEISFSI .................................... I ......................................E U 1 .1 ................................................................................Cold Shutdown / Refueling System Malfunction ...............C A I .1 ................................................................................C U 1 .1 ................................................................................C U 2 .1 ................................................................................CG3.1 .............................................CG3.2 ..............................................CS3.1 .......................................................C S 3 .2 ...............................................................................C S 3 .3 ...............................................................................C A 3 .1 ................................................................................CU3.1 ................................................................................C U 3 .2 ................................................................................C U 3 .3 ................................................................................C A 4 .1 ...............................................................................C U 4 .1 ...............................................................................C U 4 .2 ...............................................................................C U 5 .1 ...............................................................................C U 6 .1 ...............................................................................System Malfunction ..........................................................S G 1 .1 ...............................................................................S S 1 .1 ...............................................................................S A 1 .1 ...............................................................................S U 1 .1 ...............................................................................SS2.1 ................................................................. ....S G 3 .1 ..............................................................................S S 3 .1 ..............................................................................S A 3 .1 ..............................................................................S U 3 .1 ..............................................................................S U 4 .1 ..............................................................................S S 5 .1 ..............................................................................S A 5 .1 ...............................................................................S U 5 .1 ..............................................................................S U 6 .1 ..............................................................................S U 7 .1 ..............................................................................Page viEPMP-EPP-0101Rev xx Draft F SECTION6.2Table of ContentsTITLE P4Category S (cont'd)S U 7 .2 ...............................................................................S U 8 .1 ...............................................................................Category F Fission Product Barrier Degradation ................................F G I .1 ...............................................................................F S 1 .1 ...............................................................................F A I .1 ...............................................................................F U 1 .1 ...............................................................................Attachment 2 -Fission Product Barrier Loss / Potential LossMatrix and Bases ...................................................................................FC Loss A.1 .....................................................................FC Loss D.2 .....................................................................FC Loss D.3 .....................................................................FC Loss EA .....................................................................FC Potential Loss A.1 .......................................................FC Potential Loss E.2 ......................................................RCS Loss A.1 ..................................................................RCS Loss B.2 ..................................................................RCS Loss C.3 ..................................................................RCS Loss C0 ..................................................................RCS Loss D.5 ..................................................................RCS Loss E.6 ..................................................................RCS Potential Loss C0 ...................................................RCS Potential Loss C.2 ............................RCS Potential Loss E.3 ...................................................PC Loss B.1 .....................................................................PC Loss B.2 .....................................................................PC Loss C.3 .....................................................................PC Loss C0 .....................................................................PC Loss C.5 .....................................................................PC Loss E.6 .....................................................................PC Potential Loss A.1 ......................................................PC Potential Loss B.2 ......................................................PC Potential Loss B.3 ......................................................PC Potential Loss BA ......................................................PC Potential Loss D.5 ......................................................Page vii EPMP-EfWGE,P-0101Rev xx Draft F Table of ContentsSECTION6.2TITLE PAGEAttachment 2 (cont'd)PC Potential Loss E.6 ......................................................Page viiiEPMP-EPP-0101Rev xx Draft F ABBEVIATIONS / ACRONYMSAC .................................................................................................................. Alternating CurrentAPRM ............................................................................................. Average Power Range MeterATW S ................................................................................. Anticipated Transient W ithout ScramBLDG ............................................................................................................................... BuildingBW R .......................................................................................................... Boiling W ater ReactorODE .................................................................................................. Com m itted Dose EquivalentCFR ................................................................................................. Code of Federal RegulationsDC .......................................................................................................................... Direct CurrentEAL ......................................................................................................... Em ergency Action LevelEC ............................................................................................................ Em ergency CondenserECCS ........................................................................................ Em ergency Core Cooling SystemED ................................................................................................................. Em ergency DirectorEL .................................................................................................................................. ElevationEO F .............................................................................................. Em ergency O perations FacilityEOP ......................................... Emergency Operating ProcedureEPA .......................................................................................... Environm ental Protection AgencyEPM P .......................................................................... Em ergency Plan Maintenance ProcedureEPRI ......................................................................................... Electric Power Research InstituteEPIP ........................................................................... Emergency Plan Im plementing ProcedureFBI ............................................................................................... Federal Bureau of InvestigationG E ................................................................................................................ General EmergencyHCTL ........................................................................................ Heat Capacity Tem perature Lim itHOO ............................................................................... Headquarters (NRC) Operations OfficerIC ..................................................................................................................... Initiating ConditionISFSI ......... ....................... ............. INDEPENDENT SPENT FUEL STORAGE INSTALLATIONJAFNPP ...................................................................... Jam es A. FitzPatrick Nuclear Power PlantLCO ............................................................................................. Lim iting Condition of OperationLOCA ..................................................................................................... Loss of Coolant Accidentm R .......................................................................................................................... m illiRoentgenM SCRW L ................................................................... M inim um Steam Cooling RPV W ater LevelM SIV .................................................................................................. Main Steam Isolation ValveM SL ................................................................................................................... Main Steam LineNEI ......................................................................................................... Nuclear Energy InstituteNESP ............................................................................... National Environm ental Studies ProjectNRC .......................................................................................... Nuclear Regulatory Com m issionNO RA D ............................................................... North Am erican Aerospace Defense Com m andNUMARC .............................................................. Nuclear Managem ent and Resources CouncilO BE .................................................................................................. O perating Basis EarthquakePage ix EPMP-EPP-0101Rev xx Draft F ACRONYMS & ABBREVIATIONS (continued)OCA .......................................................................................................... Owner Controlled AreaO DCM ....................................................................................... Off-site Dose Calculation ManualOG ESM .......................................................................................... Offgas Effl uent Stack MonitorO RO ............................................................................................ Off-site Response OrganizationPAG ................................................................................................... Protective Action GuidelinePC ............................................................................................................... Prim ary Containm entPSIG ........................................................................................... Pounds per Square Inch GaugeR ................................................................................................................................... RoentgenRB ...................................................................................................................... Reactor BuildingRCS ....................................................................................................... Reactor Coolant SystemRem ..................................................................................................... Roentgen Equivalent ManRPS .................................................................................................... Reactor Protection SystemRPV ....................................................................................................... Reactor Pressure VesselRW ......................................................... Raw WaterRW CU ..................................................................................................... Reactor W ater CleanupSAE ............................................................................................................ Site Area Em ergencySBO .................................................................................................................... Station BlackoutSPDS ....................................................................................... Safety Param eter Display SystemTEDE ........................................................................................... Total Effective Dose EquivalentTSC ...................................................................................................... Technical Support CenterUE ......................................................................................................................... Unusual EventUSAR .......................................................................................... Updated Safety Analysis ReportPage x EPMP-EPP-0101Rev xx Draft F 1.0 PURPOSEThis document provides an explanation and rationale for each Emergency Action Level(EAL) included in the EAL Upgrade Project for Nine Mile Point Nuclear Station Unit 1(NMP1). It should be used to facilitate review of the NMP1 EALs and provide historicaldocumentation for future reference. Decision-makers responsible for implementation ofEPIP-EPP-01, "Classification of Emergency Conditions at Unit 1," and the EmergencyAction Level Matrices, may use this document as a technical reference in support of EALinterpretation. This information may assist the Emergency Director in makingclassifications, particularly those involving judgment or multiple events. The basisinformation may also be useful in training, for explaining event classifications to offsiteofficials, and facilitates regulatory review and approval of the classification scheme.The expectation is that emergency classifications are to be made as soon as conditionsare present and recognizable for the classification, but within 15 minutes in all cases ofconditions present. Use of this document for assistance is not intended to delay theemergency classification.2.0 DISCUSSION2.1 BackgroundEALs are the plant-specific indications, conditions or instrument readings that are utilizedto classify emergency conditions defined in the Nine Mile Point Site Emergency Plan.In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development ofEmergency Action Levels" as an alternative to NUREG-0654 EAL guidance.NEI 99-01 (NUMARC/NESP-007) Revision 4 was subsequently issued for industryimplementation. Enhancements over earlier revisions included:Consolidating the system malfunction initiating conditions and example emergencyaction levels which address conditions that may be postulated to occur during plantshutdown conditions.Page 1 EPMP-EPP-01 01Rev 00 Draft A

• Initiating conditions and example emergency action levels that fully addressconditions that may be postulated to occur at permanently Defueled Stations andINDEPENDENT SPENT FUEL STORAGE INSTALLATIONs (ISFSls)." Simplifying the fission product barrier EAL threshold for a Site Area Emergency.Subsequently, Revision 5 of NEI 99-01 has been issued which incorporates resolutions tonumerous implementation issues including the NRC EAL FAQs. Using NEI 99-01 Revision5 Final, February 2008 (ADAMS Accession Number ML080450149), NMP1 conducted anEAL implementation upgrade project that produced the EALs discussed herein.2.2 Fission Product BarriersMany of the EALs derived from the NEI methodology are fission product barrier based.That is, the conditions that define the EALs are based upon loss or potential loss of one ormore of the three fission product barriers. "Loss" and "Potential Loss" signify the relativedamage and threat of damage to the barrier. "Loss" means the barrier no longer assurescontainment of radioactive materials; "potential loss" implies an increased probability ofbarrier loss and decreased certainty of maintaining the barrier.The primary fission product barriers are:A. Fuel Clad (FC): Zirconium tubes which house the ceramic uranium oxide pelletsalong with the end plugs which are welded into each end of the fuel rods comprisethe FC barrier.B. Reactor Coolant System (RCS): The reactor vessel shell, vessel head, CRDhousings, vessel nozzles and penetrations, and all primary systems directlyconnected to the RPV up to the outermost Primary Containment isolation valvecomprise the RCS barrier.C. Containment (PC): The drywell, the torus, their respective interconnecting paths,and other connections up to and including the outermost containment isolationvalves comprise the Primary Containment barrier.2.3 Emergency Classification Based on Fission Product Barrier DegradationThe following criteria are the bases for event classification related to fission product barrierloss or potential loss:Unusual Event:Any loss or any potential loss of ContainmentPage 2 EPMP-EPP-0101Rev 00 Draft A Alert:Any loss or any potential loss of either Fuel Clad or RCSSite Area Emergency:Loss or potential loss of any two barriersGeneral Emergency:Loss of any two barriers and loss or potential loss of third barrier2.4 EAL Relationship to EOPsWhere possible, the EALs have been made consistent with and utilize the conditionsdefined in the NMP1 Emergency Operating Procedures (EOPs). While the symptoms thatdrive operator actions specified in the EOPs are not indicative of all possible conditionswhich warrant emergency classification, they define the symptoms, independent ofinitiating events, for which reactor plant safety and/or fission product barrier integrity arethreatened. When these symptoms are clearly representative of one of the NEI InitiatingConditions, they have been utilized as an EAL. This permits rapid classification ofemergency situations based on plant conditions without the need for additional evaluationor event diagnosis. Although some of the EALs presented here are based on conditionsdefined in the EOPs, classification of emergencies using these EALs is not dependentupon EOP entry or execution. The EALs can be utilized independently or in conjunctionwith the EOPs.2.5 Symptom-Based vs. Event-Based ApproachTo the extent possible, the EALs are symptom-based. That is, the action level threshold isdefined by values of key plant operating parameters that identify emergency or potentialemergency conditions. This approach is appropriate because it allows the full scope ofvariations in the types of events to be classified as emergencies. However, a purelysymptom-based approach is not sufficient to address all events for which emergencyclassification is appropriate. Particular events to which no predetermined symptoms can beascribed have also been utilized as EALs since they may be indicative of potentially moreserious conditions not yet fully realized.Page 3 EPMP-EPP-0101Rev 00 Draft A 2.6 EAL OrganizationThe NMP1 EAL scheme includes the following features:" Division of the EAL set into three broad groups:o EALs applicable under all plant operating modes -This group would bereviewed by the EAL-user any time emergency classification is considered.o EALs applicable only under hot operating modes -This group would only bereviewed by the EAL-user when the plant is in Hot Shutdown or PowerOperation mode.o EALs applicable only under cold operating modes -This group would only bereviewed by the EAL-user when the plant is in Cold Shutdown, Refuel orDefueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant isin a cold condition and avoid review of cold condition EALs when the plant is in ahot condition. This approach significantly minimizes the total number of EALs thatmust be reviewed by the EAL-user for a given plant condition, reduces EAL-userreading burden and, thereby, speeds identification of the EAL that applies to theemergency." Within each of the above three groups, assignment of EALs tocategories/subcategories -Category and subcategory titles are selected torepresent conditions that are operationally significant to the EAL-user.Subcategories are used as necessary to further divide the EALs of a category intological sets of possible emergency classification thresholds. The NMP1 EALcategories/subcategories and their relationship to NEI 99-01 Rev. 5 RecognitionCategories are listed below.Page 4 EPMP-EPP-0101Rev 00 Draft A EAL Groups, Categories and SubcategoriesEAL Group/Category EAL SubcategoryAny Operating Mode:R -Abnormal Radiation Levels / 1 -Offsite Rad ConditionsRadiological Effluents 2 -Onsite Rad Conditions & Spent Fuel Events3 -CR/CAS RadH -Hazards and Other Conditions 1 -Natural or Destructive PhenomenaAffecting Plant Safety 2 -FIRE or EXPLOSION3 -Hazardous Gas4 -Security5 -Control Room Evacuation6 -JudgmentE -ISFSI NoneCold Conditions:C -Cold Shutdown I Refueling System 1 -Loss of AC PowerMalfunction 2 -Loss of DC Power3 -RPV Water Level4 -RCS Temperature5 -Inadvertent Criticality6 -CommunicationsHot Conditions:S -System Malfunction 1 -Loss of AC Power2 -Loss of DC Power3 -Criticality & RPS Failure4 -Inability to Reach or Maintain Shutdown Conditions5 -Instrumentation6 -Communications7 -Fuel Clad Degradation8 -RCS LeakageF -Fission Product Barrier Degradation NonePage 5 EPMP-EPP-0101Rev 00 Draft A The primary tool for determining the emergency classification level is the EALClassification Matrix. The user of the EAL Classification Matrix may (but is not required to)consult the EAL Technical Bases Document in order to obtain additional informationconcerning the EALs under classification consideration. The user should consult Sections2.7 and 2.8, and Attachments 1 and 2 of this document for such information.2.7 Technical Bases InformationEAL technical bases are provided in Attachment 1 for each EAL according to EAL group(Any, Hot, Cold), EAL category (R, H, E, C, S and F) and EAL subcategory. A summaryexplanation of each category and subcategory is given at the beginning of the technicalbases discussions of the EALs included in the category. For each EAL, the followinginformation is provided:Catecqory Letter & TitleSubcategory Number & TitleInitiating Condition (IC)Site-specific description of the generic IC given in NEI 99-01 Rev. 5.EAL Identifier (enclosed in rectangle)Each EAL is assigned a unique identifier to support accurate communication of theemergency classification to onsite and offsite personnel. Four characters define eachEAL identifier:1. First character (letter): Corresponds to the EAL category as described above (R,H, E, C, S or F)2. Second character (letter): The emergency classification (G, S, A or U)G = General EmergencyS = Site Area EmergencyA = AlertU = Unusual Event3. Third character (number): Subcategory number within the given category.Subcategories are sequentially numbered beginning with the number one (1). IfPage 6 EPMP-EPP-0101Rev 00 Draft A a category does not have a subcategory, this character is assigned the numberone (1).4. Fourth character (number): The numerical sequence of the EAL within the EALsubcategory. If the subcategory has only one EAL, it is given the number one(1).Classification (enclosed in rectangle):Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)EAL (enclosed in rectangle)Wording enclosed in the rectangle appears as it is displayed in the EAL ClassificationMatrix. Selected terms are highlighted for emphasis:" Bold, uppercase print is assigned to: "ANY," EAL identifiers, and logic termssuch as AND, OR, EITHER, etc. (When used as conjunctions, the words "and"and "or" are not highlighted.)° Bold, mixed case print is assigned to: "all," "only," "both," table titles andcolumn headings, numbers following the word "ANY," and negative terms (e.g.,"not," "cannot," etc.)° Uppercase print is assigned to acronyms, abbreviations, and terms defined inSection 4.0.Mode ApplicabilityOne or more of the following plant operating conditions comprise the mode to whicheach EAL is applicable: 1 -Power Operation, 2 -Hot Shutdown, 3 -Cold Shutdown, 4 -Refuel, D -Defueled, or All. (See Section 2.8 for operating mode definitions.)Basis:A Plant-Specific basis section provides NMP1-relevant information concerning the EAL.This is followed by a Generic basis section that provides a description of the rationalefor the EAL as provided in NEI 99-01 Rev. 5.NMP1 Basis Reference(s):Site-specific source documentation from which the EAL is derived2.8 Operating Mode Applicability (Technical Specifications Definitions 1.1)Page 7 EPMP-EPP-0101Rev 00 Draft A 1 Power Operation* Reactor mode switch is in startup or run position.* Reactor is critical or criticality is possible due to control rod withdrawal.2 Hot Shutdown" The reactor mode switch is in the shutdown position.* No core alterations leading to an addition of reactivity are being performed." Reactor coolant temperature is greater than 212'F.3 Cold Shutdown" The reactor mode switch is in the shutdown position or refuel position.* No Core alterations leading to an addition of reactivity are being performed..* Reactor coolant temperature is less than or equal to 212'F.4 Refuel" The reactor mode switch is in the refuel position.* The reactor coolant temperature is less than 212°F." Fuel may be loaded or unloaded." No more than one operable control rod may be withdrawn.D DefueledNo fuel is in the reactor. (Note: this is equivalent to the technical specificationdefined condition "Major Maintenance")The plant operating mode that exists at the time that the event occurs (prior to anyprotective system or operator action is initiated in response to the condition) should becompared to the mode applicability of the EALs. If a lower or higher plant operating modeis reached before the emergency classification is made, the declaration shall be based onthe mode that existed at the time the event occurred.2.9 Validation of Indications, Reports and ConditionsAll emergency classifications shall be based upon VALID indications, reports or conditions.An indication, report, or condition, is considered to be VALID when it is verified by (1) aninstrument channel check, or (2) indications on related or redundant indicators, or (3) bydirect observation by plant personnel, such that doubt related to the indicator's operability,Page 8 EPMP-EPP-0101Rev 00 Draft A the condition's existence, or the report's accuracy is removed. Implicit in this definition isthe need for timely assessment.2.10 Planned vs. UNPLANNED EventsPlanned evolutions involve preplanning to address the limitations imposed by thecondition, the performance of required surveillance testing, and the implementation ofspecific controls prior to knowingly entering the condition in accordance with the specificrequirements of the site's Technical Specifications. Activities which cause the site tooperate beyond that allowed by the site's Technical Specifications, planned orUNPLANNED, may result in an EAL threshold being met or exceeded. Planned evolutionsto test, manipulate, repair, perform maintenance or modifications to systems andequipment that result in an EAL value being met or exceeded are not subject toclassification and activation requirements as long as the evolution proceeds as plannedand is within the operational limitations imposed by the specific operating license.However, these conditions may be subject to the reporting requirements of 10 CFR 50.72.2.11 Classifying Transient EventsFor some events, the condition may be corrected before a declaration has been made.The key consideration in this situation is to determine whether or not further plant damageoccurred while the corrective actions were being taken. In some situations, this can bereadily determined, in other situations, further analyses may be necessary (e.g., coolantradiochemistry following an ATWS event, plant structural examination following anearthquake, etc.). Classify the event as indicated and terminate the emergency onceassessment shows that there were no consequences from the event and other terminationcriteria are met.Existing guidance for classifying transient events addresses the period of time of eventrecognition and classification (15 minutes). However, in cases when EAL declarationcriteria may be met momentarily during the normal expected response of the plant,declaration requirements should not be considered to be met when the conditions are apart of the designed plant response, or result from appropriate Operator actions.Page 9 EPMP-EPP-01 01Rev 00 Draft A There may be cases in which a plant condition that exceeded an EAL was not recognizedat the time of occurrence but is identified well after the condition has occurred (e.g., as aresult of routine log or record review), and the condition no longer exists. In these cases,an emergency should not be declared. Reporting requirements of 10 CFR 50.72 areapplicable and the guidance of NUREG-1022, Event Reporting Guidelines 10 CFR 50.72and 50.73, should be applied.2.12 Multiple Simultaneous Events and IMMINENT EAL ThresholdsWhen multiple simultaneous events occur, the emergency classification level is based onthe highest EAL reached. For example, two Alerts remain in the Alert category. Or, an Alertand a Site Area Emergency is a Site Area Emergency. Further guidance is provided inRIS 2007-02, Clarification of NRC Guidance for Emergency Notifications During QuicklyChanging Events.Since NMP1 is at a multi-unit site, emergency classification level upgrading must alsoconsider the effects of a loss of a common system on more than one unit (e.g., potentialfor radioactive release from more than one core).Although the majority of the EALs provide very specific thresholds, the Emergency Director(ED) must remain alert to events or conditions that lead to the conclusion that exceedingthe EAL threshold is IMMINENT. If, in the judgment of the ED, an IMMINENT situation is athand, the classification should be made as if the threshold has been exceeded. While thisis particularly prudent at the higher emergency classes (the early classification may permitmore effective implementation of protective measures), it is nonetheless applicable to allemergency classes.2.13 Emergency Classification Level DowngradingAnother important aspect of usable EAL guidance is the consideration of what to do whenthe risk posed by an emergency is clearly decreasing. A combination approach involvingrecovery from General Emergencies and some Site Area Emergencies and terminationfrom Unusual Events, Alerts, and certain Site Area Emergencies causing no long termplant damage appears to be the best choice. Downgrading to lower emergencyclassification levels adds notifications but may have merit under certain circumstances.Page 10 EPMP-EPP-0101Rev 00 Draft A 3.0 REFERENCES3.1 Developmental3.1.1 NEI 99-01 Rev. 5 Final, Methodology for Development of EmergencyAction Levels, February 2008, ADAMS Accession Number ML080450149.3.1.2 NRC Regulatory Issue Summary (RIS) 2003-18, Supplement 2, Use ofNuclear Energy Institute (NEI) 99-01, Methodology for Development ofEmergency Action Levels Revision 4, Dated January 2003 (December 12,2005)3.1.3 RIS 2007-02 Clarification of NRC Guidance for Emergency NotificationsDuring Quickly Changing Events3.1.4 Nine Mile Point Site Emergency Plan3.2 Implementing3.2.1 EPIP-EPP-01 Classification of Emergency Conditions at Unit 13.2.2 EAL Comparison Matrix3.3 CommitmentsNonePage 11EPMP-EPP-0101Rev 00 Draft A 4.0 DEFINITIONS (ref. 3.1.1 except as noted)AFFECTING SAFE SHUTDOWNEvent in progress has adversely affected functions that are necessary to bring the plant toand maintain it in the applicable hot or cold shutdown condition. Plant conditionapplicability is determined by Technical Specification LCOs in effect.Example 1: Event causes damage that results in entry into an LCO that requires theplant to be placed in hot shutdown. Hot shutdown is achievable, but cold shutdown isnot. This event is not "AFFECTING SAFE SHUTDOWN."Example 2: Event causes damage that results in entry into an LCO that requires theplant to be placed in cold shutdown. Hot shutdown is achievable, but cold shutdown isnot. This event is "AFFECTING SAFE SHUTDOWN."AIRLINER/LARGE AIRCRAFTAny size or type of aircraft with the potential for causing significant damage to the plant(refer to the Security Plan for a more detailed definition).BOMBRefers to an explosive device suspected of having sufficient force to damage plantsystems or structures.CIVIL DISTURBANCEA group of people violently protesting station operations or activities at the site.CONFINEMENT BOUNDARYThe barrier(s) between areas containing radioactive substances and the environment.CONTAINMENT CLOSUREThe procedurally defined actions taken to secure containment (primary or secondary) andits associated structures, systems, and components as a functional barrier to fissionproduct release under existing plant conditions.EXPLOSIONA rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energizedequipment that imparts energy of sufficient force to potentially damage permanentstructures, systems, or components.EXTORTIONAn attempt to cause an action at the station by threat of force.FIRECombustion characterized by heat and light. Sources of smoke such as slipping drive beltsor overheated electrical equipment do not constitute FIREs. Observation of flame ispreferred but is not required if large quantities of smoke and heat are observed.HOSTAGEPage 12 EPMP-EPP-0101Rev 00 Draft A A person(s) held as leverage against the station to ensure that demands will be met by thestation.HOSTILE ACTIONAn act toward NMP1 or its personnel that includes the use of violent force to destroyequipment, take HOSTAGEs, and/or intimidate the licensee to achieve an end. Thisincludes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, orother devices used to deliver destructive force. Other acts that satisfy the overall intentmay be included.HOSTILE ACTION should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on NMP1. Non-terrorism-based EALsshould be used to address such activities, (e.g., violent acts between individuals in theowner controlled area).HOSTILE FORCEOne or more individuals who are engaged in a determined assault, overtly or by stealthand deception, equipped with suitable weapons capable of killing, maiming, or causingdestruction.IMMINENTMitigation actions have been ineffective, additional actions are not expected to besuccessful, and trended information indicates that the event or condition will occur. WhereIMMINENT timeframes are specified, they shall apply.INTACTThe RCS should be considered INTACT when the RCS pressure boundary is inits normalcondition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).INTRUSIONThe act of entering without authorization. Discovery of a BOMB in a specified area isindication of INTRUSION into that area by a HOSTILE FORCE.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)A complex that is designed and constructed for the interim storage of spent nuclear fueland other radioactive materials associated with spent fuel storage.NORMAL LEVELSAs applied to radiological IC/EALs, the highest reading in the past twenty-four hoursexcluding the current peak value.NORMAL PLANT OPERATIONSActivities at the plant site associated with routine testing, maintenance, or equipmentoperations, in accordance with normal operating or administrative procedures. Entry intoabnormal or emergency operating procedures, or deviation from normal security orradiological controls posture, is a departure from NORMAL PLANT OPERATIONS.PROJECTILEPage 13 EPMP-EPP-0101Rev 00 Draft A An object directed toward NMP1 that could cause concern for its continued operability,reliability, or personnel safety.PROTECTED AREAThe area which normally encompasses all controlled areas within the securityPROTECTED AREA fence. NMP1 and NMP2 share a common PROTECTED AREAborder. NMP1 and NMP2 PROTECTED AREA boundaries are illustrated in USAR Figure1.2-1.SABOTAGEDeliberate damage, mis-alignment, or mis-operation of plant equipment with the intent torender the equipment inoperable. Equipment found tampered with or damaged due tomalicious mischief may not meet the definition of SABOTAGE until this determination ismade by security supervision.SAFETY-RELATED STRUCTUREs, SYSTEMs and COMPONENTs (as defined in1 OCFR50.2)Those structures, systems and components that are relied upon to remain functionalduring and following design basis events to assure:(1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition;(3) The capability to prevent or mitigate the consequences of accidents which couldresult in potential offsite exposures.SECURITY CONDITIONAny security event as listed in the approved security contingency plan that constitutes athreat/compromise to site security, threat/risk to site personnel, or a potential degradationto the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILEACTION.SITE BOUNDARYPer NMP2 ODCM Figure D 1.0-1, the line around the Nine Mile Point Nuclear Stationbeyond which the land is not owned, leased or otherwise controlled by the owners andoperators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear PowerPlant.STRIKE ACTIONWork stoppage within the PROTECTED AREA by a body of workers to enforcecompliance with demands made on NMP1. The STRIKE ACTION must threaten to interruptNORMAL PLANT OPERATIONS.UNISOLABLEA breach or leak that cannot be promptly isolated.UNPLANNEDPage 14 EPMP-EPP-0101Rev 00 Draft A A parameter change or an event, the reasons for which may be known or unknown, that isnot the result of an intended evolution or expected plant response to a transient.VALIDAn indication, report, or condition, is considered to be VALID when it is verified by (1) aninstrument channel check, or (2) indications on related or redundant indicators, or (3) bydirect observation by plant personnel, such that doubt related to the indicator's operability,the condition's existence, or the report's accuracy is removed. Implicit in this definition isthe need for timely assessment.VISIBLE DAMAGEDamage to equipment or structure that is readily observable without measurements,testing, or analysis. Damage is sufficient to cause concern regarding the continuedoperability or reliability of affected SAFETY-RELATED STRUCTURE, SYSTEM orCOMPONENT. Example damage includes: deformation due to heat or impact, denting,penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping,scratches) should not be included.VITAL AREAAny areas, normally within the NMP1 PROTECTED AREA, that contains equipment,systems, components, or material, the failure, destruction, or release of which coulddirectly or indirectly endanger the public health and safety by exposure to radiation.Page 15EPMP-EPP-0101Rev 00 Draft A 5.0 NMPI-TO-NEI 99-01 EAL CROSSREFERENCEThis cross-reference is provided to facilitate association and location of a NMP1 EALwithin the NEI 99-01 IC/EAL identification scheme. Further information regarding thedevelopment of the NMP1 EALs based on the NEI guidance can be found in the EALComparison Matrix.NMP1 NEI 99-01EAL IC ExampleEALRG1.2 AG1 2RG1.3 AG1 4RS1.1 AS1 1RS1.2 AS1 2RS1.3 AS1 4RA1.1 AA1 1RA1.2 AA1 2RA1.3 AA1 3RU1.1 AU1 1RU1.2 AU1 2RU1.3 AU1 3RA2.1 AA2 2RA2.2 AA2 1RU2.1 AU2 1RU2.2 AU2 2RA3.1 AA3 1HAI.1 HA1 1HA1.2 HA1 2Page 16EPMP-EPP-0101Rev 00 Draft A NMPI NEI 99-01EAL IC ExampleEALHA1.3 HAl 3HA1.4 HA1 4HA1.5 HA1 6HA1.6 HA1 5HU1.1 HU1 1HU1.2 HU1 2HU1.3 HU1 3HU1.4 HU1 4HU1.5 HU1 5HA2.1 HA2 1HU2.1 HU2 1HU2.2 HU2 2HA3.1 HA3 1HU3.1 HU3 1HU3.2 HU3 2HG4.1 HG1 1HG4.2 HG1 2HS4.1 HS4 1HA4.1 HA4 1,2HU4.1 HU4 1,2,3HS5.1 HS2 1HA5.1 HA5 1HG6.1 HG2 1HS6.1 HS3 1Page 17EPMP-EPP-0101Rev 00 Draft A NMP1 NEI 99-01EAL IC ExampleEALHA6.1 HA6 1HU6.1 HU5 1EU1.1 E-HU1 1CA1.1 CA3 1CU1.1 CU3 1CU2.1 CU7 1CG3.1 CG1 1CG3.2 CG1 2CS3.1 CS1 1CS3.2 CS1 2CS3.3 CS1 3CA3.1 CAl 1,2CU3.1 CU1 1CU3.2 CU2 1CU3.3 CU2 2CA4.1 CA4 1,2CU4.1 CU4 1CU4.2 CU4 2CU5.1 CU8 1CU6.1 CU6 1,2SG1.1 SG1 1SS1.1 SS1 1SA1.1 SA5 1SUI.1 SUl 1Page 18EPMP-EPP-0101Rev 00 Draft A NMPI NEI 99-01EAL IC ExampleEALSS2.1 SS3 1SG3.1 SG2 1SS3.1 SS2 1SA3.1 SA2 1SU3.1 SU8 1SU4.1 SU2 1SS5.1 SS6 1SA5.1 SA4 1SU5.1 SU3 1SU6.1 SU6 1,2SU7.1 SU4 2SU7.2 SU4 1SU8.1 SU5 1,2FG1.1 FG1 1FS1.1 FS1 1FA1.1 FA1 1FU1.1 FU1 1Page 19EPMP-EPP-01 01Rev 00 Draft A 6.0 ATTACHMENTS6.16.2Attachment 1, Emergency Action Level Technical BasesAttachment 2, Fission Product Barrier Loss / Potential Loss Matrix and BasisPage 20EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory R -Abnormal Rad Release / Rad EffluentEAL Group: ANY (EALs in this category are applicable toany plant condition, hot or cold.)Many EALs are based on actual or potential degradation of fission product barriersbecause of the elevated potential for offsite radioactivity release. Degradation of fissionproduct barriers though is not always apparent via non-radiological symptoms. Therefore,direct indication of elevated radiological effluents or area radiation levels are appropriatesymptoms for emergency classification.At lower levels, abnormal radioactivity releases may be indicative of a failure ofcontainment systems or precursors to more significant releases. At higher release rates,offsite radiological conditions may result which require offsite protective actions. Elevatedarea radiation levels in plant may also be indicative of the failure of containment systemsor preclude access to plant vital equipment necessary to ensure plant safety.Events of this category pertain to the following subcategories:1. Offsite Rad ConditionsDirect indication of effluent radiation monitoring systems provides a rapid assessmentmechanism to determine releases in excess of classifiable limits. Projected offsitedoses, actual offsite field measurements or measured release rates via samplingindicate doses or dose rates above classifiable limits.2. Onsite Rad Conditions & Spent Fuel EventsSustained general area radiation levels in excess of those indicating loss of control ofradioactive materials or those levels which may preclude access to vital plant areasalso warrant emergency classification.3. CR/CAS RadSustained general area radiation levels which may preclude access to areas requiringcontinuous occupancy also warrant emergency classification.Page 21 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:R -Abnormal Rad Release / Rad Effluent1 -Offsite Rad ConditionsOffsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity > 1,000 mRem TEDE or 5,000 mRem thyroidCDE for the actual or projected duration of the release using actualmeteorologyEAL:RG1.2 General EmergencyDose assessment using actual meteorology indicates doses > 1,000 mRem TEDE or5,000 mRem thyroid CDE at or beyond the SITE BOUNDARYMode Applicability:AllBasis:Plant-SpecificThe 1,000 mRem TEDE dose is set at 100% of the EPA PAG, while the 5,000 mRemthyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDEand thyroid CDE.Dose assessment is performed using EPIP-EPP-08 "Off-site Dose Assessment andProtective Action Recommendation" (ref. 1).The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 2).GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.Releases of this magnitude are associated with the failure of plant systems needed for theprotection of the public and likely involve fuel damage.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementingprocedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when theclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP1 Basis Reference(s):Page 22EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis1. EPIP-EPP-08 Off-site Dose Assessment and Protective Action Recommendation2. NMP1 ODCM Figure 5.1.3-13. NEI 99-01 IC AG1Page 23EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:R -Abnormal Rad Release / Rad Effluent1 -Offsite Rad ConditionsOffsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity > 1,000 mRem TEDE or 5,000 mRem thyroidCDE for the actual or projected duration of the release using actualmeteorologyEAL:RG1.3 General EmergencyField survey results indicate closed window dose rates > 1,000 mRem/hr expected tocontinue for _ 60 min. at or beyond the SITE BOUNDARY (Note 1)ORAnalyses of field survey samples indicate thyroid CDE > 5,000 mRem for 1 hr of inhalationat or beyond the SITE BOUNDARY (Note 1)Note 1: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition will likely exceed the applicable timeMode Applicability:AllBasis:Plant-SpecificReal time field surveys and sample analysis is performed by offsite field monitoring teamsper EPIP-EPP-07, "Downwind Radiological Monitoring ," (ref. 1) and assessed forradiological dose consequences per EPIP-EPP-08, "Off-site Dose Assessment andProtective Action Recommendation "(ref. 2).The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 3).GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.Releases of this magnitude are associated with the failure of plant systems needed for theprotection of the public and likely involve fuel damage.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted. For thisreason, emergency implementing procedures should call for the timely performance of doseassessments using actual meteorology and release information. If the results of these dosePage 24EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical Basisassessments are available when the classification is made (e.g., initiated at a lower classificationlevel), the dose assessment results override the monitor reading EAL.NMP1 Basis Reference(s):1. EPIP-EPP-07 Downwind Radiological Monitoring2. EPIP-EPP-08 Off-site Dose Assessment and Protective Action Recommendation3. NMP1 ODCM Figure 5.1.3-14. NEI 99-01 IC AG1Page 25EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:R -Abnormal Rad Release / Rad Effluent1 -Offsite Rad ConditionsOffsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity exceeds 100 mRem TEDE or 500 mRemthyroid CDE for the actual or projected duration of the releaseusing actual meteorologyEAL:RSI.1 Site Area EmergencyANY monitor reading > Table R-1 "SAE" column for >_ 15 min. (Note 1)" Do not delay declaration awaiting dose assessment results* If dose assessment results are available, declaration should be based on doseassessment instead of radiation monitor values (see EAL RS1.2)Note 1: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition will likely exceed the applicable timeTable R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousStack (RN 1OA/B) N/A N/A 3.0E4 cps 300 cpsEC Vent N/A 300 mRem/hr 30 mRem/hr 10mRem/hrLiquidSW Effluent N/A N/A 90,000 cpm 900 cpmRW Discharge N/A N/A 200 x batch 2 x batchMode Applicability:AllBasis:Plant-SpecificThe SAE value for the EC Vent is based on the boundary dose resulting from an actual orIMMINENT release of gaseous radioactivity that exceeds 100 mRem whole body or 500mRem child thyroid for the actual or projected duration of the release (ref. 1). The 100mRem integrated dose is based on 10% of the PAG level for TEDE whole body dose. The500 mRem integrated child thyroid dose was established in consideration of the 1:5 ratio ofthe EPA Protective Action Guidelines for whole body thyroid.Page 26EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisLiquid effluent radiation monitors are not addressed in Table R-1 at the Site AreaEmergency and General Emergency levels because the dose assessment code used tocalculate these Table R-1 readings only considers a release through the Radwaste/RBVent or the Main Stack.Certain gaseous effluent radiation monitor readings are "N/A" at the Site Area Emergencyand General Emergency levels because the dose rate corresponding to the EPA PAGswould generate a reading that is beyond the upper range of the monitors.A radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude areassociated with the failure of plant systems needed for the protection of the public.The site specific monitor is the only monitor in potential release pathways that will still be on scale.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementing'procedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when theclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP1 Basis Reference(s):1. Calculation 1 H21 C003, Rev 02. NEI 99-01 IC AS1Page 27EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: Offsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity exceeds 100 mRem TEDE or 500 mRemthyroid CDE for the actual or projected duration of the releaseusing actual meteorologyEAL:RS1.2 Site Area EmergencyDose assessment using actual meteorology indicates doses > 100 mRem TEDE or500 mRem thyroid CDE at or beyond the SITE BOUNDARYMode Applicability:AllBasis:Plant-SpecificThe 100 mRem TEDE dose is set at 10% of the EPA PAG, while the 500 mRem thyroidCDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE andthyroid CDE (ref. 1).The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 2).GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude areassociated with the failure of plant systems needed for the protection of the public.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementingprocedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when theclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP1 Basis Reference(s):1. EPIP-EPP-08 Offsite Dose Assessment and Protective Action Recommendation2. NMP1 ODCM Figure 5.1.3-13. NEI 99-01 IC AS1Page 28 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:R -Abnormal Rad Release / Rad Effluent1 -Offsite Rad ConditionsOffsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity exceeds 100 mRem TEDE or 500 mRemthyroid CDE for the actual or projected duration of the releaseusing actual meteorologyEAL:RS1.3 Site Area EmergencyField survey results indicate closed window dose rates > 100 mRem/hr expected tocontinue for _ 60 min. at or beyond the SITE BOUNDARY (Note 1)ORAnalyses of field survey samples indicate thyroid CDE > 500 mRem for 1 hr of inhalation ator beyond the SITE BOUNDARY (Note 1)Note 1: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition will likely exceed the applicable timeMode Applicability:AllBasis:Plant-SpecificReal time field surveys and sample analysis is performed by offsite field monitoring teamsper EPIP-EPP-07 "Downwind Radiological Monitoring" (ref. 1) and assessed forradiological dose consequences per EPIP-EPP-08 "Off-site Dose Assessment andProtective Action Recommendation" (ref. 2).The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 3).GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude areassociated with the failure of plant systems needed for the protection of the public.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementingprocedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when thePage 29EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP1 Basis Reference(s):1. EPIP-EPP-07 Downwind Radiological Monitoring2. EPIP-EPP-08 Off-site Dose Assessment and Protective Action Recommendation3. NMP1 ODCM Figure 5.1.3-14. NEI 99-01 IC AS1Page 30EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 200 times the ODCM for 15 minutes or longerEAL:RAI.1 AlertANY gaseous monitor reading > Table R-1 "Alert" column for > 15 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Table R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousStack (RN1OA/B) N/A N/A 3.0E4 cps 300 cpsEC Vent N/A 300 mRem/hr 30 mRem/hr 10 mRem/hrLiquidSW Effluent N/A N/A 90,000 cpm 900 cpmRW Discharge N/A N/A 200 x batch 2 x batchMode Applicability:AllBasis:Plant-SpecificThe EC Vent monitor value in Table R-1 for the Alert level is not based on 200 times theODCM value (ref. 1). The method used to determine ODCM values differs from the methodused to determine the Table R-1 SAE levels and, if applied to the EC Vent monitor, wouldyield a value greater than the SAE level. Instead, a value of 30 mRem/hr has beenselected because it provides a graded escalation between the UE level and the SAE level(ref. 2).The Reactor Building Vent Monitors are not included in this EAL because the ReactorBuilding ventilation discharges to the main stack. Radioactivity release from the ReactorBuilding would, therefore, be assessed by the main stack monitor.Page 31 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisA radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses an actual or substantial potential decrease in the level of safety of the plant asindicated by a radiological release that exceeds regulatory commitments for an extended period oftime.Nuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 200 x ODCM limit multiples are specified only to distinguish between non-emergencyconditions. While these multiples obviously correspond to an off-site dose or dose rate, theemphasis in classifying these events is the degradation in the level of safety of the Releasesshould not be prorated or averaged. For example, a release exceeding 600x ODCM for'5 minutesdoes not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL is intended for sites that have established effluent monitoring on non-routine releasepathways for which a discharge permit would not normally be prepared.NMP1 Basis Reference(s):1. NMP1 Offsite Dose Calculation Manual (ODOM)2. Calculation 1H21 C003, Rev 03. N1-ARP-H1 Annunciator Hl 84. Nl-OP-50B Process Radiation Monitoring System5. NEI 99-01 IC AA1Page 32 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 200 times the ODCM for 15 minutes or longerEAL:RA1.2 AlertANY liquid monitor > Table R-1 "Alert" column for -15 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Table R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousStack (RN 1OA/B) N/A N/A 3.0E4 cps 300 cpsEC Vent N/A 300 mRem/hr 30 mRem/hr 10 mRem/hrLiquidSW Effluent N/A N/A 90,000 cpm 900 cpmRW Discharge N/A N/A 200 x batch 2 x batchMode Applicability:AllBasis:Plant-SpecificThe Containment Spray Raw Water Monitors (RW) are not included in this EAL becausethese monitors detect radiation in the discharge from their respective processes. Themonitors are located upstream of the Service Water monitor. Therefore, the Service Waterradiation monitor adequately detects offsite radioactivity releases from these systems.A radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericPage 33 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses an actual or substantial potential decrease in the level of safety of the plant asindicated by a radiological release that exceeds regulatory commitments for an extended period oftime.Nuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 200 x ODCM limit multiples are specified only to distinguish between non-emergencyconditions. While these multiples obviously correspond to an off-site dose or dose rate, theemphasis in classifying these events is the degradation in the level of safety of the plant, not themagnitude of the associated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 600x ODCM for 5minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiationmonitor readings to exceed the threshold identified in the EAL established by the radioactivitydischarge permit. This value may be associated with a planned batch release, or a continuousrelease path.NMP1 Basis Reference(s):1. NMP1 Offsite Dose Calculation Manual (ODCM)2. N1-ARP-H1 Annunciator Hl 53. Nl-OP-50B Process Radiation Monitoring System4. N1 -CSP-Q215, Service Water Alarm Setpoint Determination, Attachment 25. N1-CSP-Q308, Attachment 26. NEI 99-01 IC AA1Page 34 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 200 times the ODCM for 15 minutes or longerEAL:RA1.3 AlertConfirmed sample analyses for gaseous or liquid releases indicate concentrations orrelease rates > 200 x ODCM limits for -15 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Mode Applicability:AllBasis:Plant-SpecificReleases in excess of two hundred times the site Offsite Dose Calculation Manual(ODCM) (ref. 1) instantaneous limits that continue for 15 minutes or longer represent anuncontrolled situation and hence, a potential significant degradation in the level of safety.The final integrated dose (which is very low in the Alert emergency class) is not theprimary concern here; it is the degradation in plant control implied by the fact that therelease was not isolated within 15 minutes. Therefore, it is not intended that the release beaveraged over 15 minutes. For example, a release of 400 times the ODCM limit for 7.5minutes does not exceed this initiating condition. Further, the ED should not wait until 15minutes has elapsed, but should declare the event as soon as it is determined that therelease duration has or will likely exceed 15 minutes.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses an actual or substantial potential decrease in the level of safety of the plant asindicated by a radiological release that exceeds regulatory commitments for an extended period oftime.Page 35 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisNuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 200 x ODCM limit are specified only to distinguish between non-emergency conditions. Whilethese multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifyingthese events is the degradation in the level of safety of the plant, not the magnitude of theassociated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 600 x ODCM for5 minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses uncontrolled releases that are detected by sample analyses, particularly onunmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage.NMPI Basis Reference(s):1. NMP1 Offsite Dose Calculation Manual (ODOM)2. NEI 99-01 IC AA1Page 36EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 2 times the ODCM for 60 minutes or longerEAL:RUI.1 Unusual EventANY gaseous monitor reading > Table R-1 "UE" column for _ 60 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Table R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousStack (RN1OA/B) N/A N/A 3.0E4 cps 300 cpsEC Vent N/A 300 mRem/hr 30 mRem/hr 10 mRem/hrLiquidSW Effluent N/A N/A 90,000 cpm 900 cpmRW Discharge N/A N/A 200 x batch 2 x batchMode Applicability:AllBasis:Plant-SpecificThe Emergency Condenser (EC) Vent monitor value shown for the Table R-1 UE level istwo times the high alarm setpoint; the main Stack (OGESM, RN IOA/B) monitor value istwo times the high-high alarm setpoint. The alarm setpoints for the listed monitors areconservatively set to ensure ODCM radioactivity release limits are not exceeded (ref. 1).The Reactor Building Vent Monitors are not included in this EAL because the ReactorBuilding ventilation discharges to the main stack. Radioactivity release from the ReactorBuilding would, therefore, be assessed by the main stack monitor.Page 37EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisA radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses a potential decrease in the level of safety of the plant as indicated by aradiological release that exceeds regulatory commitments for an extended period of time.Nuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 2 x ODCM limit multiples are specified only to distinguish between non-emergency conditions.While these multiples obviously correspond to an off-site dose or dose rate, the emphasis inclassifying these events is the degradation in the level of safety of the plant, not the magnitude ofthe associated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 4x ODCM for 30minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiationmonitor readings to exceed the threshold identified in the IC.This EAL is intended for sites that have established effluent monitoring on non-routine releasepathways for which a discharge permit would not normally be prepared.NMP1 Basis Reference(s):1. NMP1 Offsite Dose Calculation Manual (ODOM)2. Calculation 1H210C003, Rev 03. N1-ARP-H1 Annunciator H1 84. Nl-OP-50B Process Radiation Monitoring System5. NEI 99-01 IC AU1Page 38 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:R -Abnormal Rad Release / Rad Effluent1 -Offsite Rad ConditionsANY release of gaseous or liquid radioactivity to the environment> 2 times the ODCM for 60 minutes or longerEAL:RU1.2 Unusual EventANY liquid monitor reading > Table R-1 "UE" column for > 60 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Table R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousStack (RN 1OA/B) N/A N/A 3.0E4 cps 300 cpsEC Vent N/A 300 mRem/hr 30 mRem/hr 10 mRem/hrLiquidSW Effluent N/A N/A 90,000 cpm 900 cpmRW Discharge N/A N/A 200 x batch 2 x batchMode Applicability:AllBasis:Plant-SpecificThe Service Water effluent monitor value shown for the UE level is two times the highalarm setpoint. The alarm setpoint is conservatively set to ensure ODCM radioactivityrelease limits are not exceeded (ref. 1).The Containment Spray Raw Water Monitors (RW) are not included in this EAL becausethese monitors detect radiation in the discharge from their respective processes. Themonitors are located upstream of the Service Water monitor. Therefore, the Service Waterradiation monitor adequately detects offsite radioactivity releases from these systems.Page 39EPMP-EPP-0101Rev 00 Draft A Attachment I -Emergency Action Level Technical BasisA radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This IC addresses a potential decrease in the level of safety of the plant as indicated by aradiological release that exceeds regulatory commitments for an extended period of time.Nuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 2 x ODCM limit multiples are specified only to distinguish between non-emergency conditions.While these multiples obviously correspond to an off-site dose or dose rate, the emphasis inclassifying these events is the degradation in the level of safety of the plant, not the magnitude ofthe associated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 4x ODCM for 30minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiationmonitor readings to exceed the threshold identified in the EAL established by the radioactivitydischarge permit. This value may be associated with a planned batch release, or a continuousrelease path.NMP1 Basis Reference(s):1. NMP1 Offsite Dose Calculation Manual (ODOM)2. N1-ARP-H1 Annunciator Hl 53. Nl-OP-50B Process Radiation Monitoring System4. Ni -CSP-Q215, Service Water Alarm Setpoint Determination, Attachment 25. N1-CSP-Q308, Attachment 26. NEI 99-01 IC AU1Page 40 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 2 times the ODCM for 60 minutes or longerEAL:RU1.3 Unusual EventConfirmed sample analyses for gaseous or liquid releases indicate concentrations orrelease rates > 2 x ODCM limits for > 60 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Mode Applicability:AllBasis:Plant-SpecificReleases in excess of two times the site Offsite Dose Calculation Manual (ODCM) (ref. 1)instantaneous limits that continue for 60 minutes or longer represent an uncontrolledsituation and hence, a potential degradation in the level of safety. The final integrated dose(which is very low in the Unusual Event emergency class) is not the primary concern here;it is the degradation in plant control implied by the fact that the release was not isolatedwithin 60 minutes. Therefore, it is not intended that the release be averaged over 60minutes. For example, a release of 4 times the ODCM limit for 30 minutes does not exceedthis initiating condition. Further, the ED should not wait until 60 minutes has elapsed, butshould declare the event as soon as it is determined that the release duration has or willlikely exceed 60 minutes.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses a potential decrease in the level of safety of the plant as indicated by aradiological release that exceeds regulatory commitments for an extended period of time.Page 41 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisNuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 2 x ODCM limit multiples are specified only to distinguish between non-emergency conditions.While these multiples obviously correspond to an off-site dose or dose rate, the emphasis inclassifying these events is the degradation in the level of safety of the plant, not the magnitude ofthe associated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 4x ODCM for 30minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses uncontrolled releases that are detected by sample analyses, particularly onunmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakagein lake water systems, etc.NMP1 Basis Reference(s):1. NMP1 Off-Site Dose Calculation Manual (ODCM)2. NEI 99-01 IC AU1Page 42EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 2 -Onsite Rad Conditions & Spent Fuel EventsInitiating Condition: Damage to irradiated fuel or loss of water level that has resulted orwill result in the uncovering of irradiated fuel outside the ReactorVesselEAL:RA2.1 AlertAlarm on ANY of the following radiation monitors due to damage to irradiated fuel or lossof water level:" ARM 18 (West end of shield wall)* ARM 25 (Rx building -east wall)" ARM 29 (Refuel bridge (LOW RANGE))" Refuel Bridge (HIGH RANGE)" Reactor Building Vent Radiation MonitorMode Applicability:AllBasis:Plant-SpecificThis EAL is defined by the specific areas where irradiated fuel is located such as therefueling cavity, RPV or Spent Fuel Pool.The bases for the area radiation high include a spent fuel handling accident and are,therefore, appropriate for this EAL (ref. 1).Elevated readings on ventilation monitors may also be indication of a radioactivity releasefrom the fuel, confirming that damage has occurred. However, elevated background at themonitor due to water level lowering may mask elevated ventilation exhaust airborne activityand needs to be considered. The Reactor Building Ventilation Radiation MonitoringSystem is used to monitor gross gamma radiation levels within the Reactor BuildingVentilation System exhaust. Two detectors located in the exhaust plenum upstream of theVentilation System isolation valves are utilized to signal off normal radiation levels and toautomatically isolate the Reactor Building Ventilation System by closing the isolationvalves and tripping the normal ventilating fans (ref. 2).Page 43 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisHowever, while radiation monitors may detect a rise in dose rate due to a drop in the waterlevel, it might not be a reliable indication of whether or not the fuel is covered. Forexample, the monitor could in fact be properly responding to a known event involvingtransfer or relocation of a source stored in or near the fuel pool or responding to a plannedevolution such as removal of the reactor head. Interpretation of these EAL thresholdsrequires some understanding of the actual radiological conditions present in the vicinity ofthe monitors.GenericThis EAL addresses increases in radiation dose rates within plant buildings, and may be aprecursor to a radioactivity release to the environment. These events represent a loss of controlover radioactive material and represent an actual or substantial potential degradation in the level ofsafety of the plant.This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage.Increased ventilation monitor readings may be indication of a radioactivity release from the fuel,confirming that damage has occurred. Increased background at the ventilation monitor due towater level decrease may mask increased ventilation exhaust airborne activity and needs to beconsidered.While a radiation monitor could detect an increase in dose rate due to a drop in the water level, itmight not be a reliable indication of whether or not the fuel is covered.Escalation of this emergency classification level, if appropriate, would be based on RS1.1, RS1.2,RS1.3, RG1.2 or RG1.3.NMPI1 Basis Reference(s):1. Nl-OP-50A ARM System Attachment 22. Nl-OP-50B Process Radiation Monitoring System3. NEI 99-01 IC AA2Page 44EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 2 -Onsite Rad Conditions & Spent Fuel EventsInitiating Condition: Damage to irradiated fuel or loss of water level that has resulted orwill result in the uncovering of irradiated fuel outside the ReactorVesselEAL:RA2.2 AlertA water level drop in a reactor refueling pathway that will result in irradiated fuel becominguncoveredMode Applicability:AllBasis:Plant-SpecificThe reactor cavity and Spent Fuel Pool (SFP) comprise the reactor refueling pathway (ref.1).Allowing level to decrease could result in spent fuel being uncovered, reducing spent fueldecay heat removal and creating an extremely hazardous radiation environment.There is no indication that water level in the spent fuel pool has dropped to the level of thefuel other than by visual observation by personnel on the refueling floor. N1-SOP-6.1, Lossof SFP/Rx Cavity Level/Decay Heat Removal, provides appropriate instructions to report avisual observation of irradiated fuel uncovery (ref. 2).GenericThis event represents a loss of control over radioactive material and represents an actual orsubstantial potential degradation in the level of safety of the plant.Escalation of this emergency classification level, if appropriate, would be based on RS1.1, RS1.2,RS1.3, RG1.2 or RG1.3.NMP1 Basis Reference(s):1. UFSAR Section X.J.2.12. N1-SOP-6.1 Loss of SFP/Rx Cavity Level/Decay Heat Removal3. NEI 99-01 IC AA2Page 45 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 2 -Onsite Rad Conditions & Spent Fuel EventsInitiating Condition: UNPLANNED rise in plant radiation levelsEAL:RU2.1 Unusual EventUNPLANNED water level drop in a reactor refueling pathway as indicated by inability torestore and maintain SFP level > low water level alarm (Note 3)ANDArea radiation monitor reading rise on ANY of the following:* ARM 18 (West end of shield wall)" ARM 25 (Rx building -east wall)* ARM 29 (Refuel bridge (LOW RANGE))" Refuel Bridge (HIGH RANGE)Note 3: If loss of water level in the refueling pathway occurs while in Mode 3, 4 or D, consider classification underEALs CU3.1, CU3.2 or CU3.3Mode Applicability:AllBasis:Plant-SpecificThe reactor cavity and Spent Fuel Pool (SFP) comprise the reactor refueling pathway (ref.1).The SFP is equipped with level switch LSE-(1S77)54-26C that actuates a low level alarmat El 338' 0" (ref. 2).The definition of "... inability to restore and maintain SFP level >..." allows the operator tovisually observe the low water level condition, if possible, and to attempt water levelrestoration instructions as long as water level remains above the top of irradiated fuel.Water level restoration instructions are performed in accordance with procedure N1 -SOP-6.1, Loss of SFP/Rx Cavity Level/Decay Heat Removal (ref. 3).The listed Area radiation monitors are located in the proximity of where spent fuel may belocated and have been selected to be indicative of a decrease in radiation shielding due todecreasing refueling pathway water level (ref. 4). While a radiation monitor could detect aPage 46 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical Basisrise in dose due to a drop in the water level, it might not be a reliable indication, in and ofitself, of whether or not the fuel is uncovered. For example, the reading on an arearadiation monitor located on the refueling bridge may rise due to planned evolutions suchas head lift, or even a fuel assembly being raised in the manipulator mast. Elevatedradiation monitor indications will need to be combined with another indicator (or personnelreport) of water loss.This event escalates-to an Alert if irradiated fuel outside the RPV is uncovered.GenericThis EAL addresses increased radiation levels as a result of water level decreases aboveirradiated fuel or events that have resulted, or may result, in UNPLANNED increases in radiationdose rates within plant buildings. These radiation increases represent a loss of control overradioactive material and represent a potential degradation in the level of safety of the plant.The refueling pathway is a combination of cavities, tubes, canals and pools. While a radiationmonitor could detect an increase in dose rate due to a drop in the water level, it might not be areliable indication of whether or not the fuel is covered.For refueling events where the water level drops below the RPV flange classification would be viaEAL CU3.1, CU3.2 or CU3.3. This event escalates to an Alert per EAL RA2.1 if irradiated fueloutside the reactor vessel is uncovered. For events involving irradiated fuel in the reactor vessel,escalation would be via the Fission Product Barrier Table for events in operating modes 1 and 2.NMP1 Basis Reference(s):1. UFSAR Section X.J.2.12. N1-ARP-L1 Annunciator Ll 53. N1-SOP-6.1 Loss of SFP/Rx Cavity Level/Decay Heat Removal4. Nl-OP-50A ARM System Attachment 25. NEI 99-01 IC AU2Page 47EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisR -Radioactivity Release / Area Radiation2 -Onsite Rad Conditions & Spent Fuel EventsCategory:Subcategory:Initiating Condition: UNPLANNED rise in plant radiation levelsEAL:RU2.2 Unusual EventUNPLANNED area radiation readings rise by a factor of 1,000 over NORMAL LEVELSMode Applicability:AllBasis:Plant-SpecificAssessment of this EAL may be made with survey readings using portable instruments aswell as installed radiation monitors.GenericThis EAL addresses increased radiation levels as a result of water level decreases aboveirradiated fuel or events that have resulted, or may result, in UNPLANNED increases in radiationdose rates within plant buildings. These radiation increases represent a loss of control overradioactive material and represent a potential degradation in the level of safety of the plant.This EAL addresses increases in plant radiation levels that represent a loss of control ofradioactive material resulting in a potential degradation in the level of safety of the plant.This EAL excludes radiation level increases that result from planned activities such as use ofradiographic sources and movement of radioactive waste materials. A specific list of ARMs is notrequired as it would restrict the applicability of the threshold. The intent is to identify loss of controlof radioactive material in any monitored area.NMP1 Basis Reference(s):1. NEI 99-01 IC AU2Page 48EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: R -Abnormal Rad Release / Rad EffluentSubcategory: 3 -CR/CAS RadInitiating Condition: Rise in radiation levels within the facility that impedes operation ofsystems required to maintain plant safety functionsEAL:RA3.1 AlertDose rates > 15 mRem/hr in EITHER of the following areas requiring continuousoccupancy to maintain plant safety functions:Control RoomORCASMode Applicability:AllBasis:Plant-SpecificThe Control Room and Central Alarm Station (CAS) must be continuously occupied in allplant operating modes at NMPI.Area radiation monitor (ARM) #3 monitors radiation levels in the vicinity of the main ControlRoom. This ARM alarms at 1 mR/hr giving personnel sufficient warning of changing levels(ref. 1). There is no area radiation monitoring system at NMP1 for the CAS. Abnormalradiation levels may be initially detected by routine radiological surveys (ref. 1).It is the impaired ability to operate the plant that results in the actual or potentialdegradation of the level of safety of the plant. The cause or magnitude of the increase inradiation levels is not a concern of this EAL. The Emergency Director must consider thesource or cause of the increased radiation levels and determine if any other EALs may beinvolved. For example, a dose rate of 15 mRem/hr in the Control Room may be a problemin itself. However, the increase may also be indicative of high dose rates in the primarycontainment due to a LOCA. In the latter case, a Site Area Emergency or a GeneralEmergency may be indicated by other EAL categories.Page 49 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisThis EAL could result in declaration of an Alert at NMP1 due to a radioactivity release orradiation shine resulting from a major accident at NMP2 or JAFNPP. Such a declarationwould be appropriate if the increase impairs safe plant operation.This EAL is not intended to apply to anticipated temporary radiation increases due toplanned events (e.g., radwaste container movement, depleted resin transfers, etc.).GenericThis EAL addresses increased radiation levels that: impact continued operation in areas requiringcontinuous occupancy to maintain safe operation or to perform a safe shutdown.The cause and/or magnitude of the increase in radiation levels is not a concern of this EAL. TheEmergency Director must consider the source or cause of the increased radiation levels anddetermine if any other EAL may be involved.Areas requiring continuous occupancy include the Control Room and any other control stationsthat are staffed continuously, such as the security alarm station CAS.NMP1 Basis Reference(s):1. Ni-OP-50A ARM System2. NEI 99-01 IC AA3Page 50 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory H -Hazards and Other Conditions Affecting Plant SafetyEAL Group: ANY (EALs in this category are applicable toany plant condition, hot or cold.)Hazards are non-plant, system-related events that can directly or indirectly affect plantoperation, reactor plant safety or personnel safety.The events of this category pertain to the following subcategories:1. Natural or Destructive PhenomenaNatural events include hurricanes, earthquakes or tornados that have potential tocause plant structure or equipment damage of sufficient magnitude to threatenpersonnel or plant safety. Non-naturally occurring events that can cause damage toplant facilities and include aircraft crashes, missile impacts, etc.2. FIRE or EXPLOSIONFIREs can pose significant hazards to personnel and reactor safety. Appropriate forclassification are FIREs within the site PROTECTED AREA or which may affectoperability of equipment needed for safe shutdown3. Hazardous GasNon-naturally occurring events that can cause damage to plant facilities and includetoxic, asphyxiant, corrosive or flammable gas leaks.4. SecurityUnauthorized entry attempts into the PROTECTED AREA, BOMB threats, SABOTAGEattempts, and actual security compromises threatening loss of physical control of theplant.5. Control Room EvacuationEvents that are indicative of loss of Control Room habitability. If the Control Room mustbe evacuated, additional support for monitoring and controlling plant functions isnecessary through the emergency response facilities.Page 51 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis6. JudgmentThe EALs defined in other categories specify the predetermined symptoms or eventsthat are indicative of emergency or potential emergency conditions and thus warrantclassification. While these EALs have been developed to address the full spectrum ofpossible emergency conditions which may warrant classification and subsequentimplementation of the Emergency Plan, a provision for classification of emergenciesbased on operator/management experience and judgment is still necessary. The EALsof this category provide the Emergency Director the latitude to classify emergencyconditions consistent with the established classification criteria based upon EmergencyDirector judgment.Page 52EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HAI.1 AlertNMP-2 seismic instrumentation indicates > 0.075 gANDEarthquake confirmed by ANY of the following:* Earthquake felt in plant* JAFNPP seismic instrumentation" Control Room indication of degraded performance of systems required for thesafe shutdown of the plantMode Applicability:AllBasis:Plant-SpecificThis EAL is based on the NMP2 USAR design basis operating earthquake of 0.075g (ref.1, 2). Seismic events of this magnitude can cause damage to plant safety functions.The method of detection relies on actuation of the NMP2 seismic monitor OBE alarmconfirmed by one or more indications such as shift operators on duty in the Control Roomdetermining that the ground motion was felt, indication received from JAFNPPinstrumentation or degraded system performance.The NMP1 design basis operating earthquake is 0.1 1g. However, due to the seismicinstrumentation available at NMP1, determination of seismic activity levels beyond theSeismic Event value of 0.01 g will require evaluation of data recorded by the SeismicMonitoring Recorders. Since this could cause unnecessary delay in classification, action istaken at the lower NMP2 level which is indicated in real time by the NMP2 seismicinstrumentation.NMP1 and NMP2 share a common PROTECTED AREA border. Consideration should begiven to the opposite unit when classifying under this EAL.Page 53 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisGenericThese EALs escalate from HU1.1 in that the occurrence of the event has resulted in VISIBLEDAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or hascaused damage to the safety systems in those structures evidenced by control room indications ofdegraded system response or performance. The occurrence of VISIBLE DAMAGE and/ordegraded system response is intended to discriminate against lesser events. The initial reportshould not be interpreted as mandating a lengthy damage assessment prior to classification. Noattempt is made in this EAL to assess the actual magnitude of the damage. The significance hereis not that a particular system or structure was damaged, but rather, that the event was of sufficientmagnitude to cause this degradation.Escalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.Seismic events of this magnitude can result in a VITAL AREA being subjected to forces beyonddesign limits, and thus damage may be assumed to have occurred to plant safety systems.NMPI Basis Reference(s):1. NMP2 USAR Section 3.7A.1.12. N2-SOP-90 Natural Events3. NMP2 USAR Section 2.1.1.14. Nl-SOP-28 Seismic Event5. USAR Section I.B.13 Characteristics -Structural Design6. NEI 99-01 IC HA1Page 54EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HA1.2 AlertTornado strikingORSustained high winds > 90 mphresulting in EITHER:VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM orCOMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas* Reactor Building (including Primary Containment)* Control Room* Screenhouse* Turbine Building* Battery Rooms* Battery Board Rooms* Cable Spreading Room* Main Steam Isolation Valve Room* Diesel Generator Engine and Board Rooms* Security* Central Alarm Station* Secondary Alarm Station* Security Uninterruptible Power Supply RoomMode Applicability:AllPage 55EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisBasis:Plant-SpecificAll Category I structures are designed for a wind velocity of 125 mph. NMP2 design windspeed is 90 mph. The more limiting wind speed, therefore, has been selected for NMP1.(ref. 1,3)Weather conditions are monitored at three locations:* The 200 foot high Primary OR Main Meteorological Tower located 0.6 miles west-southwest of NMP2" The 90 foot Backup Tower located east of JAFNPP" The 30 foot Inland Tower located at the Oswego County Airport near FultonMeteorological parameters such as wind speed are sent to the Control Rooms andTechnical Support Centers (TSC) at NMP1, NMP2, JAFNPP and the EmergencyOperations Facility (EOF). Data from sensors mounted on these towers are sent to bothdigital and analog systems for display, processing and storage. Wind speed and winddirection, as well as wind speed deviation and differential temperatures are monitored inNMP1 Control Room and recorded on strip chart recorders on the G panel. (ref. 2)Wind speed can be measured up to 100 mph.Weather information may be obtained from (ref. 4):" National Weather Service: Buffalo 716-565-9001 or 800-462-7751; or Binghamton607-729-7629" Accu-Weather: 815-235-8650 or 814-237-5803The PROTECTED AREA Boundary is depicted in NMP2 USAR Figure 1.2-1, Plot Plan(ref. 5).This threshold addresses events that may have resulted in a Safe Shutdown Area beingsubjected to forces beyond design limits and thus damage may be assumed to haveoccurred to plant safety systems. Safe Shutdown Areas are areas that house equipmentthe operation of which may be needed to ensure the reactor safely reaches and ismaintained in cold shutdown. Safe Shutdown Areas include structures that contain thePage 56 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical Basisequipment of concern. The Alert classification is appropriate if relevant plant parametersindicate that the performance of safety systems in the affected Safe Shutdown Areas hasbeen degraded. No attempt should be made to fully inventory the actual magnitude of thedamage or quantify the degradation of safety system performance prior to declaration ofan Alert under this threshold.Table H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 6).NMP1 and NMP2 share a common PROTECTED AREA border. Consideration should begiven to the opposite unit when classifying under this EAL.GenericThis EAL escalates from HU1.2 in that the occurrence of the event has resulted in VISIBLEDAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or hascaused damage to the safety systems in those structures evidenced by control room indications ofdegraded system response or performance. The occurrence of VISIBLE DAMAGE and/ordegraded system response is intended to discriminate against lesser events. The initial reportshould not be interpreted as mandating a lengthy damage assessment prior to classification. Noattempt is made in this EAL to assess the actual magnitude of the damage. The significance hereis not that a particular system or structure was damaged, but rather, that the event was of sufficientmagnitude to cause this degradation.Escalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.This EAL is based on a tornado striking (touching down) or high winds that have caused VISIBLEDAMAGE to structures containing functions or systems required for safe shutdown of the plant.NMPI Basis Reference(s):1. USAR Section VI.C.1.1 Wind and Snow Loadings2. N1-OP-64 Meteorological Monitoring3. NMP2 USAR Section 3.3.1.14. N2-SOP-64 High Winds5. NMP2 USAR Figure 1.2-16. USAR Section X7. NEI 99-01 IC HA1Page 57 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HA1.3 AlertInternal floodingresulting in EITHER:An electrical shock hazard that precludes access to operate or monitor ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas* Reactor Building (including Primary Containment)" Control Room" Screenhouse" Turbine Building* Battery Rooms* Battery Board Rooms* Cable Spreading Room* Main Steam Isolation Valve Room* Diesel Generator Engine and Board Rooms" Security* Central Alarm Station* Secondary Alarm Station* Security Uninterruptible Power Supply RoomMode Applicability:AllBasis:Plant-SpecificThis threshold addresses the affect of flooding caused by internal events such ascomponent failures, Circulating, Component Cooling or Service Water line ruptures,Page 58 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisequipment misalignment, FIRE suppression system actuation, and outage activitymishaps. The internal flooding areas contain systems that are:" Required for safe shutdown of the plant" Not designed to be wetted or submerged" Susceptible to internal flooding eventsTable H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1).GenericEscalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.This EAL addresses the effect of internal flooding caused by events such as component failures,equipment misalignment, or outage activity mishaps. It is based on the degraded performance ofsystems, or has created industrial safety hazards (e.g., electrical shock) that preclude necessaryaccess to operate or monitor safety equipment. The inability to access, operate or monitor safetyequipment represents an actual or substantial potential degradation of the level of safety of theplant.Flooding as used in this EAL describes a condition where water is entering the room faster thaninstalled equipment is capable of removal, resulting in a rise of water level within the room.Classification of this EAL should not be delayed while corrective actions are being taken to isolatethe water source.NMP1 Basis Reference(s):1. USAR Section X2. NEI 99-01 IC HA1Page 59EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HA1.4 AlertTurbine failure-generated PROJECTILEsresulting in EITHER:VISIBLE DAMAGE to or penetration of ANY SAFETY-RELATED STRUCTURE,SYSTEM or COMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas" Reactor Building (including Primary Containment)" Control Room" Screenhouse" Turbine Building* Battery Rooms* Battery Board Rooms* Cable Spreading Room* Main Steam Isolation Valve Room* Diesel Generator Engine and Board Rooms* Security* Central Alarm Station* Secondary Alarm Station* Security Uninterruptible Power Supply RoomMode Applicability:AllBasis:Plant-SpecificThe turbine generator stores large amounts of rotational kinetic energy in its rotor. In theunlikely event of a major mechanical failure, this energy may be transformed into bothPage 60EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisrotational and translational energy of rotor fragments. These fragments may impact thesurrounding stationary parts. If the energy-absorbing capability of these stationary turbinegenerator parts is insufficient, external PROJECTILEs will be released. These ejectedPROJECTILEs may impact various plant structures, including those housing safety relatedequipment.Table H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1).GenericThis EAL escalates from HU1.4 in that the occurrence of the event has resulted in VISIBLEDAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or hascaused damage to the safety systems in those structures evidenced by control room indications ofdegraded system response or performance. The occurrence of VISIBLE DAMAGE and/ordegraded system response is intended to discriminate against lesser events. The initial reportshould not be interpreted as mandating a lengthy damage assessment prior to classification. Noattempt is made in this EAL to assess the actual magnitude of the damage. The significance hereis not that a particular system or structure was damaged, but rather, that the event was of sufficientmagnitude to cause this degradation.Escalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.This EAL addresses the threat to safety related equipment imposed by PROJECTILEs generatedby main turbine rotating component failures. Therefore, this EAL is consistent with the definition ofan Alert in that the potential exists for actual or substantial potential degradation of the level ofsafety of the plant.NMP1 Basis Reference(s):1. USAR Section X2. NEI 99-01 IC HA1Page 61 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HA1.5 AlertLake water level > 254 ftORIntake water level < 236 ftMode Applicability:AllBasis:Plant-SpecificThis threshold covers high and low water level conditions that may have resulted in a plantVITAL AREA being subjected to levels beyond design limits, and thus damage may beassumed to have occurred to plant safety systems.The high lake level is based upon the maximum probable flood level (ref. 1).The low intake water level corresponds to the minimum level before damage may occur tothe service water pumps (ref. 2-6).GenericThis EAL addresses other site specific phenomena that result in VISIBLE DAMAGE to VITALAREAs or results in indication of damage to SAFETY STRUCTURES, SYSTEMS, orCOMPONENTS containing functions and systems required for safe shutdown of the plant that canalso be precursors of more serious events.NMP1 Basis Reference(s):1. USAR Section Ill-F Screenhouse, Intake and Discharge Tunnels2. USAR Section X-F Service Water System3. N1-SOP-18.1 Service Water Failure/Low Intake Level4. $13.1-100-F0035. $14-93-F0036. S16.9NPSHAM0027. NEI 99-01 IC HA1Page 62 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HA1.6 AlertVehicle crashresulting in EITHER:VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM orCOMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas" Reactor Building (including Primary Containment)* Control Room" Screenhouse" Turbine Building* Battery Rooms* Battery Board Rooms* Cable Spreading Room* Main Steam Isolation Valve Room* Diesel Generator Engine and Board Rooms" Security* Central Alarm Station* Secondary Alarm Station* Security Uninterruptible Power Supply RoomMode Applicability:AllBasis:Plant-SpecificThis EAL is intended to address crashes of vehicle types large enough to cause significantdamage to plant structures containing functions and systems required for safe shutdown ofPage 63 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisthe plant. Vehicle types include automobiles, aircraft, trucks, cranes, forklifts, waterbornecraft, etc.Table H-1 Safe Shutdown Areas include all Class I Structures and structures containingClass I equipment and systems needed for safe shutdown (ref. 1).GenericThe occurrence of VISIBLE DAMAGE and/or degraded system response is intended todiscriminate against lesser events. The initial report should not be interpreted as mandating alengthy damage assessment prior to classification. No attempt is made in this EAL to assess theactual magnitude of the damage. The significance here is not that a particular system or structurewas damaged, but rather, that the event was of sufficient magnitude to cause this degradation.Escalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.This EAL addresses vehicle crashes within the PROTECTED AREA that results in VISIBLEDAMAGE to VITAL AREAs or indication of damage to SAFETY STRUCTURES, SYSTEMS, orCOMPONENTS containing functions and systems required for safe shutdown of the plant.NMP1 Basis Reference(s):1. USAR Section X2. NEI 99-01 IC HA1Page 64EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting the PROTECTEDAREAEAL:HUI.1 Unusual EventSeismic event identified by ANY two of the following:" Annunciator H2 6 SEISMIC DETECTION EQUIPMENT EVENT indicates seismicevent detected* Confirmation of earthquake received on NMP-2 or JAFNPP seismic instrumentation" Earthquake felt in plantMode Applicability:AllBasis:Plant-SpecificThe NMPlseismic instrumentation actuates at 0.01 g causing Annunciator H2 6 to bereceived. This annunciator provides the most direct indication in the Control Room that aseismic event has occurred. Other methods are indication received from NMP-2 orJAFNPP instrumentation.Evaluation of the magnitude of the event will require evaluation of data recorded by theSeismic Monitoring Recorders.NMP1 and NMP2 share a common PROTECTED AREA border. Consideration should begiven to the opposite unit when classifying under this EAL.GenericThis EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be ofconcern to plant operators.Damage may be caused to some portions of the site, but should not affect ability of safetyfunctions to operate.As defined in the EPRI-sponsored Guidelines for Nuclear Plant Response to an Earthquake, datedOctober 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) thevibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based onPage 65 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical Basisa consensus of control room operators on duty at the time, and (b) for plants with operable seismicinstrumentation, the seismic switches of the plant are activated.NMP-1 Basis Reference(s):1. N1-ARP-H2 annunciator H2 62. Nl-SOP-28 Seismic Event3. USAR Section I.B.13 Characteristics4. NEI 99-01 IC HU1Page 66EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting the PROTECTEDAREAEAL:HU1.2 Unusual EventTornado striking within PROTECTED AREA boundaryORSustained high winds > 90 mphMode Applicability:AllBasis:Plant-SpecificAll Category 1 structures are designed for a wind velocity of 125 mph. This EAL isdeclared on a site-wide basis. NMP2 design wind speed is 90 mph. The more limiting windspeed, therefore, has been selected for NMP1. (ref. 1, 3)Weather conditions are monitored at three locations:* The 200 foot high Primary OR Main Meteorological Tower located 0.6 miles west-southwest of NMP2* The 90 foot Backup Tower located east of JAFNPP" The 30 foot Inland Tower located at the Oswego County Airport near FultonMeteorological parameters such as wind speed are sent to the Control Rooms andTechnical Support Centers (TSC) at NMP1, NMP2, JAFNPP and the EmergencyOperations Facility (EOF). Data from sensors mounted on these towers are sent to bothdigital and analog systems for display, processing and storage. Wind speed and winddirection, as well as wind speed deviation and differential temperatures are monitored inNMP1 Control Room and recorded on strip chart recorders on the G panel. (ref. 2)Wind speed can be measured up to 100 mph.Page 67 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisWeather information may be obtained from (ref. 4):* National Weather Service: Buffalo 716-565-9001 or 800-462-7751; or Binghamton607-729-7629" Accu-Weather: 815-235-8650 or 814-237-5803NMP1 and NMP2 share a common PROTECTED AREA border. Consideration should begiven to the opposite unit when classifying under this EAL.GenericThis EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be ofconcern to plant operators.This EAL is based on a tornado striking (touching down) or high winds within the PROTECTEDAREA.Escalation of this emergency classification level, if appropriate, would be based on VISIBLEDAMAGE, or by other in plant conditions, via EAL HA1.2.NMP1 Basis Reference(s):1. USAR Section VI.C.1.1 Wind and Snow Loadings2. N1-OP-64 Meteorological Monitoring3. NMP2 USAR Section 3.3.1.14. N1-SOP-64 High Winds5. NEI 99-01 IC HU1Page 68 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety1 -Natural or Destructive PhenomenaNatural or destructive phenomena affecting the PROTECTEDAREAEAL:HU1.3 Unusual EventInternal flooding that has the potential to affect ANY SAFETY-RELATED STRUCTURE,SYSTEM or COMPONENT required by Technical Specifications for the current operatingmode in ANY Table H-1 areaTable H-1 Safe Shutdown Areas" Reactor Building (including Primary Containment)" Control Room" Screenhouse* Turbine Building* Battery Rooms* Battery Board Rooms* Cable Spreading Room* Main Steam Isolation Valve Room* Diesel Generator Engine and Board Rooms* Security* Central Alarm Station* Secondary Alarm Station* Security Uninterruptible Power Supply RoomMode Applicability:AllBasis:Plant-SpecificPlant structures evaluated for impact of internal flooding in the NMP-1 Internal FloodingHazards Analysis (ref. 1) are:" Reactor Building" Turbine Building" ScreenhousePage 69EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisTable H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 2).GenericThis EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be ofconcern to plant operators.This EAL addresses the effect of internal flooding caused by events such as component failures,equipment misalignment, or outage activity mishaps.Escalation of this emergency classification level, if appropriate, would be based VISIBLE DAMAGEvia EAL HA1.3, or by other plant conditions.NMP1 Basis Reference(s):1. Calculation SO-FLOOD-FOO1 Internal Flooding Hazard Analysis2. USAR Section X3. NEI 99-01 IC HU1Page 70EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting the PROTECTEDAREAEAL:HU1.4 Unusual EventTurbine failure resulting in ANY of the following:" Casing penetration* Damage to turbine seals" Damage to generator sealsMode Applicability:AllBasis:Plant-SpecificThe turbine generator stores large amounts of rotational kinetic energy in its rotor. In theunlikely event of a major mechanical failure, this energy may be transformed into bothrotational and translational energy of rotor fragments. These fragments may impact thesurrounding stationary parts. If the energy-absorbing capability of these stationary turbinegenerator parts is insufficient, external PROJECTILEs will be released. These ejectedPROJECTILEs may impact various plant structures, including those housing safety relatedequipment.In the event of PROJECTILE ejection, the probability of a strike on a plant region is afunction of the energy and direction of an ejected PROJECTILE and of the orientation ofthe turbine with, respect to the plant region.Failure of turbine or generator seals may be indicated by a loss of seal oil pressure or lossof condenser vacuum (ref. 3, 4, 5).GenericThese EALs are categorized on the basis of the occurrence of an event of sufficient magnitude tobe of concern to plant operators.Page 71 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisThis EAL addresses main turbine rotating component failures of sufficient magnitude to causeobservable damage to the turbine casing or to the seals of the turbine generator. Generator sealdamage observed after generator purge does not meet the intent of this EAL because it did notimpact normal operation of the plant.Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases(hydrogen cooling) to the plant environs. Actual FIRES and flammable gas build up areappropriately classified via EAL HU2.1 and EAL HU3.1.This EAL is consistent with the definition of a UE while maintaining the anticipatory nature desiredand recognizing the risk to non-safety related equipment.Escalation of this emergency classification level, if appropriate, would be to EAL HA1.4 based ondamage done by PROJECTILES generated by the failure or in conjunction with a steam generatortube rupture. These latter events would be classified by the Category R EALs or Category F EALs.NMP1 Basis Reference(s):1. N1-OP-31 Tandem Compound Reheat Turbine2. N1-SOP-31.1 Turbine Trip3. N1 -ARP-A1 3-4 Condenser Vacuum Below 24" Hg4. N1-ARP-A1 4-1 Generator H2 Seal Oil Pressure Low5. N1-SOP-25.1 Unplanned Loss of Condenser Vacuum6. NEI 99-01 IC HU1Page 72EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting the PROTECTEDAREAEAL:HUI.5 Unusual EventLake water level > 248.2 ftORIntake water level < 238.8 ftMode Applicability:AllBasis:Plant-SpecificThis threshold addresses high and low bay water level conditions that could be a precursorof more serious events (ref. 1, 2).The high lake level is based upon the maximum attainable uncontrolled lake water level asspecified in the NMP 2 USAR. Dams on the St. Lawrence River, under the authority of theInternational St. Lawrence River Board of Control, are now used to regulate the lake level.The low limit is set for el 74.37 m (244 ft) on April 1 and is maintained at or above thatelevation during the entire navigation season (April 1 to November 30). The upper limit ofthe lake level is el 75.59 m (248.2 ft) (ref. 3).The low level is based on intake forebay level and corresponds to the minimum intakewater level for operability of Emergency Service Water, Emergency Diesel Generatorcooling water, Containment Spray Raw Water and Diesel and Electric FIRE Pump (ref. 4-9).During planned evolutions such as intake water gate manipulation for reverse flowoperations in which continuous monitoring of the intake level is being accomplished, entryinto this EAL would not be warranted unless UNPLANNED /unexpected conditions and/orindications occur.Page 73 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisGenericThis EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be ofconcern to plant operators.This EAL addresses other site specific phenomena that can also be precursors of more seriousevents.NMP1 Basis Reference(s):1. USAR Section III-F Screenhouse, Intake and Discharge Tunnels2. USAR Section X-F Service Water System3. NMP 2 USAR Section 2.4.11.24. N1-ARP-H2 Annunciator H2 35. N1-SOP-18.1 Service Water Failure/Low Intake Level6. $13.1-100F0037. $14-93F0038. S16.9NPSHAM0029. Calc No. S14-93-F00710. NEI 99-01 IC HU1Page 74EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 2 -FIRE or EXPLOSIONInitiating Condition: FIRE or EXPLOSION affecting the operability of plant safetysystems required to establish or maintain safe shutdownEAL:HA2.1 AlertFIRE or EXPLOSIONresulting in EITHER:VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM orCOMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-I1 Safe Shutdown Areas* Reactor Building (including Primary Containment)* Control Room* Screenhouse* Turbine Building* Battery Rooms* Battery Board Rooms* Cable Spreading Room* Main Steam Isolation Valve Room* Diesel Generator Engine and Board RoomsSecurity* Central Alarm Station* Secondary Alarm Station* Security Uninterruptible Power Supply RoomMode Applicability:AllBasis:Plant-SpecificTable H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1).Page 75 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisGenericVISIBLE DAMAGE is used to identify the magnitude of the FIRE or EXPLOSION and todiscriminate against minor FIREs and EXPLOSIONs.The reference to structures containing safety systems or components is included to discriminateagainst FIREs or EXPLOSIONs in areas having a low probability of affecting safe operation. Thesignificance here is not that a safety system was degraded but the fact that the FIRE orEXPLOSION was large enough to cause damage to these systems.The use of VISIBLE DAMAGE should not be interpreted as mandating a lengthy damageassessment prior to classification. The declaration of an Alert and the activation of the TechnicalSupport Center will provide the Emergency Director with the resources needed to perform detaileddamage assessments.The Emergency Director also needs to consider any security aspects of the EXPLOSION.Escalation of this emergency classification level, if appropriate, will be based on EALs in CategoryS, Category F or Category R.NMPI Basis Reference(s):1. USAR Section X2. NEI 99-01 IC HA2Page 76 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 2 -FIRE or EXPLOSIONInitiating Condition: FIRE within the PROTECTED AREA not extinguished within 15min. of detection or EXPLOSION within the PROTECTED AREAEAL:HU2.1 Unusual EventFIRE not extinguished within 15 min. of Control Room notification or verification of aControl Room FIRE alarm in ANY Table H-1 area, RadWaste Solidification and StorageBldg, or Security West Bldg (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table H-1 Safe Shutdown Areas" Reactor Building (including Primary Containment)" Control Room" Screenhouse" Turbine Building* Battery Rooms* Battery Board Rooms* Cable Spreading Room* Main Steam Isolation Valve Room* Diesel Generator Engine and Board Rooms" Security* Central Alarm Station* Secondary Alarm Station* Security Uninterruptible Power Supply RoomMode Applicability:AllBasis:Plant-SpecificTable H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1). The RadWaste Solidification and StorageBldg. and Security West Bldg. are included because they are immediately adjacent to onePage 77EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisor more Table H-1 areas and a FIRE within either of those buildings may potentially impactsafe shutdown equipment should the FIRE not be controlled (ref. 1).GenericThis EAL addresses the magnitude and extent of FIREs that may be potentially significantprecursors of damage to safety systems. It addresses the FIRE, and not the degradation inperformance of affected systems that may result.As used here, detection is visual observation and either report by plant personnel or sensor alarmindication.The 15 minute time period begins with a credible notification that a FIRE is occurring, or indicationof a FIRE detection system alarm/actuation. Verification of a FIRE detection systemalarm/actuation includes actions that can be taken within the control room or other nearby sitespecific location to ensure that it is not spurious. An alarm is assumed to be an indication of a FIREunless it is disproved within the 15 minute period by personnel dispatched to the scene. In otherwords, a personnel report from the scene may be used to disprove a sensor alarm if receivedwithin 15 minutes of the alarm, but shall not be required to verify the alarm.The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIREsthat are readily extinguished (e.g., smoldering waste paper basket).NMP1 Basis Reference(s):1. USAR Section X2. NEI 99-01 IC HU2Page 78EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 2 -FIRE or EXPLOSIONInitiating Condition: FIRE within the PROTECTED AREA not extinguished within 15min. of detection or EXPLOSION within the PROTECTED AREAEAL:HU2.2 Unusual EventEXPLOSION of sufficient force to damage permanent structures or equipment within thePROTECTED AREAMode Applicability:AllBasis:Plant-SpecificWhile some EXPLOSIONs may also result in FIREs that exceed EAL HU2.1, no FIRE isnecessary to declare an emergency in the event of an EXPLOSION. If a FIRE also occursas a result or with an EXPLOSION, declare the Unusual Event based on the EXPLOSIONand monitor the progress of the FIRE for potential escalation due to FIRE damage.NMP1 and NMP2 share a common PROTECTED AREA border. NMP1 and NMP2PROTECTED AREA boundaries are illustrated in NMP-2 USAR Figure 1.2-1 (ref. 1).GenericThis EAL addresses the magnitude and extent of EXPLOSIONs that may be potentially significantprecursors of damage to safety systems. It addresses the EXPLOSION, and not the degradation inperformance of affected systems that may result.This EAL addresses only those EXPLOSIONs of sufficient force to damage permanent structuresor equipment within the PROTECTED AREA.No attempt is made to assess the actual magnitude of the damage. The occurrence of theEXPLOSION is sufficient for declaration.The Emergency director also needs to consider any security aspects of the EXPLOSION, ifapplicable.Escalation of this emergency classification level, if appropriate, would be based on EAL HA2.1.NMP1 Basis Reference(s):1. NMP-2 USAR Figure 1.2-1Page 79 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical Basis2. NEI 99-01 IC HU2Page 80EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisH -Hazards and Other Conditions Affecting Plant Safety3 -Hazardous GasCategory:Subcategory:Initiating Condition: Access to a VITAL AREA is prohibited due to toxic, corrosive,asphyxiant or flammable gases which jeopardize operation ofoperable equipment required to maintain safe operations or safelyshutdown the reactorEAL:HA3.1 AlertAccess to ANY Table H-1 area is prohibited due to toxic, corrosive, asphyxiant orflammable gases which jeopardize operation of systems required to maintain safeoperations or safely shutdown the reactor (Note 5)Note 5: If the equipment in the stated area was already inoperable, or out of service, before the event occurred, thenEAL HA3.1 should not be declared as it will have no adverse impact on the ability of the plant to safelyoperate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.Table H-1 Safe Shutdown Areas" Reactor Building (including Primary Containment)" Control Room* Screenhouse" Turbine Building* Battery Rooms* Battery Board Rooms* Cable Spreading Room* Main Steam Isolation Valve Room* Diesel Generator Engine and Board Rooms" Security* Central Alarm Station* Secondary Alarm Station* Security Uninterruptible Power Supply RoomMode Applicability:AllBasis:Plant-SpecificPage 81EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisTable H-1 Safe Shutdown Areas include all Class I Structures and structures containingClass I equipment and systems needed for safe shutdown (ref. 1).For areas that contain no safety-related structure, system or component that wouldpotentially be required to be operated or for which the structure, system or component wasalready out of service or inoperable before the event, this EAL would not be applicable.For purposes of this EAL, any gas (C02 included) is considered toxic when oxygenconcentrations in the affected areas have been or could be expected to be reduced to<19.5% or toxicity of the gas will be injurious to persons inhaling it. For discharges ofHalon, NMP's systems are designed for discharge concentration from 5% up to 6.5%. Inaccordance with NFPA 12 A, Halon 1301 Fire Extinguishing Systems, exposures to levelsof up to 7% produce little if any noticeable effect (ref. 2).GenericGases in a Safe Shutdown Area can affect the ability to safely operate or safely shutdown thereactor.The fact that SCBA may be worn does not eliminate the need to declare the event.Declaration should not be delayed for confirmation from atmospheric testing if the atmosphereposes an immediate threat to life and health or an immediate threat of severe exposure to gases.This could be based upon documented analysis, indication of personal ill effects from exposure, oroperating experience with the hazards.If the equipment in the stated area was already inoperable, or out of service, before the eventoccurred, then this EAL should not be declared as it will have no adverse impact on the ability ofthe plant to safely operate or safely shutdown beyond that already allowed by TechnicalSpecifications at the time of the event.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.Most commonly, asphyxiants work by merely displacing air in an enclosed environment. Thisreduces the concentration of oxygen below the normal level of around 19%, which can lead tobreathing difficulties, unconsciousness or even death.An uncontrolled release of flammable gasses within a facility structure has the potential to affectsafe operation of the plant by limiting either operator or equipment operations due to the potentialfor ignition and resulting equipment damage/personnel injury. Flammable gasses, such ashydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repairequipment/components (acetylene -used in welding). This EAL assumes concentrations offlammable gasses which can ignite/support combustion.Escalation of this emergency classification level, if appropriate, will be based on EALs in CategoryS, Category F or Category R.Page 82 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisNMP1 Basis Reference(s):1. USAR Section X2. NFPA 12 A Halon 1301 Fire Extinguishing Systems3. NEI 99-01 IC HA3Page 83EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Hazardous GasInitiating Condition: Release of toxic, corrosive, asphyxiant or flammable gasesdeemed detrimental to NORMAL PLANT OPERATIONSEAL:HU3.1 Unusual EventToxic, corrosive, asphyxiant or flammable gases in amounts that have or could adverselyaffect NORMAL PLANT OPERATIONSMode Applicability:AllBasis:Plant-SpecificNORMAL PLANT OPERATIONS is defined to mean activities at the plant site associatedwith routine testing, maintenance, or equipment operations, in accordance with normaloperating or administrative procedures. Entry into abnormal or emergency operatingprocedures, or deviation from normal security or radiological controls posture, is adeparture from NORMAL PLANT OPERATIONS.For purposes of this EAL, any gas (C02 included) is considered toxic when oxygenconcentrations in the affected areas have been or could be expected to be reduced to<19.5% or toxicity of the gas will be injurious to persons inhaling it. For discharges ofHalon, NMP's systems are designed for discharge concentration from 5% up to 6.5%. Inaccordance with NFPA 12 A, Halon 1301 Fire Extinguishing Systems, exposures to levelsof up to 7% produce little if any noticeable effect (ref. 1).GenericThis EAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficientquantity to affect NORMAL PLANT OPERATIONS.The fact that SCBA may be worn does not eliminate the need to declare the event.This EAL is not intended to require significant assessment or quantification. It assumes anuncontrolled process that has the potential to affect plant operations. This would preclude small orincidental releases, or releases that do not impact structures needed for plant operation.Page 84 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisAn asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.Most commonly, asphyxiants work by merely displacing air in an enclosed environment. Thisreduces the concentration of oxygen below the normal level of around 19%, which can lead tobreathing difficulties, unconsciousness or even death.Escalation of this emergency classification level, if appropriate, would be based on EAL HA3.1.NMP1 Basis Reference(s):1. NFPA 12 A Halon 1301 Fire Extinguishing Systems2. NEI 99-01 IC HU3Page 85EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Hazardous GasInitiating Condition: Release of toxic, corrosive, asphyxiant or flammable gasesdeemed detrimental to NORMAL PLANT OPERATIONSEAL:HU3.2 Unusual EventRecommendation by local, county or state officials to evacuate or shelter site personnelbased on an offsite eventMode Applicability:AllBasis:Plant-SpecificA recommendation by offsite officials that a potential evacuation of site personnel may berequired based on an offsite event assumes that the plant lies within an evacuation areaestablished by offsite officials due to a release of toxic, corrosive, asphyxiant or flammablegas. In this case, it can be assumed that an actual or potential release of such hazardousgas is anticipated to enter the PROTECTED AREA in amounts that could affect the, healthof plant personnel or NORMAL PLANT OPERATIONS.GenericEscalation of this emergency classification level, if appropriate, would be based on EAL HA3.1.NMP1 Basis Reference(s):1. NEI 99-01 IC HU3Page 86 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety4 -SecurityHOSTILE ACTION resulting in loss of physical control of thefacilityEAL:HG4.1 General EmergencyA HOSTILE ACTION has occurred such that plant personnel are unable to operateequipment required to maintain safety functionsMode Applicability:AllBasis:Plant-SpecificSafety functions include:" Reactivity control -ability to shut down the reactor and keep it shutdown" RPV water level control -ability to cool the core* Decay heat removal -ability to maintain a heat sinkGenericThis EAL encompasses conditions under which a HOSTILE ACTION has resulted in a loss ofphysical control of VITAL AREAs (containing vital equipment or controls of vital equipment)required to maintain safety functions and control of that equipment cannot be transferred to andoperated from another location.If control of the plant equipment necessary to maintain safety functions can be transferred toanother location, then the threshold is not met.NMP1 Basis Reference(s):1. NEI 99-01 IC HG1Page 87 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 4 -SecurityInitiating Condition: HOSTILE ACTION resulting in loss of physical control of the facilityEAL:HG4.2 General EmergencyA HOSTILE ACTION has caused failure of Spent Fuel Cooling systemsANDIMMINENT fuel damage is likelyMode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses failure of spent fuel cooling systems as a result of HOSTILE ACTION ifIMMINENT fuel damage is likely.NMP1 Basis Reference(s):1. NEI 99-01 IC HG1Page 88EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisH -Hazards and Other Conditions Affecting Plant Safety4 -SecurityCategory:Subcategory:Initiating Condition: HOSTILE ACTION within the PROTECTED AREAEAL:HS4.1 Site Area EmergencyA HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA asreported by the Security Site SupervisorMode Applicability:AllBasis:Plant-SpecificNoneGenericThis condition represents an escalated threat to plant safety above that contained in the Alert inthat a HOSTILE FORCE has progressed from the Owner Controlled Area to the PROTECTEDAREA.This EAL addresses the contingency for a very rapid progression of events, such as thatexperienced on September 11, 2001. It is not premised solely on the potential for a radiologicalrelease. Rather the issue includes the need for rapid assistance due to the possibility for significantand indeterminate damage from additional air, land or water attack elements.The fact that the site is under serious attack with minimal time available for further preparation oradditional assistance to arrive requires Offsite Response Organization (ORO) readiness andpreparation for the implementation of protective measures.This EAL addresses the potential for a very rapid progression of events due to a HOSTILEACTION. It is not intended to address incidents that are accidental events or acts of civildisobedience, such as small aircraft impact, hunters, or physical disputes between employeeswithin the PROTECTED AREA. Those events are adequately addressed by other EALs.Escalation of this emergency classification level, if appropriate, would be based on actual plantstatus after impact or progression of attack.NMP1 Basis Reference(s):1. NMP Site Security Plan2. NEI 99-01 IC HS4Page 89EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisH -Hazards and Other Conditions Affecting Plant Safety4 -SecurityCategory:Subcategory:Initiating Condition: HOSTILE ACTION within the Owner Controlled Area or airborneattack threatEAL:HA4.1 AlertA HOSTILE ACTION is occurring or has occurred within the Owner Controlled Area asreported by the Security Site SupervisorORA validated notification from NRC of an AIRLINER attack threat within 30 min. of the siteMode Applicability:AllBasis:Plant-SpecificNoneGenericNote: Timely and accurate communication between the Security Site Supervisor and the ControlRoom is crucial for the implementation of effective Security EALs.This EAL addresses the contingency for a very rapid progression of events, such as thatexperienced on September 11, 2001. They are not premised solely on the potential for aradiological release. Rather the issue includes the need for rapid assistance due to the possibilityfor significant and indeterminate damage from additional air, land or water attack elements.The fact that the site is under serious attack or is an identified attack target with minimal timeavailable for further preparation or additional assistance to arrive requires a heightened state ofreadiness and implementation of protective measures that can be effective (such as on-siteevacuation, dispersal or sheltering).First ConditionThis condition addresses the potential for a very rapid progression of events due to a HOSTILEACTION. It is not intended to address incidents that are accidental events or acts of civildisobedience, such as small aircraft impact, hunters, or physical disputes between employeeswithin the Owner Controlled Area. Those events are adequately addressed by other EALs.Note that this condition is applicable for any HOSTILE ACTION occurring, or that has occurred, inthe Owner Controlled Area.Second ConditionPage 90EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisThis condition addresses the immediacy of an expected threat arrival or impact on the site within arelatively short time.The intent of this condition is to ensure that notifications for the AIRLINER attack threat are madein a timely manner and that Offsite Response Organizations (OROs) and plant personnel are at astate of heightened awareness regarding the credible threat. AIRLINER is meant to be a LARGEAIRCRAFT with the potential for causing significant damage to the plant.This condition is met when a plant receives information regarding an AIRLINER attack threat fromNRC and the AIRLINER is within 30 minutes of the plant. Only the plant to which the specific threatis made need declare the Alert.The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threatinvolves an AIRLINER (AIRLINER is meant to be a LARGE AIRCRAFT with the potential forcausing significant damage to the plant). The status and size of the plane may be provided byNORAD through the NRC.NMP1 Basis Reference(s):1. NMP Site Security Plan2. NEI 99-01 IC HA4Page 91EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety4 -SecurityConfirmed SECURITY CONDITION or threat which indicates apotential degradation in the level of safety of the plantHU4.1 Unusual EventA SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by theSecurity Site SupervisorORA credible site-specific security threat notificationORA validated notification from NRC providing information of an aircraft threatMode Applicability:AllBasis:Plant-SpecificIf the Security Site Supervisor determines that a threat notification is credible, the SecuritySite Supervisor will notify the Operations Shift Manager that a "Credible Threat" conditionexists for NMP. Generally, NMP Security Procedures address standard practices fordetermining credibility. The three main criteria for determining credibility are: technicalfeasibility, operational feasibility, and resolve. For NMP1, a validated notification deliveredby the FBI, the NRC or similar agency is treated as credible.GenericNote: Timely and accurate communication between the Security SiteSupervisor and the Control Room is crucial for the implementation of effective Security EALs.Security events which do not represent a potential degradation in the level of safety of the plant arereported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. Security events assessed asHOSTILE ACTIONs are classifiable under EAL HA4.1, EAL HS4.1 and EAL HG4.1.A higher initial classification could be made based upon the nature and timing of the security threatand potential consequences. The licensee shall consider upgrading the emergency responsestatus and emergency classification level in accordance with the NMP Site Security Plan.First ConditionPage 92 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisReference is made to security shift supervision because these individuals are the designatedpersonnel on-site qualified and trained to confirm that a security event is occurring or has occurred.Training on security event classification confirmation is closely controlled due to the strict secrecycontrols placed on the NMP Site Security Plan.This threshold is based on the NMP Site Security Plan. The NMP Site Security Plan is based onguidance provided by NEI 03-12.Second ConditionThis threshold is included to ensure that appropriate notifications for the security threat are made ina timely manner. This includes information of a credible threat. Only the plant to which the specificthreat is made need declare the Unusual Event.The determination of "credible" is made through use of information found in the NMP Site SecurityPlan.Third ConditionThe intent of this EAL is to ensure that notifications for the aircraft threat are made in a timelymanner and that Offsite Response Organizations and plant personnel are at a state of heightenedawareness regarding the credible threat. It is not the intent of this EAL to replace existing non-hostile related EALs involving aircraft.This EAL is met when a plant receives information regarding an aircraft threat from NRC.Validation is performed by calling the NRC or by other approved methods of authentication. Onlythe plant to which the specific threat is made need declare the Unusual Event.The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threatinvolves an AIRLINER (AIRLINER is meant to be a LARGE AIRCRAFT with the potential forcausing significant damage to the plant). The status and size of the plane may be provided byNORAD through the NRC.Escalation to Alert emergency classification level via EAL HA4.1 would be appropriate if the threatinvolves an AIRLINER within 30 minutes of the plant.NMP1 Basis Reference(s):1. NMP Site Security Plan2. NEI 99-01 IC HU4Page 93 EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisH -Hazards and Other Conditions Affecting Plant Safety5 -Control Room EvacuationCategory:Subcategory:Initiating Condition: Control Room evacuation has been initiated and plant controlcannot be establishedEAL:HS5.1 Site Area EmergencyControl Room evacuation has been initiatedANDControl of the plant cannot be established within 15 min.Mode Applicability:AllBasis:Plant-SpecificN1-SOP-21.2, Control Room Evacuation, provides specific instructions for evacuating theControl Room/Building and establishing plant control in alternate locations.GenericThe intent of this EAL is to capture those events where control of the plant cannot be reestablishedin a timely manner. In this case, expeditious transfer of control of safety systems has not occurred(although fission product barrier damage may not yet be indicated).The intent of the EAL is to establish control of important plant equipment and knowledge ofimportant plant parameters in a timely manner. Primary emphasis should be placed on thosecomponents and instruments that supply protection for and information about safety functions.Typically, these safety functions are reactivity control (ability to reach and maintain reactorshutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintaina heat sink).The determination of whether or not control is established at the remote shutdown panel is basedon Emergency Director (ED) judgment. The Emergency Director is expected to make a reasonable,informed judgment within the site specific time for transfer that the licensee has control of the plantfrom the remote shutdown panel.Escalation of this emergency classification level, if appropriate, would be by EALs in Category F orCategory R.NMP1 Basis Reference(s):1. N1-SOP-21.2 Control Room Evacuation2. NEI 99-01 IC HS2Page 94EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety5 -Control Room EvacuationControl Room evacuation has been initiatedEAL:HA5.1 AlertControl Room evacuation has been initiatedMode Applicability:AllBasis:Plant-SpecificN1 -SOP-21.2, Control Room Evacuation, provides specific instructions for evacuating theControl Room/Building and establishing plant control in alternate locations.GenericWith the control room evacuated, additional support, monitoring and direction through theTechnical Support Center and/or other emergency response facilities may be necessary.Inability to establish plant control from outside the control room will escalate this event to a SiteArea Emergency.NMP1 Basis Reference(s):1. N1-SOP-21.2 Control Room Evacuation2. NEI 99-01 IC HA5Page 95EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety6 -JudgmentOther conditions exist that in the judgment of the EmergencyDirector warrant declaration of a General EmergencyHG6.1 General EmergencyOther conditions exist which in the judgment of the Emergency Director indicate thatevents are in progress or have occurred which involve actual or IMMINENT substantialcore degradation or melting with potential for loss of containment integrity or HOSTILEACTION that results in an actual loss of physical control of the facility. Releases can bereasonably expected to exceed EPA Protective Action Guideline exposure levels (1,000mRem TEDE or 5,000 mRem thyroid CDE) offsite for more than the immediate site areaMode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDirector to fall under the emergency classification level description for General Emergency.NMP1 Basis Reference(s):1. NEI 99-01 IC HG2Page 96EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety6 -JudgmentOther conditions existing that in the judgment of the EmergencyDirector warrant declaration of a Site Area EmergencyHS6.1 Site Area EmergencyOther conditions exist which in the judgment of the Emergency Director indicate thatevents are in progress or have occurred which involve actual or likely major failures ofplant functions needed for protection of the public or HOSTILE ACTION that results inintentional damage or malicious acts; (1) toward site personnel or equipment that couldlead to the likely failure of or; (2) that prevent effective access to equipment needed for theprotection of the public. ANY releases are not expected to result in exposure levels whichexceed EPA Protective Action Guideline exposure levels (1,000 mRem TEDE or 5,000mRem thyroid CDE) beyond the SITE BOUNDARYMode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDirector to fall under the emergency classification level description for Site Area Emergency.NMP1 Basis Reference(s):1. NEI 99-01 IC HS3Page 97EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety6 -JudgmentOther conditions exist that in the judgment of the EmergencyDirector warrant declaration of an AlertHA6.1 AlertOther conditions exist which in the judgment of the Emergency Director indicate thatevents are in progress or have occurred which involve an actual or potential substantialdegradation of the level of safety of the plant or a security event that involves probable lifethreatening risk to site personnel or damage to site equipment because of HOSTILEACTION. ANY releases are expected to be limited to small fractions of the EPA ProtectiveAction Guideline exposure levels (1,000 mRem TEDE ord 5,000 mRem thyroid CDE)Mode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDirector to fall under the Alert emergency classification level.NMP1 Basis Reference(s):1. NEI 99-01 IC HA6Page 98EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety6 -JudgmentOther conditions existing that in the judgment of the EmergencyDirector warrant declaration of a UEHU6.1 Unusual EventOther conditions exist which in the judgment of the Emergency Director indicate thatevents are in progress or have occurred which indicate a potential degradation of the levelof safety of the plant or indicate a security threat to facility protection has been initiated. Noreleases of radioactive material requiring offsite response or monitoring are expectedunless further degradation of safety systems occursMode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDirector to fall under the UE emergency classification level.NMP1 Basis Reference(s):1. NEI 99-01 IC HU5Page 99EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory E -INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)EAL Group: Not Applicable (the EAL in this category isapplicable independent of plant operatingmode)An INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) is a complex that isdesigned and constructed for the interim storage of spent nuclear fuel and otherradioactive materials associated with spent fuel storage. A significant amount of theradioactive material contained within a cask/canister must escape its packaging and enterthe biosphere for there to be a significant environmental effect resulting from an accidentinvolving the dry storage of spent nuclear fuel. Formal offsite planning is not requiredbecause the postulated worst-case accident involving an ISFSI has insignificantconsequences to the public health and safety.A Notification of Unusual Event is declared on the basis of the occurrence of an event ofsufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged orviolated. This includes classification based on a loaded fuel storage cask/canisterCONFINEMENT BOUNDARY loss leading to the degradation of the fuel during storage orposing an operational safety problem with respect to its removal from storage.A hostile security event that leads to a potential loss in the level of safety of the ISFSI is aclassifiable event under Security category EAL HA4.1.Minor surface damage that does not affect storage cask/canister boundary is excludedfrom the scope of these EALs.Page 100 EPMP-EPP-0101Rev 00 Draft A Attachment I -Emergency Action Level Technical BasisCategory: E -ISFSISubcategory: Not ApplicableInitiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARYEAL:EUI.1 Unusual EventDamage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured doserates > then ANY of the following:* 400 mRem/hr at 3 feet from the HSM surface* 100 mRem/hr outside HSM door on centerline* 20 mRem/hr end shield wall exteriorMode Applicability:AllBasis:Plant-SpecificThe NMP site ISFSI utilizes the NUHOMS Horizontal Modular Storage System.This EAL addresses any condition which indicates a loss of a cask CONFINEMENTBOUNDARY and thus a potential degradation in the level of safety of the ISFSI. The caskCONFINEMENT BOUNDARY is the NUHOMS 61 BT Dry Shielded Canister (DSC). TheDSC is the pressure-retaining component of the storage system (ref. 1). Each loaded DSCis housed within a Horizontal Storage Module (HSM). Indication of a loss ofCONFINEMENT BOUNDARY is any increase in external HSM radiation levels in excess ofTechnical Specification limits (ref. 2).GenericAn UE in this EAL is categorized on the basis of the occurrence of an event of sufficient magnitudethat a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includesclassification based on a loaded fuel storage cask CONFINEMENT BOUNDARY loss leading tothe degradation of the fuel during storage or posing an operational safety problem with respect toits removal from storage.NMP1 Basis Reference(s):1. CDP No. N1-07-092/N2-07-070 Nine Mile Point Nuclear Station -Conceptual Design,Independent Spent Fuel Storage InstallationPage 101 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis2. Transnuclear, Inc. Standardized NUHOMS Horizontal Modular Storage SystemCertificate of Compliance No. 1004, Attachment A Technical Specifications Section1.2.7 HSM Dose Rates with a Loaded 24P, 52B or 61 BT DSC3. NEI 99-01 IC E-HU1Page 102 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory C -Cold Shutdown / Refueling System MalfunctionEAL Group: Cold Conditions (RCS temperature -2120F);EALs in this category are applicable only inone or more cold operating modes.Category C EALs are directly associated with cold shutdown or refueling system safetyfunctions. Given the variability of plant configurations (e.g., systems out-of-service formaintenance, containment open, reduced AC power redundancy, time since shutdown)during these periods, the consequences of any given initiating event can vary greatly. Forexample, a loss of decay heat removal capability that occurs at the end of an extendedoutage has less significance than a similar loss occurring during the first week aftershutdown. Compounding these events is the likelihood that instrumentation necessary forassessment may also be inoperable. The cold shutdown and refueling system malfunctionEALs are based on performance capability to the extent possible with consideration givento RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicableoperating modes (3 -Cold Shutdown, 4 -Refuel, D -Defueled).The events of this category pertain to the following subcategories:1. Loss of AC PowerLoss of emergency plant electrical power can compromise plant safety systemoperability including decay heat removal and emergency core cooling systems whichmay be necessary to ensure fission product barrier integrity. This category includesloss of onsite and offsite power sources for the 4.16 kV emergency buses.2. Loss of DC PowerLoss of emergency plant electrical power can compromise plant safety systemoperability including decay heat removal and emergency core cooling systems whichmay be necessary to ensure fission product barrier integrity. This category includesloss of power to the 125 VDC buses.Page 103 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis3. RPV Water LevelRPV water level is a measure of inventory available to ensure adequate core coolingand, therefore, maintain fuel clad integrity. The RPV provides a volume for the coolantthat covers the reactor core. The RPV and associated pressure piping (reactor coolantsystem) together provide a barrier to limit the release of radioactive material should thereactor fuel clad integrity fail.4. RCS TemperatureUncontrolled or inadvertent temperature or pressure increases are indicative of apotential loss of safety functions.5. Inadvertent CriticalityInadvertent criticalities pose potential personnel safety hazards as well being indicativeof losses of reactivity control.6. CommunicationsCertain events that degrade plant operator ability to effectively communicate withessential personnel within or external to the plant warrant emergency classification.Page 104 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -Loss of AC PowerInitiating Condition: Loss of all offsite and all onsite AC power to 4.16 kV emergencybuses for > 15 min.EAL:CAl.1 AlertLoss of all offsite and all onsite AC power, Table C-1, to 4.16 kV emergency busesfor __ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.F _ Table C-1 AC Power Sources I* DG 102* DG 10300 T-101N* T-101So

• T-10 backfed from offsite through T-1or T-2 (only if already aligned)Mode Applicability:3 -Cold Shutdown, 4 -Refuel, D -DefueledBasis:Plant-SpecificNMP1 4.16 kV emergency buses are buses PB102 and PB103, which feed all Stationredundant safety-related loads.. There are three offsite power sources available to thesebuses (ref. 1, 2):" Offsite 115 kV through transformer 101N. This is the normal power supply toPB 102.* Offsite 115 kV through transformer 101S. This is the normal power supply toPB103.Page 105 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis* Offsite 345 kV through transformer T-1 or T-2 backfed through transformer T-1 0.Based on operational experience, if the backfeed is not already aligned, this cannot beconsidered available/capable of supplying the bus due to the time it will take to align it. Inany case, if this cannot be accomplished within 15 minutes, it is not available and anUnusual Event must be declared.There are two onsite AC power sources:* DG102forPB102* DG103 for PB103The fifteen-minute interval was selected as a threshold to exclude transient power losses.If multiple sources fail to energize the unit safety-related buses within 15 minutes, anUnusual Event is declared under this EAL. The subsequent loss of the single remainingpower source escalates the event to an Alert under EAL CA1.1.This EAL is the cold condition equivalent of the hot condition loss of all AC power EALSS1.1.GenericLoss of all AC power compromises all plant safety systems requiring electric power including RHR,ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink.The event can be classified as an Alert when in cold shutdown, refuel, or defueled mode becauseof the significantly reduced decay heat and lower temperature and pressure, increasing the time torestore one of the emergency busses, relative to that specified for the Site Area Emergency EAL.Escalating to Site Area Emergency, if appropriate, is by EALs in Category R.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.NMP1 Basis Reference(s):1. Nl-OP-30 4.16 kV, 600V, and 480V House Service2. USAR section IX Electrical Systems3. NEI 99-01 IC CA3Page 106 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction1 -Loss of AC PowerAC power capability to 4.16 kV emergency buses reduced to asingle power source for >_ 15 min. such that ANY additional singlefailure would result in a complete loss of all 4.16 kV emergencybus powerEAL:CUI.1 Unusual EventAC power capability to 4.16 kV emergency buses reduced to a single power source, TableC-1, for _ 15 min. (Note 4)ANDANY additional single power source failure will result in a loss of all 4.16 kV emergencybus powerNote 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-1 AC Power Sources* DG 102,- 0 DG 10300 T-101N0 T-101SO

• T-10 backfed from offsite through T-1or T-2 (only if already aligned)Mode Applicability:3 -Cold Shutdown, 4 -Refuel, D -DefueledBasis:Plant-SpecificNMP1 4.16 kV emergency buses are buses PB102 and PB103, which feed all Stationredundant safety-related loads.. There are three offsite power sources available to thesebuses (ref. 1, 2):Page 107EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis" Offsite 115 kV through transformer 101 N. This is the normal power supply toPB102.* Offsite 115 kV through transformer 101S. This is the normal power supply toPB103." Offsite 345 kV through transformer T-1 or T-2 backfed through transformer T-1 0.Based on operational experience, if the backfeed is not already aligned, this cannot beconsidered available/capable of supplying the bus due to the time it will take to align it. Inany case, if this cannot be accomplished within 15 minutes, it is not available and anUnusual Event must be declared.There are two onsite AC power sources:* DG 102 for PB102* DG103 for PB103The fifteen-minute interval was selected as a threshold to exclude transient power losses.If multiple sources fail to energize the unit safety-related buses within 15 minutes, an-Unusual Event is declared under this EAL. The subsequent loss of the single remainingpower source escalates the event to an Alert under EAL CA1.1.GenericThe condition indicated by this EAL is the degradation of the off-site and on-site AC power systemssuch thatany additional single failure would result in a complete loss of 4.16 kV emergency busAC power. This condition could occur due to a loss of off-site power with a concurrent failure of oneemergency generator to supply power to its emergency bus. The subsequent loss of this singlepower source would escalate the event to an Alert in accordance with EAL CA1.1.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.NMP1 Basis Reference(s):1. Nl-OP-30 4.16 kV, 600V, and 480V House Service2. USAR section IX Electrical Systems3. NEI 99-01 IC CU3Page 108 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 2 -Loss of DC PowerInitiating Condition: Loss of required DC power for _ 15 min.EAL:CU2.1 Unusual Event< 106 VDC on required 125 VDC buses (Battery board 11, Battery board 12) for _ 15 min.(Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Mode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificA Safety Related (SR) system and a Quality Related (QR) system comprise the 125 VDCPower System. The two SR 125 VDC systems (Battery board 11 and Battery board 12)each consist of: one battery, two Static Chargers in parallel, and a DC distribution board.The one QR 125 VDC system consists of: a battery, one Static Charger, and one batteryboard. This EAL addresses only the Safety Related battery boards (ref. 1, 2).106 VDC is the minimum voltage for battery operability (ref.3).This EAL is the cold condition equivalent of the hot condition loss of DC powerEAL SS2.1.GenericThe purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor andcontrol the removal of decay heat during Cold Shutdown or Refueling operations.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.NMP1 Basis Reference(s):1. N1-OP-47A 125 VDC Power System2. USAR section IX Electrical Systems3. NMP1 Technical Specification 3.6.34. NEI 99-01 IC CU7Page 109 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction3 -RPV Water LevelLoss of RPV inventory affecting fuel clad integrity withContainment challengedEAL:CG3.1 General EmergencyRPV water level < -84 in. for __ 30 min. (Note 4)ANDANY Containment Challenge Indication, Table C-3Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-3 Containment Challenge Indications* CONTAINMENT CLOSURE not established0 Explosive mixture exists inside PrimaryContainment (H2 > 6% and 02 > 5%)0 UNPLANNED rise in Primary Containmentpressure0 RB area radiation > 8 R/hrMode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificWhen RPV water level drops the top of active fuel, an indicated RPV water level of -84 in.(rounded from 84 7/16"), core uncovery starts to occur (ref. 1, 2).Four conditions are associated with a challenge to Primary Containment integrity:CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteriagoverning Primary and Secondary Containment integrity. If the TechnicalPage 110 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisSpecification criteria are met, therefore, CONTAINMENT CLOSURE has beenestablished. (ref. 3, 4)Explosive (deflagration) mixtures in the Primary Containment are assumed to beelevated concentrations of hydrogen and oxygen. BWR industry evaluation ofhydrogen generation for development of EOPs/SAGs indicates that any hydrogenconcentration above minimum detectable is not to be expected within the shortterm. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowlyevolving, long-term condition. Hydrogen concentrations that rapidly develop aremost likely caused by metal-water reaction. A metal-water reaction is indicative ofan accident more severe than accidents considered in the plant design basis andwould be indicative, therefore, of a potential threat to Primary Containment integrity.Hydrogen concentration of approximately 6% is considered the global deflagrationconcentration limit.The specified values for this threshold are the minimum global deflagrationconcentration limits (6% hydrogen and 5% oxygen), and readily recognizablebecause 6% hydrogen is well above the EOP flowchart entry condition. Theminimum global deflagration hydrogen/oxygen concentrations (6%/5%, respectively)require intentional Primary Containment venting, which is defined to be a loss of thePrimary Containment barrier. (ref. 2, 5)If the hydrogen or oxygen monitor is unavailable, sampling and analysis maydetermine gas concentrations. The validity of sample results must be judged basedupon plant conditions, since drawing and analyzing samples may take some time. Ifsample results cannot be relied upon and hydrogen concentrations cannot bedetermined by any other means, the concentrations must be considered "unknown."The monitors should not be considered "unavailable" until an attempt has beenmade to place them in service. (ref. 2)Any UNPLANNED rise in Primary Containment pressure in the Cold Shutdown orRefuel mode indicates CONTAINMENT CLOSURE cannot be assured and thePrimary Containment cannot be relied upon as a barrier to fission product release.Page 111 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisRB (Reactor Building) area radiation monitors should provide indication of increasedrelease that may be indicative of a challenge to CONTAINMENT CLOSURE. TheEOP Maximum Safe Operating level is 8 R/hr and is indicative of problems in thesecondary containment that are spreading. The locations into which the primarysystem discharge is of concern correspond to the areas addressed in Detail S ofN1-EOP-5 (ref. 6).If RPV water level is restored and maintained above the top of active fuel before aContainment Challenge condition occurs and subsequently a Containment Challengecondition is reached, this EAL is not met.GenericThis EAL represents the inability to restore and maintain RPV water level to above the top of activefuel with containment challenged. Fuel damage is probable if RPV water level cannot be restored,as available decay heat will cause boiling, further reducing the RPV water level. With theContainment breached or challenged then the potential for unmonitored fission product release tothe environment is high. This represents a direct path for radioactive inventory to be released tothe environment. This is consistent with the definition of a GE. The GE is declared on theoccurrence of the loss or IMMINENT loss of function of all three barriers.A number of variables can have a significant impact on heat removal capability challenging the fuelclad barrier. Examples include: initial vessel level and shutdown heat removal system design.Analysis indicates that core damage may occur within an hour following continued core uncoverytherefore, 30 minutes was conservatively chosen.If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute core uncoverytime limit then escalation to General Emergency would not occur.NMP1 Basis Reference(s):1. N2-EOP-2 RPV Control2. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document3. NIP-OUT-01 Shutdown Safety4. NMP1 Technical Specifications Definitions 1.11 and 1.125. N1-EOP-4.2 Hydrogen Control6. N1-EOP-5 Secondary Containment Control7. NEI 99-01 IC CG1Page 112 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction3 -RPV Water LevelLoss of Reactor Vessel inventory affecting fuel clad integrity withContainment challengedEAL:CG3.2 General EmergencyRPV water level cannot be monitored with core uncovery indicated by ANY of thefollowing for __ 30 min. (Note 4):" ANY UNPLANNED RPV leakage indication, Table C-2" Erratic Source Range Monitor indicationANDANY Containment Challenge Indication, Table C-3Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeTable C-2 RPV Leakage Indications* Drywell equipment drain tank level rise* Drywell floor drain tank level rise* Reactor building equipment sump level rise* Reactor Building floor drain sump level rise* Torus water level rise* UNPLANNED rise in RPV make-up rate* Observation of UNISOLABLE RCS leakageTable C-3 Containment Challenge Indications* CONTAINMENT CLOSURE not established* Explosive mixture exists inside PrimaryContainment (H2 > 6% and 02 > 5%)* UNPLANNED rise in Primary Containmentpressure* RB area radiation > 8 R/hrPage 113EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisMode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificIf RPV water level monitoring capability is unavailable, all RPV water level indication wouldbe unavailable and, the RPV inventory loss must be detected by Table C-2, RPV LeakageIndications. Level increases must be evaluated against other potential sources of leakagesuch as cooling water sources inside the drywell to ensure they are indicative of RPVleakage. Drywell equipment and floor drain tank level rise is the normal method ofmonitoring and calculating leakage from the RPV. A Reactor Building equipment or floordrain sump level rise may also be indicative of RPV inventory losses external to thePrimary Containment from systems connected to the RPV. A rise in torus water level couldbe indicative of valve misalignment or leakage. If the make-up rate to the RPVunexplainably rises above the pre-established rate, a loss of RPV inventory may beoccurring even if the source of the leakage cannot be immediately identified. Visualobservation of leakage from systems connected to the RCS in areas outside the PrimaryContainment that cannot be isolated could be indicative of a loss of RPV inventory. (ref.1through 6)Four channels of log count rate meters are available in the Control Room to detect erraticsource range monitor indications (ref. 7): SRM Channels 11, 12, 13, and 14 located on theE console.Post-TMI studies indicated that the installed nuclear instrumentation will operate erraticallywhen the core is uncovered and that source range monitors can be used as a tool formaking such determinations. Figure C-2 shows the response of the source range monitorduring the first few hours of the TMI-2 accident. The instrument reported an increasingsignal about 30 minutes into the accident. At this time, the reactor coolant pumps wererunning and the core was adequately cooled as indicated by the core outletthermocouples. Hence, the increasing signal was the result of an increasing two-phasevoid fraction in the reactor core and vessel downcomer and the reduced shielding that thetwo-phase mixture provides to the source range monitor.Page 114 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisFour conditions are associated with a challenge to Primary Containment integrity:* CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteriagoverning Primary and Secondary Containment integrity. If the TechnicalSpecification criteria are met, therefore, CONTAINMENT CLOSURE has beenestablished. (ref. 10, 11)* Explosive (deflagration) mixtures in the Primary Containment are assumed to beelevated concentrations of hydrogen and oxygen. BWR industry evaluation ofhydrogen generation for development of EOPs/SAGs indicates that any hydrogenconcentration above minimum detectable is not to be expected within the shortterm. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowlyevolving, long-term condition. Hydrogen concentrations that rapidly develop aremost likely caused by metal-water reaction. A metal-water reaction is indicative ofan accident more severe than accidents considered in the plant design basis andwould be indicative, therefore, of a potential threat to Primary Containment integrity.Hydrogen concentration of approximately 6% is considered the global deflagrationconcentration limit.The specified values for this threshold are the minimum global deflagrationconcentration limits (6% hydrogen and 5% oxygen), and readily recognizablebecause 6% hydrogen is well above the EOP flowchart entry condition. Theminimum global deflagration hydrogen/oxygen concentrations (6%/5%, respectively)require intentional Primary Containment venting, which is defined to be a loss of thePrimary Containment barrier. (ref. 9, 12)If the hydrogen or oxygen monitor is unavailable, sampling and analysis maydetermine gas concentrations. The validity of sample results must be judged basedupon plant conditions, since drawing and analyzing samples may take some time. Ifsample results cannot be relied upon and hydrogen concentrations cannot bedetermined by any other means, the concentrations must be considered "unknown."Page 115 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisThe monitors should not be considered "unavailable" until an attempt has beenmade to place them in service. (ref. 9)" Any UNPLANNED rise in Primary Containment pressure in the Cold Shutdown orRefuel mode indicates CONTAINMENT CLOSURE cannot be assured and thePrimary Containment cannot be relied upon as a barrier to fission product release." RB (Reactor Building) area radiation monitors should provide indication of increasedrelease that may be indicative of a challenge to CONTAINMENT CLOSURE. TheEOP Maximum Safe Operating level is 8 R/hr and is indicative of problems in thesecondary containment that are spreading. The locations into which the primarysystem discharge is of concern correspond to the areas addressed in Detail S ofN1-EOP-5 (ref. 13).If RPV water level is restored and maintained above the top of active fuel before aContainment Challenge condition occurs and subsequently a Containment Challengecondition is reached, this EAL is not met.GenericThis EAL represents the inability to restore and maintain RPV water level to above the top of activefuel with containment challenged. Fuel damage is probable if RPV water level cannot be restored,as available decay heat will cause boiling, further reducing the RPV water level. With theContainment breached or challenged then the potential for unmonitored fission product release tothe environment is high. This represents a direct path for radioactive inventory to be released tothe environment. This is consistent with the definition of a GE. The GE is declared on theoccurrence of the loss or IMMINENT loss of function of all three barriers.A number of variables can have a significant impact on heat removal capability challenging the fuelclad barrier. Examples include: initial RPV water level and shutdown heat removal system design.Analysis indicates that core damage may occur within an hour following continued core uncoverytherefore, 30 minutes was conservatively chosen.If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute core uncoverytime limit then escalation to General Emergency would not occur.Sump and tank level increases must be evaluated against other potential sources of leakage suchas cooling water sources inside the containment to ensure they are indicative of RCS leakage.As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due tothis core shine should result in site specific monitor indication and possible alarm.Page 116 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisPost-TMI studies indicated that the installed nuclear instrumentation will operate erratically whenthe core is uncovered and that this should be used as a tool for making such determinations.NMP1 Basis Reference(s):1. S-ODP-OPS-0110 Containment Leakage Evaluation2. USAR 1.4 Primary Coolant Leakage3. Annunciator H2 1 DRYWELL FLOOR DRAIN LEVEL-HIGH4. Annunciator H2 7 DRYWELL WATER LEAK DETECTION SYS5. Annunciator H2 1 R BLDG FL DR SUMPS 11-16 AREA WTR LVL LEVEL HIGH6. Annunciator H2 2 R BUILDING EQUIP DRAIN LEVEL-HIGH7. Nl-OP-38A Source Range Monitor8. N1-EOP-2 RPV Control9. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document10.NIP-OUT-01 Shutdown Safety11.NMP1 Technical Specifications Definitions 1.11 and 1.1212.N1-EOP-4.2 Hydrogen Control13. N1-EOP-5 Secondary Containment Control14. NEI 99-01 IC CG1Page 117 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisFigure C-2: Response of the TMI-2 Source Range MeasurementDuring the First Six Hours of the AccidentC\J0C'J(D(DWE0L0 .0 o.CD(sapeoap bol) puooeS jad sjunooPage 118EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: Loss of RPV inventory affecting core decay heat removal capabilityEAL:CS3.1 Site Area EmergencyWith CONTAINMENT CLOSURE not established, RPV water level < -1 in.Mode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificWhen RPV water level decreases to -1 in., water level is six inches below the Core Sprayinitiation setpoint (ref. 1).The inability to restore and maintain level after reaching this setpoint infers a failure of theRCS barrier and Potential Loss of the Fuel Clad barrier.CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteria governingPrimary and Secondary Containment integrity. If the Technical Specification criteria aremet, therefore, CONTAINMENT CLOSURE has been established. (ref. 3)GenericUnder the conditions specified by this EAL, continued decrease in /RPV water level is indicative ofa loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundaryleakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.Escalation to a General Emergency is via EAL CG3.1, EAL CG3.2, RG1.2 or RG1.3.NMP1 Basis Reference(s):1. NMP1 Technical Specification 3.6.2 Table 3.6.2.d2. NIP-OUT-01 Shutdown Safety3. NMP1 Technical Specification Definitions 1.11 and 1.124. NEI 99-01 IC CS1Page 119 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: Loss of RPV inventory affecting core decay heat removal capabilityEAL:CS3.2 Site Area EmergencyWith CONTAINMENT CLOSURE established, RPV water level < -84 in.Mode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificWhen RPV water level drops the top of active fuel, an indicated RPV water level of -84 in.(rounded from 84 7/16"), core uncovery starts to occur (ref. 1, 2).CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteria governingPrimary and Secondary Containment integrity. If the Technical Specification criteria aremet, therefore, CONTAINMENT CLOSURE has been established. (ref. 3, 4)GenericUnder the conditions specified by this EAL, continued decrease in RCS level is indicative of a lossof inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, orcontinued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.Escalation to a General Emergency is via EAL CG3.1, EAL CG3.2, RG1.2 or RG1.3.NMP1 Basis Reference(s):1. N1-EOP-2 RPV Control2. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document3. NIP-OUT-01 Shutdown Safety4. NMP1 Technical Specification Definitions 1.11 and 1.125. NEI 99-01 IC CS1Page 120 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: Loss of RPV inventory affecting core decay heat removal capabilityEAL:CS3.3 Site Area EmergencyRPV water level cannot be monitored for __ 30 min. (Note 4) with a loss of RPV inventoryas indicated by ANY of the following:" ANY UNPLANNED RPV leakage indication, Table C-2" Erratic Source Range Monitor indicationNote 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-2 RPV Leakage Indications* Drywell equipment drain tank level rise* Drywell floor drain tank level rise* Reactor building equipment sump level rise* Reactor Building floor drain sump level rise* Torus water level rise* UNPLANNED rise in RPV make-up rate* Observation of UNISOLABLE RCS leakageMode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificIf RPV water level monitoring capability is unavailable, all RPV water level indication wouldbe unavailable and, the RPV inventory loss must be detected by Table C-2, RPV LeakageIndications. Level increases must be evaluated against other potential sources of leakagesuch as cooling water sources inside the drywell to ensure they are indicative of RPVleakage. Drywell equipment and floor drain tank level rise is the normal method ofmonitoring and calculating leakage from the RPV. A Reactor Building equipment or floorPage 121 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisdrain sump level rise may also be indicative of RPV inventory losses external to thePrimary Containment from systems connected to the RPV. A rise in torus water level couldbe indicative of valve misalignment or leakage. If the make-up rate to the RPVunexplainably rises above the pre-established rate, a loss of RPV inventory may beoccurring even if the source of the leakage cannot be immediately identified. Visualobservation of leakage from systems connected to the RCS in areas outside the PrimaryContainment that cannot be isolated could be indicative of a loss of RPV inventory. (ref.1through 6)Four channels of log count rate meters are available in the Control Room to detect erraticsource range monitor indications (ref. 7): SRM Channels 11, 12, 13, and 14 located on theE console.Post-TMI studies indicated that the installed nuclear instrumentation will operate erraticallywhen the core is uncovered and that source range monitors can be used as a tool formaking such determinations. Figure C-2 shows the response of the source range monitorduring the first few hours of the TMI-2 accident. The instrument reported an increasingsignal about 30 minutes into the accident. At this time, the reactor coolant pumps wererunning and the core was adequately cooled as indicated by the core outletthermocouples. Hence, the increasing signal was the result of an increasing two-phasevoid fraction in the reactor core and vessel downcomer and the reduced shielding that thetwo-phase mixture provides to the source range monitor.GenericUnder the conditions specified by this EAL, continued decrease in /RPV water level is indicative ofa loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundaryleakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.Escalation to a General Emergency is via EAL CG3.1, EAL CG3.2, RG1.2 or RG1.3.The 30-minute duration allows sufficient time for actions to be performed to recover inventorycontrol equipment.As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due tothis core shine should result in site specific monitor indication and possible alarm.NMP1 Basis Reference(s):1. S-ODP-OPS-0110 Containment Leakage EvaluationPage 122 EPMP-EPP-0101Rev 00 Draft A 2.3.4.5.6.7.8.Attachment 1 -Emergency Action Level Technical BasisUSAR 1.4 Primary Coolant LeakageAnnunciator H2 1 DRYWELL FLOOR DRAIN LEVEL-HIGHAnnunciator H2 7 DRYWELL WATER LEAK DETECTION SYSAnnunciator H2 1 R BLDG FL DR SUMPS 11-16 AREA WTR LVL LEVEL HIGHAnnunciator H2 2 R BUILDING EQUIP DRAIN LEVEL-HIGHN1-OP-38A Source Range MonitorNEI 99-01 IC CS1Page 123EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisFigure C-2: Response of the TMI-2 Source Range MeasurementDuring the First Six Hours of the AccidentV\0a,~1)cia,cI-a,a,E1=CL1C)(0U' I(sapeoap bol) puooaS jad sjunocoPage 124EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: Loss of RPV inventoryEAL:CA3.1 AlertRPV water level < +5 in.ORRPV water level cannot be monitored for -15 min. with ANY UNPLANNED RPV leakageindication, Table C-2 (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-2 RPV Leakage Indications* Drywell equipment drain tank level rise* Drywell floor drain tank level rise* Reactor building equipment sump level rise* Reactor Building floor drain sump level rise* Torus water level rise* UNPLANNED rise in RPV make-up rate* Observation of UNISOLABLE RCS leakageMode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificThe threshold RPV water level of +5 in. is the Core Spray initiation setpoint (ref. 1). RPVwater level is normally monitored using the instruments in Figure C-1 (ref. 2).In Cold Shutdown mode, the RCS will normally be INTACT and standard RPV water levelmonitoring means are available. In the Refuel mode, the RCS is not INTACT and RPVwater level may be monitored by different means, including the ability to monitor levelvisually.Page 125EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisIn the second condition of this EAL, all RPV water level indication is unavailable and theRPV inventory loss must be detected by the leakage indications listed in Table C-2. Levelincreases must be evaluated against other potential sources of leakage such as coolingwater sources inside the drywell to ensure they are indicative of RPV leakage. Drywellequipment and floor drain tank level rise is the normal method of monitoring andcalculating leakage from the RPV (ref. 3 through 8). A Reactor Building equipment or floordrain sump level rise may also be indicative of RPV inventory losses external to thePrimary Containment from systems connected to the RPV. A rise in torus water level couldbe indicative of valve misalignment or leakage. If the make-up rate to the RPVunexplainably rises above the pre-established rate, a loss of RPV inventory may beoccurring even if the source of the leakage cannot be immediately identified. Visualobservation of leakage from systems connected to the RCS in areas outside the PrimaryContainment that cannot be isolated could be indicative of a loss of RPV inventory.The 15-minute interval for the loss of level indication was chosen because it is half of theSite Area Emergency EAL duration. The interval allows this EAL to be an effectiveprecursor to the Site Area Emergency EAL CS3.1. Therefore this EAL meets the definitionfor an Alert emergency.GenericThis EAL serves as a precursor to a loss of ability to adequately cool the fuel. The magnitude ofthis loss of water indicates that makeup systems have not been effective and may not be capableof preventing further RPV water level decrease and potential core uncovery. This condition willresult in a minimum emergency classification level of an Alert.The inability to restore and maintain level after reaching this setpoint would be indicative of afailure of the RCS barrier.If RPV water level continues to lower then escalation to Site Area Emergency will be via EALCS3.1, EAL CS3.2 or EAL CS3.3.NMP1 Basis Reference(s):1. NMP1 Technical Specification 3.6.2 Table 3.6.2.d2. P&ID C-18015-C, Reactor Vessel Instrumentation3. S-ODP-OPS-01 10 Containment Leakage Evaluation4. USAR 1.4 Primary Coolant Leakage5. Annunciator H2 1 DRYWELL FLOOR DRAIN LEVEL-HIGH6. Annunciator H2 7 DRYWELL WATER LEAK DETECTION SYSPage 126 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis7. Annunciator H2 1 R BLDG FL DR SUMPS 11-16 AREA WTR LVL LEVEL HIGH8. Annunciator H2 2 R BUILDING EQUIP DRAIN LEVEL-HIGH9. NEI 99-01 IC CA1Page 127 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisFigure C-1 RPV Water Levels-Instrument zero= 297'4"324'-7 W3215'-8 25,32'SURFACE b *'OF 3r519&"0 3"9&10NORM LEVEL- KI ERM.NORM LEVEL 70 .-l,.MM( SCRAM HpCI.T!P.TSV)EL 297'- 'r LO LO LEVEL {ISOLATION. EMERG CONC..CCRE&10 CCONT. SPPAY, TRIP RECIRC PtiMPI,.(ARNATWS)t 'EL (AID.S)EL 3 IT STEAM Nr;7ý FEL 278'-25716"OTrTOM OF ACEIVEEL 274- 9R ECIRCULLATiONPage 128EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: RCS leakageEAL:CU3.1 Unusual EventRCS leakage results in the inability to maintain or restore RPV water level > +53 in.for >_ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeMode Applicability:3 -Cold ShutdownBasis:Plant-SpecificFigure C-1 illustrates the elevations of the RPV water level instrument ranges (ref. 1).+53 in. is the RPV low water level scram setpoint (ref. 5).RPV water level is monitored from the bottom of active fuel to the top of the RPV head toensure adequate coverage for expected and postulated conditions of RPV water level.RPV water level measurement is derived by the differential pressure that exists between areference leg and variable leg. All instruments are referenced to a benchmark at 439 in.above the inside bottom head of the reactor vessel with the exception of the Suppressedor Flange Level range. This benchmark corresponds to 84 in. (rounded from 84 7/16")above the top of active fuel and is the 0 in. reference indication on the RPV water levelinstruments (except the suppressed or flange level). RPV water level monitoring issubdivided into five ranges identified as:* Control range provides indication and control signals for normal plant operation andprotection system actuation.* Lo-lo-lo range provides indication and control signals for transient conditions belowthe normal operating band and emergency equipment actuation.Page 129 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis* Wide range provides indication for transient conditions above normal operatingband.* Suppressed or flange level provides indication relative to the reactor vessel flange." Fuel Zone provides indication for long term accident conditions where RPV waterlevel cannot be restored.In preparation for refueling operations, the Wide and Suppressed RPV water levelinstruments are modified to provide continuous level indication from within the RPV to thereactor cavity.GenericThis EAL is considered to be a potential degradation of the level of safety of the plant. The inabilityto maintain or restore level is indicative of loss of RCS inventory.Relief valve normal operation should be excluded from this EAL. However, a relief valve thatoperates and fails to close per design should be considered applicable to this EAL if the relief valvecannot be isolated.Prolonged loss of RCS inventory may result in escalation to the Alert emergency classification levelvia either EAL CA2.1 or EAL CA3.1.NMP1 Basis Reference(s):1. P&ID C-18015-C, Reactor Vessel Instrumentation2. N1-OP-58 RPV Level Backfill Injection System3. USAR Section VIII.A.2.1, Reactor Water Level4. USAR Section VIII.C.2.1.1, Reactor Water Level5. NMP1 Technical Specification 3.6.2 Table 3.6.2.a6. NEI 99-01 IC CU1Page 130 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisFigure C-1 RPV Water Levels-Instrument zero= 297'4"2IIN. SURFACflI C P20 20 IM L.2EL -' 10 10 LEE ISLTM O.EEOCO OE3F _ -,A.BMIw EC.35H. CITRIP.TSV}2" EL.29,7'- 9" LO LO LEVEL {ISOILATION. EMERG CONO. CORE&I 10 10 CCST> SPPAY. TRIP PECIPC PULIPtARI(ATWS)----EL (A DS.)EL 3 .17-0 .EL.278"-2 5.6BOTOM OF ACTIVEFULEL 274' 9"PECIRCU CAT OjNOuZTLCPage 131EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: RCS leakageEAL:CU3.2 Unusual EventUNPLANNED RPV water level drop below EITHER of the following for > 15 min. (Note 4):* 0 ft Flange Level (RPV flange)0 RPV water level band (when the RPV water level band is established below theRPV flange)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeMode Applicability:4 -RefuelBasis:Plant-SpecificFigure C-1 illustrates the elevations of the RPV water level instrument ranges (ref. 1).The RPV flange mating surface is at 0 ft on the Flange Level instrument or EL 315'-825/32" (ref. 1).RPV water level is monitored from the bottom of active fuel to the top of the RPV head toensure adequate coverage for expected and postulated conditions of RPV water level.RPV water level measurement is derived by the differential pressure that exists between areference leg and variable leg. All instruments are referenced to a benchmark at 439 in.above the inside bottom head of the reactor vessel with the exception of the Suppressedor Flange Level range. This benchmark corresponds to 84 in. (rounded from 84 7/16")above the top of active fuel and is the 0 in. reference indication on the RPV water levelinstruments (except the suppressed or flange level). RPV water level monitoring issubdivided into five ranges identified as:* Control range provides indication and control signals for normal plant operation andprotection system actuation.Page 132 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis" Lo-lo-lo range provides indication and control signals for transient conditions belowthe normal operating band and emergency equipment actuation." Wide range provides indication for transient conditions above normal operatingband.* Suppressed or flange level provides indication relative to the reactor vessel flange." Fuel Zone provides indication for long term accident conditions where RPV waterlevel cannot be restored.In preparation for refueling operations, the Wide and Suppressed RPV water levelinstruments are modified to provide continuous level indication from within the RPV to thereactor cavity.GenericThis EAL is a precursor of more serious conditions and considered to be a potential degradation ofthe level of safety of the plant.Refueling evolutions that decrease RPV water level below the RPV flange are carefully plannedand procedurally controlled. An UNPLANNED event that results in water level decreasing belowthe RPV flange, or below the planned RPV water level for the given evolution (if the planned RPVwater level is already below the RPV flange), warrants declaration of a UE due to the reduced RCSinventory that is available to keep the core covered.The allowance of 15 minutes was chosen because it is reasonable to assume that level can berestored within this time frame using one or more of the redundant means of refill that should beavailable. If level cannot be restored in this time frame then it may indicate a more seriouscondition exists.Continued loss of RCS Inventory will result in escalation to the Alert emergency classification levelvia either EAL CA2.1 or EAL CA3.1.This EAL involves a decrease in RPV water level below the top of the RPV flange that continuesfor 15 minutes due to an UNPLANNED event. This EAL is not applicable to decreases in floodedreactor cavity level, which is addressed by EAL RU2.1, until such time as the level decreases tothe level of the vessel flange.NMP1 Basis Reference(s):1. P&ID C-18015-C, Reactor Vessel Instrumentation2. N1-OP-58 RPV Level Backfill Injection System3. USAR Section VIII.A.2.1, Reactor Water Level4. USAR Section VIII.C.2.1.1, Reactor Water Level5. NMP1 Technical Specification 3.6.2 Table 3.6.2.aPage 133 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis6. NEI 99-01 IC CU2Page 134EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisFigure C-1 RPV Water Levels-Instrument zero= 297'4VING SURtFACE o 'i 5L 297r-r9"LO LO LEVEL CISCLATION.OEMER COND.CORE&"0 I 0 a CONT. SPRAYT O3-M0 (I. 0 0MI< NO MLEE F ., EIXOEL 278-2 51166OTT7M OF ACTIVEFUELEL .T4'- 9"RECIRCULATONoWIU ---Page 135EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: RCS leakageEAL:CU3.3 Unusual EventRPV water level cannot be monitored with a loss of RPV inventory as indicated by ANYUNPLANNED RPV leakage indication, Table C-2Table C-2 RPV Leakage Indications* Drywell equipment drain tank level rise* Drywell floor drain tank level rise* Reactor building equipment sump level rise* Reactor Building floor drain sump level rise* Torus water level rise* UNPLANNED rise in RPV make-up rate* Observation of UNISOLABLE RCS leakageMode Applicability:4 -RefuelBasis:Plant-SpecificIn this EAL, all RPV water level indication is unavailable and the RPV inventory loss mustbe detected by the leakage indications listed in Table C-2, RPV Leakage Indications. Levelincreases must be evaluated against other potential sources of leakage such as coolingwater sources inside the drywell to ensure they are indicative of RPV leakage. Drywellequipment and floor drain tank level rise is the normal method of monitoring andcalculating leakage from the RPV (ref. 1, 2). A Reactor Building equipment or floor drainsump level rise may also be indicative of RPV inventory losses external to the PrimaryContainment from systems connected to the RPV. A rise in torus water level could beindicative of valve misalignment or leakage. If the make-up rate to the RPV unexplainablyrises above the pre-established rate, a loss of RPV inventory may be occurring even if thePage 136 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basissource of the leakage cannot be immediately identified. Visual observation of leakage fromsystems connected to the RCS in areas outside the Primary Containment that cannot beisolated could be indicative of a loss of RPV inventory.GenericThis EAL is a precursor of more serious conditions and considered to be a potential degradation ofthe level of safety of the plant.Refueling evolutions that decrease RCS water level below the Reactor Vessel flange are carefullyplanned and procedurally controlled. An UNPLANNED event that results in water level decreasingbelow the Reactor Vessel flange, or below the planned RCS water level for the given evolution (ifthe planned RCS water level is already below the Reactor Vessel flange), warrants declaration of aUE due to the reduced RCS inventory that is available to keep the core covered.Continued loss of RCS Inventory will result in escalation to the Alert emergency classification levelvia either EAL CA3.1 or EAL CA4.1.This EAL addresses conditions in the refueling mode when normal means of core temperatureindication and RPV water level indication may not be available. Redundant means of RPV waterlevel indication will normally be installed (including the ability to monitor level visually) to assurethat the ability to monitor level will not be interrupted. However, if all level indication were to be lostduring a loss of RPV inventory event, the operators would need to determine that RPV inventoryloss was occurring by observing sump and tank level changes. Sump and tank level increasesmust be evaluated against other potential sources of leakage such as cooling water sources insidethe containment to ensure they are indicative of RCS leakage.NMPI Basis Reference(s):1. S-ODP-OPS-01 10 Containment Leakage Evaluation2. USAR 1.4 Primary Coolant Leakage3. Annunciator H2 1 DRYWELL FLOOR DRAIN LEVEL-HIGH4. Annunciator H2 7 DRYWELL WATER LEAK DETECTION SYS5. Annunciator H2 1 R BLDG FL DR SUMPS 11-16 AREA WTR LVL LEVEL HIGH6. Annunciator H2 2 R BUILDING EQUIP DRAIN LEVEL-HIGH7. NEI 99-01 IC CU2Page 137 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 4 -RCS TemperatureInitiating Condition: Inability to maintain plant in cold shutdownEAL:CA4.1 AlertAn UNPLANNED event results in EITHER:RCS temperature > 212°F for > Table C-4 durationORRPV pressure increase > 10 psi due to an UNPLANNED loss of decay heat removalcapabilityTable C-4 RCS Reheat Duration ThresholdsCONTAINMENTRCS Status COSURE DurationCLOSURE StatusINTACT N/A 60 min.*Established 20 min.*Not INTACTNot established 0 min.* If an RCS heat removal system is in operation within this timeframe and RCS temperature is being reduced, the EAL is notapplicable.Mode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificSeveral instruments are capable of providing indication of RCS temperature with respect tothe Technical Specification cold shutdown temperature limit (2120F). These include (ref. 2,3):Recirc operating -Recirc pump temperature points:o 11-RRP-A427o 12-RRP -A431o 13-RRP -A435Page 138EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basiso 14-RRP -A439o 15-RRP- A443, Shutdown cooling operating -Temperature Recorder 38-146If Rx Recirc or Shutdown Cooling pumps are not in operation and reactor coolanttemperature is greater than or equal to 212'F, RCS temperature can be obtained byconverting the RPV pressure to temperature using the saturated steam tables.The RCS should be considered INTACT when the RCS pressure boundary is in its normalcondition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteria governingPrimary and Secondary Containment integrity. If the Technical Specification criteria aremet, therefore, CONTAINMENT CLOSURE has been established. (ref. 4, 5)The pressure rise of greater than 10 psig infers an RCS temperature in excess of theTechnical Specification cold shutdown limit (2120F) for which this EAL would otherwisepermit up to sixty minutes to restore RCS cooling before declaration of an Alert (RCSINTACT). This EAL therefore covers situations in which it is determined that, due to highdecay heat loads, the time provided to reestablish temperature control should be less thansixty minutes (as indicated by significant RCS re-pressurization).Wide range pressure indication (0-1600 psig) is capable of measuring pressure changes of10 psig.If RCS temperature exceeds 2120F, an operating mode change occurs. Although the eventmay have originated in cold conditions, the emergency classification shall be based on theoperating mode that existed at the time the event occurred (prior to any protective systemor operator action initiated in response to the condition). For events that occur in ColdShutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for modeapplicability, even if Hot Shutdown (or a higher mode) is entered during any subsequentPage 139 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisheat-up. In particular, the fission product barrier EALs are applicable only to events thatinitiate in Hot Shutdown or higher.Escalation to a Site Area Emergency would be under EAL CS3.1 should boiling result insignificant RPV water level loss leading to core uncovery.GenericThe RCS Reheat Duration Thresholds table addresses complete loss of functions required for corecooling for greater than 60 minutes during refuel and cold shutdown modes when RCS integrity isestablished. The 60 minute time frame should allow sufficient time to restore cooling without therebeing a substantial degradation in plant safety.The RCS Reheat Duration Thresholds table also addresses the complete loss of functions requiredfor core cooling for greater than 20 minutes during Refuel and cold shutdown modes whenCONTAINMENT CLOSURE is established but RCS integrity is not established or RCS inventory isreduced. The allowed 20 minute time frame was included to allow operator action to restore theheat removal function, if possible.Finally, complete loss of functions required for core cooling during Refuel and cold shutdownmodes when neither CONTAINMENT CLOSURE nor RCS integrity are established is ,addressed.No delay time is allowed because the evaporated reactor coolant that may be released into theContainment during this heatup condition could also be directly released to the environment.The note (*) indicates that this EAL is not applicable if actions are successful in restoring an RCSheat removal system to operation and RCS temperature is being reduced within the specified timeframe.The 10 psig pressure increase addresses situations where, due to high decay heat loads, the timeprovided to restore temperature control, should be less than 60 minutes. The RPV pressuresetpoint was chosen because it is the lowest pressure that the site can read on installed ControlBoard instrumentation that is equal to or greater than 10 psig.Escalation to Site Area Emergency would be via EAL CS3.1 should boiling result in significant RPVwater level loss leading to core uncovery.A loss of Technical Specification components alone is not intended to constitute an Alert. Thesame is true of a momentary UNPLANNED excursion above the Technical Specification coldshutdown temperature limit when the heat removal function is available.The Emergency Director must remain alert to events or conditions that lead to the conclusion thatexceeding the EAL is IMMINENT. If, in the judgment of the Emergency Director, an IMMINENTsituation is at hand, the classification should be made as if the threshold has been exceeded.NMPI Basis Reference(s):1. NMP1 Technical Specifications Definitions 1.12. N1-OP-43C Plant Shutdown3. N1-OP-4 Shutdown Cooling System4. NIP-OUT-01 Shutdown SafetyPage 140 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis5. NMP1 Technical Specifications Definitions 1.11 and 1.126. NEI 99-01 IC CA4Page 141EPMP-EPP-0101Rev 00 Draft A. -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 4- RCS TemperatureInitiating Condition: UNPLANNED loss of decay heat removal capabilityEAL:CU4.1 Unusual EventUNPLANNED event results in RCS temperature > 212°FMode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificSeveral instruments are capable of providing indication of RCS temperature with respect tothe Technical Specification cold shutdown temperature limit (2120F). These include (ref. 2,3):* Recirc operating -Recirc pump temperature points:o 11-RRP -A427o 12-RRP -A431o 13-RRP -A435o 14-RRP -A439o 15-RRP -A443* Shutdown cooling operating -Temperature Recorder 38-146If Rx Recirc or Shutdown Cooling pumps are not in operation and reactor coolanttemperature is greater than or equal to 212'F, RCS temperature can be obtained byconverting the RPV pressure to temperature using the saturated steam tables.If RCS temperature exceeds 2120F, an operating mode change occurs. Although the eventmay have originated in cold conditions, the emergency classification shall be based on theoperating mode that existed at the time the event occurred (prior to any protective systemor operator action initiated in response to the condition). For events that occur in ColdShutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for modeapplicability, even if Hot Shutdown (or a higher mode) is entered during any subsequentPage 142 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisheat-up. In particular, the fission product barrier EALs are applicable only to events thatinitiate in Hot Shutdown or higher.GenericThis EAL is a precursor of more serious conditions and, as a result, is considered to be a potentialdegradation of the level of safety of the plant. In cold shutdown the ability to remove decay heatrelies primarily on forced cooling flow. Operation of the systems that provide this forced coolingmay be jeopardized due to the unlikely loss of electrical power or RCS inventory. Since the RCSusually remains INTACT in the cold shutdown mode a large inventory of water is available to keepthe core covered.During refueling the level in the RPV will normally be maintained above the RPV flange. Refuelingevolutions that decrease water level below the RPV flange are carefully planned and procedurallycontrolled. Loss of forced decay heat removal at reduced inventory may result in more rapidincreases in RCS/RPV temperatures depending on the time since shutdown.Normal means of core temperature indication and RPV water level indication may not be availablein the Refuel mode. Redundant means of RPV water level indication are therefore procedurallyinstalled to assure that the ability to monitor level will not be interrupted. Escalation to Alert wouldbe via EAL CA3.1 based on an inventory loss or EAL CA4.1 based on exceeding its temperatureduration or pressure criteria.NMP1 Basis Reference(s):1. NMP1 Technical Specifications Definitions 1.12. N1-OP-43C Plant Shutdown3. N1-OP-4 Shutdown Cooling System4. NEI 99-01 IC CU4Page 143 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 4- RCS TemperatureInitiating Condition: UNPLANNED loss of decay heat removal capabilityEAL:CU4.2 Unusual EventLoss of all RCS temperature and RPV water level indication for >_ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeMode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificSeveral instruments are capable of providing indication of RCS temperature with respect tothe Technical Specification cold shutdown temperature limit (2120F). These include (ref. 2,3):" Recirc operating -Recirc pump temperature points:o 11-RRP -A427o 12-RRP -A431o 13-RRP -A435o 14-RRP -A439o 15-RRP -A443" Shutdown cooling operating -Temperature Recorder 38-146If Rx Recirc or Shutdown Cooling pumps are not in operation and reactor coolanttemperature is greater than or equal to 2120F, RCS temperature can be obtained byconverting the RPV pressure to temperature using the saturated steam tables.RPV water level is monitored from the bottom of active fuel to the top of the RPV head toensure adequate coverage for expected and postulated conditions of RPV water level.RPV water level measurement is derived by the differential pressure that exists between areference leg and variable leg. All instruments are referenced to a benchmark at 439 in.Page 144 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisabove the inside bottom head of the reactor vessel with the exception of the Suppressedor Flange Level range. This benchmark corresponds to 84 in. (rounded from 84 7/16")above the top of active fuel and is the 0 in. reference indication on the RPV water levelinstruments (except the suppressed or flange level). RPV water level monitoring issubdivided into five ranges identified as:" Control range provides indication and control signals for normal plant operation andprotection system actuation.* Lo-lo-lo range provides indication and control signals for transient conditions belowthe normal operating band and emergency equipment actuation.* Wide range provides indication for transient conditions above normal operatingband.* Suppressed or flange level provides indication relative to the reactor vessel flange.* Fuel Zone provides indication for long term accident conditions where RPV waterlevel cannot be restored.In preparation for refueling operations, the Wide and Suppressed RPV water levelinstruments are modified to provide continuous level indication from within the RPV to thereactor cavity.Although the event may have originated in cold conditions, the emergency classificationshall be based on the operating mode that existed at the time the event occurred (prior toany protective system or operator action initiated in response to the condition). For eventsthat occur in Cold Shutdown or Refuel, escalation is via EALs that have Cold Shutdown orRefuel for mode applicability, even if Hot Shutdown (or a higher mode) is entered duringany subsequent heat-up. In particular, the fission product barrier EALs are applicable onlyto events that initiate in Hot Shutdown or higher.GenericThis EAL is a precursor of more serious conditions and, as a result, is considered to be a potentialdegradation of the level of safety of the plant. In cold shutdown the ability to remove decay heatrelies primarily on forced cooling flow. Operation of the systems that provide this forced coolingmay be jeopardized due to the unlikely loss of electrical power or RPV inventory. Since the RCSusually remains INTACT in the cold shutdown mode a large inventory of water is available to keepthe core covered.Page 145 EPMP-EPP-0101Rev 00 Draft A Attachment I -Emergency Action Level Technical BasisDuring refueling the level in the RPV will normally be maintained above the RPV flange. Refuelingevolutions that decrease water level below the RPV flange are carefully planned and procedurallycontrolled. Loss of forced decay heat removal at reduced inventory may result in more rapidincreases in RPV temperatures depending on the time since shutdown.Normal means of core temperature indication and RPV water level indication may not be availablein the refueling mode. Redundant means of RPV water level indication are therefore procedurallyinstalled to assure that the ability to monitor level will not be interrupted. However, if all level andtemperature indication were to be lost in either the cold shutdown of refueling modes, this EALwould result in declaration of a UE if both temperature and level indication cannot be restoredwithin 15 minutes from the loss of both means of indication. Escalation to Alert would be via EALCA3.1 based on an inventory loss or EAL CA4.1 based on exceeding its temperature criteria.NMP1 Basis Reference(s):1. NMP1 Technical Specifications Definitions 1.12. Nl-OP-43C Plant Shutdown3. N1-OP-4 Shutdown Cooling System4. NEI 99-01 IC CU4Page 146EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 5 -Inadvertent CriticalityInitiating Condition: Inadvertent criticalityEAL:CU5.1 Unusual EventAn UNPLANNED sustained positive period observed on nuclear instrumentationMode Applicability:3 -Cold Shutdown, 4 -RefuelBasis:Plant-SpecificThe term "sustained" is used to allow exclusion of expected short-term positive periodsfrom planned fuel bundle or control rod movements during core alteration. These short-term positive periods are the result of the rise in neutron population due to subcriticalmultiplication.GenericThis EAL addresses criticality events that occur in Cold Shutdown or Refueling modes such as fuelmis-loading events and inadvertent dilution events. This EAL indicates a potential degradation ofthe level of safety of the plant, warranting a UE classification.Escalation would be by Emergency Director judgment.NMP1 Basis Reference(s):1. NEI 99-01 IC CU8Page 147 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisC -Cold Shutdown / Refueling System Malfunction6 -CommunicationsCategory:Subcategory:Initiating Condition: Loss of all onsite or offsite communications capabilitiesEAL:CU6.1 Unusual EventLoss of all Table C-5 onsite (internal) communication methods affecting the ability toperform routine operationsORLoss of all Table C-5 offsite (external) communication methods affecting the ability toperform offsite notificationsTable C-5 Communications SystemsSystem Onsite Offsite(internal) (external)PBX (normal dial telephones)GaitronicsHand-Held Portable Radio (station radio)Control Room installed satellite phones (non portable)ENSRECSXXXXXXXMode Applicability:3 -Cold Shutdown, 4 -Refuel, D -DefueledBasis:Plant-SpecificOnsite/offsite communications systems are listed in Table C-2 (ref. 1, 2).This EAL is the cold condition equivalent of the hot condition EAL SU6.1.GenericThe purpose of this EAL is to recognize a loss of communications capability that either defeats theplant operations staff ability to perform routine tasks necessary for plant operations or the ability toPage 148EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basiscommunicate issues with off-site authorities. The loss of off-site communications ability is expectedto be significantly more comprehensive than the condition addressed by 10 CFR 50.72.The availability of one method of ordinary off-site communications is sufficient to inform federal,state, and local authorities of plant issues. This EAL is intended to be used only whenextraordinary means (e.g., relaying of information from radio transmissions, individuals being sentto off-site locations, etc.) are being utilized to make communications possible.NMP1 Basis Reference(s):1. USAR 2.4.5 Lighting and Communication2. Nine Mile Point Nuclear Station Site Emergency Plan, Section 7.23. NEI 99-01 IC CU6Page 149EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory S -System MalfunctionEAL Group: Hot Conditions (RCS temperature > 2120F);EALs in this category are applicable only inone or more hot operating modes.Numerous system-related equipment failure events that warrant emergency classificationhave been identified in this category. They may pose actual or potential threats to plantsafety.The events of this category pertain to the following subcategories:1. Loss of AC PowerLoss of emergency plant electrical power can compromise plant safety systemoperability including decay heat removal and emergency core cooling systems whichmay be necessary to ensure fission product barrier integrity. This category includesloss of onsite and offsite power sources for the 4.16 kV safeguard buses.2. Loss of DC PowerLoss of emergency plant electrical power can compromise plant safety systemoperability including decay heat removal and emergency core cooling systems whichmay be necessary to ensure fission product barrier integrity. This category includesloss of power to the 125 VDC buses.3. Criticality & RPS FailureInadvertent criticalities pose potential personnel safety hazards as well being indicativeof losses of reactivity control.Events related to failure of the Reactor Protection System (RPS) to initiate andcomplete reactor trips. In the plant licensing basis, postulated failures of the RPS tocomplete a reactor trip comprise a specific set of analyzed events referred to asAnticipated Transient Without Scram (ATWS) events. For EAL classification however,ATWS is intended to mean any trip failure event that does not achieve reactorshutdown. If RPS actuation fails to assure reactor shutdown, positive control ofreactivity is at risk and could cause a threat to fuel clad, RCS and Containmentintegrity.Page 150 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis4. Inability to Reach or Maintain Shutdown ConditionsSystem malfunctions may lead to failure of the plant to be brought to the required plantoperating condition required by technical specifications if a limiting condition foroperation (LCO) is not met.5. InstrumentationCertain events that degrade plant operator ability to effectively assess plant conditionswithin the plant warrant emergency classification. Losses of annunciators are in thissubcategory.6. CommunicationsCertain events that degrade plant operator ability to effectively communicate withessential personnel within or external to the plant warrant emergency classification.7. Fuel Clad DegradationDuring normal operation, reactor coolant fission product activity is very low. Smallconcentrations of fission products in the coolant are primarily from the fission of trampuranium in the fuel clad or minor perforations in the clad itself. Any significant increasefrom these base-line levels (-5% clad failures) is indicative of fuel failures and iscovered under Category F, Fission Product Barrier Degradation. However, lesseramounts of clad damage may result in coolant activity exceeding TechnicalSpecification limits. These fission products will be circulated with the reactor coolantand can be detected by coolant sampling and/or the Letdown radiation monitor.8. RCS LeakageThe Reactor Vessel provides a volume for the coolant that covers the reactor core. TheReactor Vessel and associated pressure piping (reactor coolant system) togetherprovide a barrier to limit the release of radioactive material should the reactor fuel cladintegrity fail.Excessive RCS leakage greater than Technical Specification limits are utilized toindicate potential pipe cracks that may propagate to an extent threatening fuel clad,RCS and Containment integrity.Page 151 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 1 -Loss of AC PowerInitiating Condition: Prolonged loss of all offsite and all onsite AC power to 4.16 kVemergency busesEAL:SGI.1 General EmergencyLoss of all offsite and all onsite AC power, Table S-1, to 4.16 kV emergency busesAND EITHER:Restoration of at least one 4.16 kV emergency bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likelyORRPV water level cannot be restored and maintained above -84 in. or RPV waterlevel cannot be determinedTable S-1 AC Power Sources* DG 102* DG 10300 T-101N0 T-101SO 0 T-10 backfed from offsite through T-1or T-2 (only if already aligned)Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificNMP1 4.16 kV emergency buses are buses PB102 and PB103, which feed all Stationredundant safety-related loads.. There are three offsite power sources available to thesebuses (ref. 1):* Offsite 115 kV through transformer 101 N. This is the normal power supply toPB102.Page 152 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis" Offsite 115 kV through transformer 101S. This is the normal power supply toPB103.* Offsite 345 kV through transformer T-1 or T-2 backfed through transformer T-1 0.Based on operational experience, if the backfeed is not already aligned, this cannot beconsidered available/capable of supplying the bus due to the time it will take to align it.There are two onsite AC power sources:" DG102forPB102" DG103 for PB103Consideration should be given to operable loads necessary to remove decay heat orprovide RPV makeup capability when evaluating loss of all AC power to emergency buses.Even though an emergency bus may be energized, if necessary loads (i.e., loads that iflost would inhibit decay heat removal capability or RPV makeup capability) are notoperable on the energized bus then the bus should not be considered operable.Four hours is the station blackout coping period (ref. 2).An RPV water level instrument reading of -84 in. (rounded from 84 7/16") indicates RPVwater level is at the top of active fuel. When RPV water level is at or above the top ofactive fuel, the core is completely submerged. Core submergence is the most desirablemeans of core cooling. When RPV water level is below the top of active fuel, theuncovered portion of the core must be cooled by less reliable means (i.e., steam cooling orspray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme,RPV water level control measures in order to restore and maintain adequate core cooling(ref. 3). Since core uncovery begins if RPV water level drops to -84 in., the level isindicative of a challenge to core cooling and the Fuel Clad barrier.Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that RPV water level cannot be restored and maintained above the top ofactive fuel may be made at, before, or after RPV water level actually decreases to thispoint.GenericPage 153 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisLoss of all AC power to emergency busses compromises all plant safety systems requiring electricpower including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolongedloss of all AC power to emergency busses will lead to loss of fuel clad, RCS, and containment, thuswarranting declaration of a General Emergency.This EAL is specified to assure that in the unlikely event of a prolonged loss of all AC power toemergency 4.16 kV buses, timely recognition of the seriousness of the event occurs and thatdeclaration of a General Emergency occurs as early as is appropriate, based on a reasonableassessment of the event trajectory.The likelihood of restoring at least one emergency bus should be based on a realistic appraisal ofthe situation since a delay in an upgrade decision based on only a chance of mitigating the eventcould result in a loss of valuable time in preparing and implementing public protective actions.In addition, under these conditions, fission product barrier monitoring capability may be degraded.NMP1 Basis Reference(s):1. N1-OP-30 4.16 KV, 600V, and 480V House Service2. NER-1M-025 SBO Evaluation3. NER-1M-095-R02 NMP1 EOP and SAP Basis Document4. NEI 99-01 IC SG1Page 154EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 1 -Loss of AC PowerInitiating Condition: Loss of all offsite and all onsite AC power to 4.16 kV emergencybuses for __ 15 min.EAL:SSl.1 Site Area EmergencyLoss of all offsite and all onsite AC power, Table S-1, to 4.16 kV emergency busesfor __ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table S-1 AC Power Sources= DG 102i DG 1030* T-101N* T-101SO ° T-1 0 backfed from offsite through T-1or T-2 (only if already aligned)Mode Applicability:1 -Power Operation,Basis:Plant-Specific2 -Hot ShutdownNMP1 4.16 kV emergency buses are buses PB102 and PB103, which feed all Stationredundant safety-related loads. There are three offsite power sources available to thesebuses (ref. 1, 2):" Offsite 115 kV through transformer 101 N. This is the normal power supply toPB102." Offsite 115 kV through transformer 101S. This is the normal power supply toPB103.Page 155EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis0 Offsite 345 kV through transformer T-1 or T-2 backfed through transformer T-1 0.Based on operational experience, if the backfeed is not already aligned, this cannot beconsidered available/capable of supplying the bus due to the time it will take to align it. Inany case, if this cannot be accomplished within 15 minutes, it is not available and anUnusual Event must be declared.There are two onsite AC power sources:" DG102forPB102* DG103 for PB103Consideration should be given to operable loads necessary to remove decay heat orprovide RPV makeup capability when evaluating loss of all AC power to vital buses. Eventhough an essential bus may be energized, if necessary loads (i.e., loads that if lost wouldinhibit decay heat removal capability or RPV makeup capability) are not operable on theenergized bus then the bus should not be considered operable.The fifteen-minute interval was selected as a threshold to exclude transient power losses.GenericLoss of all AC power to emergency busses compromises all plant safety systems requiring electricpower including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolongedloss of all AC power to emergency 4.16 kV buses will lead to loss of Fuel Clad, RCS, andContainment, thus this event can escalate to a General Emergency.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-sitepower.Escalation to General Emergency is via EALs in Category F or EAL SG1.1.NMP1 Basis Reference(s):1. N1-OP-30 4.16 kV, 600V, and 480V House Service2. USAR section IX Electrical Systems3. NEI 99-01 IC SS1Page 156 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:Initiating Condition:S -System Malfunction1 -Loss of AC PowerAC power capability to 4.16 kV emergency buses reduced to asingle power source for 15 min. such that ANY additional singlefailure would result in a complete loss of all 4.16 kV emergencybus powerEAL:SAI.1 AlertAC power capability to 4.16 kV emergency buses reduced to a single power source, TableS-1, for >_ 15 min. (Note 4)ANDANY additional single power source failure will result in a loss of all 4.16 kV emergencybus powerNote 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table S-1 AC Power Sources* DG 102U,* DG 10300 T-101N0 T-1 01S0 0 T-10 backfed from offsite through T-1or T-2 (only if already aligned)Mode Applicability:1 -Power Operation, 2Basis:Plant-Specific-Hot ShutdownNMP1 4.16 kV emergency buses are buses PB102 and PB103, which feed all Stationredundant safety-related loads.. There are three offsite power sources available to thesebuses:Page 157 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis" Offsite 115 kV through transformer 101N. This is the normal power supply toPB1302." Offsite 115 kV through transformer 101S. This is the normal power supply toPB103." Offsite 345 kV through transformer T-1 or T-2 backfed through transformer T-1 0.Based on operational experience, if the backfeed is not already aligned, this cannot beconsidered available/capable of supplying the bus due to the time it will take to align it. Inany case, if this cannot be accomplished within 15 minutes, it is not available and anUnusual Event must be declared.There are two onsite AC power sources:" DG102forPB102" DG103 for PB103The fifteen-minute interval was selected as a threshold to exclude transient power losses.If the capability for multiple sources to energize the unit emergency buses within 15minutes is not restored, an Alert is declared under this EAL. The subsequent loss of thesingle remaining power source escalates the event to a Site Area Emergency under EALSS1.1.GenericThe condition indicated by this EAL is the degradation of the off-site and on-site AC power systemssuch that any additional single failure would result in a complete loss of 4.16 kV emergency busAC power. This condition could occur due to a loss of off-site power with a concurrent failure of allbut one emergency generator to supply power to its emergency busses. Another related conditioncould be the loss of all off-site power and loss of on-site emergency generators with only one trainof 4.16 kV emergency buses being backfed from the unit main generator, or the loss of on-siteemergency generators with only one train of 4.16 kV emergency buses being backfed from off-sitepower. The subsequent loss of this single power source would escalate the event to a Site AreaEmergency in accordance with EAL SS1.1.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.NMP1 Basis Reference(s):1. N1-OP-30 4.16 kV, 600V, and 480V House Service2. USAR section IX Electrical Systems3. NEI 99-01 IC SA5Page 158 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 1 -Loss of AC PowerInitiating Condition: Loss of all offsite AC power to 4.16 kV emergency buses for >_ 15min.EAL:SUl.1 Unusual EventLoss of all offsite AC power, Table S-1, to 4.16 kV emergency buses for __ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table S-1 AC Power Sources* DG102* DG 1030T-101N4)0 T-1 01No 9 T-10 backfed from offsite through T-1or T-2 (only if already aligned)Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificNMP1 4.16 kV emergency buses are buses PB102 and PB103, which feed all Stationredundant safety-related loads.. There are three offsite power sources available to thesebuses (ref. 1, 2):" Offsite 115 kV through transformer 101 N. This is the normal power supply toPB102.* Offsite 115 kV through transformer 101S. This is the normal power supply toPB103.Page 159 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis0 Offsite 345 kV through transformer T-1 or T-2 backfed through transformer T-1 0.Based on operational experience, if the backfeed is not already aligned, this cannot beconsidered available/capable of supplying the bus due to the time it will take to align it. Inany case, if this cannot be accomplished within 15 minutes, it is not available and anUnusual Event must be declared.There are two onsite AC power sources:" DG102forPB102," DG103 for PB103The fifteen-minute interval was selected as a threshold to exclude transient power losses.GenericProlonged loss of off-site AC power reduces required redundancy and potentially degrades thelevel of safety of the plant by rendering the plant more vulnerable to a complete loss of AC powerto emergency busses.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-sitepower.NMP1 Basis Reference(s):1. Nl-OP-30 4.16 kV, 600V, and 480V House Service2. USAR section IX Electrical Systems3. NEI 99-01 IC SU1Page 160EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 2 -Loss of DC PowerInitiating Condition: Loss of all emergency DC power for > 15 min.EAL:SS2.1 Site Area Emergency< 106 VDC on both Battery Board 11 and Battery Board 12 for > 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificA Safety Related (SR) system and a Quality Related (QR) system comprise the 125 VDCPower System. The two SR 125 VDC systems (Battery board 11 and Battery board 12)each consist of: one battery, two Static Chargers in parallel, and a DC distribution board.The one QR 125 VDC system consists of: a battery, one Static Charger, and one batteryboard. This EAL addresses only the Safety Related battery boards (ref. 1, 2).106 VDC is the minimum voltage for battery operability (ref.3).This EAL is the hot condition equivalent of the cold condition loss of DC powerEAL CU2.1.GenericLoss of all DC power compromises ability to monitor and control plant safety functions. Prolongedloss of all DC power will cause core uncovering and loss of containment integrity when there issignificant decay heat and sensible heat in the reactor system.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation to a General Emergency would occur by EALs in Category R and Category F.NMP1 Basis Reference(s):1. N1-OP-47A 125 VDC Power System2. USAR section IX Electrical Systems3. NMP1 Technical Specification 3.6.3Page 161 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis4. NEI 99-01 IC SS3Page 162EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 3 -Criticality & RPS FailureInitiating Condition: Automatic scram and all manual actions fail to shut down thereactor and indication of an extreme challenge to the ability to coolthe core existsEAL:SG3.1 General EmergencyAn automatic scram fails to shut down the reactor as indicated by reactor power > 6%ANDAll manual actions fail to shut down the reactor as indicated by reactor power > 6%AND EITHER of the following exist or have occurred:RPV water level cannot be restored and maintained above -109 in. or RPV waterlevel cannot be determinedORTorus temperature and RPV pressure cannot be maintained below the HeatCapacity Temperature Limit (N1-EOP-4 Figure M)Mode Applicability:1 -Power OperationBasis:Plant-SpecificThis EAL addresses the following:" Any automatic reactor scram signal followed by a manual scram that fails to shutdown the reactor to an extent the reactor is producing energy in excess of the heatload for which the safety systems were designed (EAL SS3.1), and" Indications that either core cooling is extremely challenged or heat removal isextremely challenged.Reactor shutdown achieved by use of the alternate control rod insertion methods of N1-EOP-3.1 is also credited as a successful manual scram provided reactor power can bereduced below the APRM downscale trip setpoint before indications of an extremechallenge to either core cooling or heat removal exist (ref. 3).Page 163 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisThe APRM downscale trip setpoint (6%) is a minimum reading on the power range scalethat indicates power production (ref. 1, 2). It also approximates the decay heat which theshutdown systems were designed to remove and is indicative of a condition requiringimmediate response to prevent subsequent core damage. At or below the APRMdownscale trip setpoint, plant response will be similar to that observed during a normalshutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters(e.g., number of open SRVs, number of open main turbine bypass valves, main steamflow, RPV pressure and suppression pool temperature trend, etc.) can be used todetermine if reactor power is greater than 6% power (ref. 2).The combination of failure of both front line and backup protection systems to function inresponse to a plant transient, along with the continued production of heat, poses a directthreat to the Fuel Clad and RCS barriers.By definition, an operating mode change occurs when the Mode Switch is moved from thestartup or run position to the shutdown position. The plant operating mode that existed atthe time the event occurs (i.e., Power Operation), however, requires emergencyclassification of at least an Alert. The operating mode change associated with movementof the Mode Switch, by itself, does not justify failure to declare an emergency for ATWSevents.Indication that core cooling is extremely challenged is manifested by:RPV water level cannot be restored and maintained above -109 in. (ref. 1, 2). TheMinimum Steam Cooling RPV Water Level (MSCRWL) is the lowest RPV waterlevel at which the covered portion of the reactor core will generate sufficient steamto preclude any clad temperature in the uncovered portion of the core fromexceeding 1500'F. Consistent with the EOP definition of "cannot be restored andmaintained," the determination that RPV water level cannot be restored andmaintained above the MSCRWL may be made at, before, or after RPV water levelactually decreases to this point.When RPV water level cannot be determined, EOPs require RPV floodingstrategies. RPV water level indication provides the primary means of knowing ifadequate core cooling is being maintained. When all means of determining RPVPage 164 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basiswater level are unavailable, the fuel clad barrier is threatened and reliance onalternate means of assuring adequate core cooling must be attempted. Theinstructions in N1-EOP-7 specify these means, which include emergencydepressurization of the RPV and injection into the RPV at a rate needed to flood tothe elevation of the main steam lines or hold RPV pressure above the MinimumSteam Cooling Pressure (in ATWS events) (ref. 4).The HCTL is the highest torus water temperature from which emergency RPVdepressurization will not raise (ref. 2):o Torus temperature above the design value (205°F), oro Torus pressure above Primary Containment Pressure Limit before the rate ofenergy transfer from the RPV to the containment is greater than the capacityof the containment vent.The HCTL is a function of RPV pressure and torus water level. It is utilized topreclude failure of the containment and equipment in the containment necessary forthe safe shutdown of the plant. Plant parameters in excess of the HCTL could be aprecursor of Primary Containment failure. (ref. 2)The HCTL is given in N1 -EOP-4 Figure M. This threshold is met when RPV BLOWDOWN is required in N1-EOP-4, Step TT-5 (ref. 5). This condition addresses loss offunctions required for hot shutdown with the reactor at pressure and temperature.GenericUnder these conditions, the reactor is producing more heat than the maximum decay heat load forwhich the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful.The reactor should be considered shutdown when it producing less heat than the maximum decayheat load for which the safety systems are designed (6% power). In the event either of thesechallenges exists at a time that the reactor has not been brought below the power associated withthe safety system design a core melt sequence exists. In this situation, core degradation can occurrapidly. For this reason, the General Emergency declaration is intended to be anticipatory of thefission product barrier table declaration to permit maximum off-site intervention time.NMP1 Basis Reference(s):1. N1-EOP-3 Failure to Scram2. NER-1M-095 NMP1 EOP/SAP Basis Document3. N1-EOP-3.1 Alternate Rod InsertionPage 165 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis4. N1-EOP-7 RPV Flooding5. N1-EOP-4 Primary Containment Control6. NEI 99-01 IC SG2Page 166EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis4. N1-EOP-7 RPV Flooding5. N1-EOP-4 Primary Containment Control6. NEI 99-01 IC SG2Page 166EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 3 -Criticality & RPS FailureInitiating Condition: Automatic scram fails to shut down the reactor and manual actionstaken from the reactor control console are not successful inshutting down the reactorEAL:SS3.1 Site Area EmergencyAn automatic scram failed to shut down the reactor as indicated by reactor power > 6%ANDManual actions taken at the reactor control console (mode switch in shutdown, manualscram push buttons and ARI) failed to shut down the reactor as indicated by reactor power> 6%Mode Applicability:1 -Power OperationBasis:Plant-SpecificThis EAL addresses any automatic reactor scram signal followed by a manual scram thatfailed to shut down the reactor to an extent the reactor is producing energy in excess of theheat load for which the safety systems were designed.The first condition of this EAL identifies the need to cease critical reactor operations byactuation of the automatic Reactor Protection System (RPS) scram function. A reactorscram is automatically initiated by the Reactor Protection System (RPS) when certaincontinuously monitored parameters exceed predetermined setpoints. A reactor scram maybe the result of manual or automatic action in response to any of the following conditions(ref. 1):Parameter SetpointHigh level in the Scram Dump 45 galVolumeMSIV closure 10%IRM neutron flux 96% of rangeAPRM neutron flux TS 2.1.2.aPage 167 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisParameter SetpointHigh reactor pressure 1080 psigLow reactor water level +53 inchesHigh drywell pressure 3.5 psigTSV closure 10%Generator load rejection TCV fast closureFollowing a successful reactor scram, rapid insertion of the control rods occurs. Nuclearpower promptly drops to a fraction of the original power level and then decays to a levelseveral decades less with a negative period. The reactor power drop continues untilreactor power reaches the point at which the influence of source neutrons on reactorpower starts to be observable. A predictable post-scram response from an automaticreactor scram signal should therefore consist of a prompt drop in reactor power as sensedby the nuclear instrumentation and a lowering of power into the source range. A successfulscram has therefore occurred when there is sufficient rod insertion from the trip of RPS tobring the reactor power to or below the APRM downscale trip setpoint of 6%. For thepurposes of this EAL, a successful automatic initiation of ARI that reduces reactor power toor below 6% is a not a successful automatic scram. If automatic actuation of ARI hasoccurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI isa backup means of inserting control rods in the unlikely event that an automatic RPSscram signal exists but the reactor continues to generate significant power. (ref. 2, 3)For the purposes of emergency classification at the Site Area emergency level, successfulmanual scram actions are those which can be quickly performed from the reactor controlconsole (i.e., Mode Switch, manual scram pushbuttons, and manual ARI actuation).Reactor shutdown achieved by use of the alternate control rod insertion methods of N1-EOP-3.1 does not constitute a successful manual scram (ref. 4).The APRM downscale trip setpoint (6%) is a minimum reading on the power range scalethat indicates power production (ref. 3). It also approximates the decay heat which theshutdown systems were designed to remove and is indicative of a condition requiringimmediate response to prevent subsequent core damage. At or below the APRMdownscale trip setpoint, plant response will be similar to that observed during a normalPage 168 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisshutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters(e.g., number of open SRVs, number of open main turbine bypass valves, main steamflow, RPV pressure and wetwell temperature trend, etc.) can be used to determine ifreactor power is greater than 6% power.By definition, an operating mode change occurs when the Mode Switch is moved from thestartup or run position to the shutdown position. The plant operating mode that existed atthe time the event occurs (i.e., Power Operation), however, requires emergencyclassification of at least an Alert. The operating mode change associated with movementof the Mode Switch, by itself, does not justify failure to declare an emergency for ATWSevents.Escalation of this event to a General Emergency would be under EAL SG3.1 orEmergency Director judgment.GenericUnder these conditions, the reactor is producing more heat than the maximum decay heat load forwhich the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful.A Site Area Emergency is warranted because conditions exist that lead to IMMINENT loss orpotential loss of both fuel clad and RCS.The reactor should be considered shutdown when it producing less heat than the maximum decayheat load for which the safety systems are designed (6% power).Manual scram actions taken at the reactor control console are any set of actions by the reactoroperator(s) at which causes or should cause control rods to be rapidly inserted into the core andshuts down the reactor.Manual scram actions are not considered successful if action away from the reactor controlconsole is required to scram the reactor. This EAL is still applicable even if actions taken awayfrom the reactor control console are successful in shutting the reactor down because the designlimits of the fuel may have been exceeded or because of the gross failure of the Reactor ProtectionSystem to shutdown the plant.Escalation of this event to a General Emergency would be due to a prolonged condition leading toan extreme challenge to either core-cooling or heat removal.NMPI Basis Reference(s):1. Technical Specifications Table 3.6.2.a2. N1-EOP-2 RPV Control3. N1-EOP-3 Failure to Scram4. N1-EOP-3.1 Alternate Rod InsertionPage 169 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis5. NEI 99-01 IC SS2Page 170EPMP-EPP-01 01Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 3 -Criticality & RPS FailureInitiating Condition: Automatic scram failed to shut down the reactor and the manualactions taken from the reactor control console are successful inshutting down the reactorEAL:SA3.1 AlertAn automatic scram failed to shut down the reactorANDManual actions taken at the reactor control console (mode switch in shutdown, manualscram push buttons or ARI) successfully shut down the reactor as indicated by reactorpower -6%Mode Applicability:1 -Power OperationBasis:Plant-SpecificThe first condition of this EAL identifies the need to cease critical reactor operations byactuation of the automatic Reactor Protection System (RPS) scram function. A reactorscram is automatically initiated by the Reactor Protection System (RPS) when certaincontinuously monitored parameters exceed predetermined setpoints. A reactor scram maybe the result of manual or automatic action in response to various conditions (ref. 1):Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclearpower promptly drops to a fraction of the original power level and then decays to a levelseveral decades less with a negative period. The reactor power drop continues untilreactor power reaches the point at which the influence of source neutrons on reactorpower starts to be observable. A predictable post-scram response from an automaticreactor scram signal should therefore consist of a prompt drop in reactor power as sensedby the nuclear instrumentation and a lowering of power into the source range. A successfulscram has therefore occurred when there is sufficient rod insertion from the trip of RPS tobring the reactor power to or below the APRM downscale trip setpoint of 6%. For thepurposes of this EAL, a successful automatic initiation of ARI that reduces reactor power toPage 171 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basisor below 6% is a not a successful automatic scram. If automatic actuation of ARI hasoccurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI isa backup means of inserting control rods in the unlikely event that an automatic RPSscram signal exists but the reactor continues to generate significant power. (ref. 2, 3)For the purposes of emergency classification at the Alert level, successful manual scramactions are those which can be quickly performed from the reactor control console (i.e.,mode switch, manual scram pushbuttons, and manual ARI actuation). Reactor shutdownachieved by use of the alternate control rod insertion methods of N1-EOP-3.1 does notconstitute a successful manual scram (ref. 4).Following any automatic RPS scram signal EOPs prescribe insertion of redundant manualscram signals to back up the automatic RPS scram function and ensure reactor shutdownis achieved. Even if the first subsequent manual scram signal inserts all control rods to thefull-in position immediately after the initial failure of the automatic scram, the lowest level ofclassification that must be declared is an Alert.If the operator determines the reactor must be scrammed before one of the RPS setpointsis reached, procedures require that the Mode Switch first be placed in the shutdownposition. Although manipulation of the Mode Switch is a manual action, the RPS logictrains are actuated as with an automatic RPS-initiated scram. If reactor power remainsabove the APRM downscale trip setpoint after the Mode Switch is placed in shutdown,RPS has failed and, as a minimum, an Alert emergency declaration is required. Ifsubsequent actuation of the reactor scram pushbuttons and manual initiation of ARI do notreduce reactor power to or below the APRM downscale trip setpoint, a Site AreaEmergency declaration is required under EAL SS3.1.In the event that the operator identifies a reactor scram is IMMINENT and initiates asuccessful manual reactor scram before the automatic scram setpoint is reached, nodeclaration is required. The successful manual scram of the reactor before it reaches itsautomatic scram setpoint or reactor scram signals caused by instrumentation channelfailures do not lead to a potential fission product barrier loss. If manual reactor scramactions fail to reduce reactor power to or below 6%, the event escalates to the Site AreaEmergency under EAL SS3.1.Page 172 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisBy procedure, operator actions include the initiation of an immediate manual scramfollowing receipt of an automatic scram signal. If there are no clear indications that theautomatic scram failed (such as a time delay following indications that a scram setpointwas exceeded), it may be difficult to determine if the reactor was shut down because ofautomatic scram or manual actions. If a subsequent review of the scram actuationindications reveals that the automatic scram did not cause the reactor to be shut down,consideration should be given to evaluating the fuel for potential damage and the reportingrequirements of 50.72 should be considered for the transient event.By definition, an operating mode change occurs when the Mode Switch is moved from thestartup or run position to the shutdown position. The plant operating mode that existed atthe time the event occurs (i.e., Power Operation), however, requires emergencyclassification of at least an Alert. The operating mode change associated with movementof the Mode Switch, by itself, does not justify failure to declare an emergency for ATWSevents.GenericThe reactor should be considered shutdown when it producing less heat than the maximum decayheat load for which the safety systems are designed (6% power).Manual scram actions taken at the reactor control console are any set of actions by the reactoroperator(s) which causes or should cause control rods to be rapidly inserted into the core andshuts down the reactor.This condition indicates failure of the automatic protection system to scram the reactor. Thiscondition is more than a potential degradation of a safety system in that a front line automaticprotection system did not function in response to a scram signal. Thus the plant safety has beencompromised because of the failure of RPS to automatically shut down the plant. An Alert isindicated because conditions may exist that lead to potential loss of fuel clad barrier or RCS barrierand because of the failure of the Reactor Protection System to automatically shut down the plant.If manual actions taken at the reactor control console fail to shut down the reactor, the event wouldescalate to a Site Area Emergency.NMP1 Basis Reference(s):1. Technical Specifications Table 3.6.2.a2. N1-EOP-2 RPV Control3. N1-EOP-3 Failure to Scram4. N1-EOP-3.1 Alternate Rod Insertion5. NEI 99-01 IC SA2Page 173 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisPage 174 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:S -System Malfunction3 -Criticality & RPS FailureInitiating Condition: Inadvertent criticalityEAL:SU3.1 Unusual EventAn UNPLANNED sustained positive period observed on nuclear instrumentationMode Applicability:2 -Hot ShutdownBasis:Plant-SpecificThe term "sustained" is used to allow exclusion of expected short-term positive periodsfrom planned fuel bundle or control rod movements during core alteration. These short-term positive periods are the result of the rise in neutron population due to subcriticalmultiplication.GenericThis EAL addresses inadvertent criticality events. While the primary concern of this EAL iscriticality This EAL addresses inadvertent criticality events. This EAL indicates a potentialdegradation of the level of safety of the plant, warranting a UE classification. This EAL excludesinadvertent criticalities that occur during planned reactivity changes associated with reactorstartups (e.g., criticality earlier than estimated).Escalation would be by EALs in Category F, as appropriate to the operating mode at the time ofthe event.NMP1 Basis Reference(s):1. NEI 99-01 IC SU8Page 175EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisS -System Malfunction4 -Inability to Reach or Maintain Shutdown ConditionsCategory:Subcategory:Initiating Condition: Inability to reach required shutdown within Technical SpecificationlimitsEAL:SU4.1 Unusual EventPlant is not brought to required operating mode within Technical Specifications LCOrequired action completion timeMode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificLimiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safeoperation of the unit. The actions associated with an LCO state conditions that typicallydescribe the ways in which the requirements of the LCO can fail to be met. Specified witheach stated condition are required action completion times. (ref. 1)GenericLimiting Conditions of Operation (LCOs) require the plant to be brought to a required operatingmode when the Technical Specification required configuration cannot be restored. Depending onthe circumstances, this may or may not be an emergency or precursor to a more severe condition.In any case, the initiation of plant shutdown required by the site Technical Specifications requires afour hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safetyenvelope when being shut down within the allowable required action completion time in theTechnical Specifications. An immediate UE is required when the plant is not brought to therequired operating mode within the allowable required action completion time in the TechnicalSpecifications. Declaration of a UE is based on the time at which the LCO-specified required actioncompletion time period elapses under the site Technical Specifications and is not related to howlong a condition may have existed.NMP1 Basis Reference(s):1. Nine Mile Point Unit ITechnical Specifications2. NEI 99-01 IC SU2Page 176EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:S -System Malfunction5 -InstrumentationInitiating Condition: Inability to monitor a significant transient in progressEAL:SS5.1 Site Area EmergencyLoss of > approximately 75% of annunciation or indication on Control Room panels L, K,H, F and G for > 15 min. (Note 4)ANDA significant transient is in progress, Table S-2ANDCompensatory indications are unavailable (Plant Process Computer, SPDS)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeTable S-2 Significant Transients" Turbine runback > 25% thermal reactor power* Electric load rejection > 25% full electrical load* Reactor scram" ECCS injection* Thermal power oscillations > 10%Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificPlant computer and SPDS are considered compensatory indication (ref. 1).Significant transients are listed in Table S-2.GenericThis EAL is intended to recognize the threat to plant safety associated with the complete loss ofcapability of the control room staff to monitor plant response to a significant transient.Page 177EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis"Planned" and "UNPLANNED" actions are not differentiated since the loss of instrumentation of thismagnitude is of such significance during a transient that the cause of the loss is not anameliorating factor.Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety systemannunciators or indicators are lost, there is an increased risk that a degraded plant condition couldgo undetected. It is not intended that plant personnel perform a detailed count of theinstrumentation lost but use the value as a judgment threshold for determining the severity of theplant conditions. It is also not intended that the Shift Manager be tasked with making a judgmentdecision as to whether additional personnel are required to provide increased monitoring of systemoperation.It is further recognized that most plant designs provide redundant safety system indication poweredfrom separate uninterruptible power supplies. While failure of a large portion of annunciators ismore likely than a failure of a large portion of indications, the concern is included in this EAL due todifficulty associated with assessment of plant conditions. The loss of specific, or several, safetysystem indicators should remain a function of that specific system or component operability status.This will be addressed by the specific Technical Specification. The initiation of a TechnicalSpecification imposed plant shutdown related to the instrument loss will be reported via 10 CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the NOUE isbased on EAL SU4.1A Site Area Emergency is considered to exist if the control room staff cannot monitor safetyfunctions needed for protection of the public while a significant transient is in progress.Annunciators for this EAL are limited to include those identified in the Abnormal OperatingProcedures, in the Emergency Operating Procedures, and in other EALs (.g., area, process, and/oreffluent rad monitors, etc.)Indications needed to monitor safety functions necessary for protection of the public include controlroom indications, computer generated indications and dedicated annunciation capability."Compensatory indications" in this context includes computer based information such as PlantProcess Computer and SPDS.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.NMP1 Basis Reference(s):1. N1-OP-42 Process Computer/SPDS2. NEI 99-01 IC SS6Page 178 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:S -System Malfunction5 -InstrumentationInitiating Condition: UNPLANNED loss of safety system annunciation or indication inthe Control Room with either (1) a significant transient in progress,or (2) compensatory indicators are unavailableEAL:SA5.1 AlertUNPLANNED loss of > approximately 75% of annunciation or indication on Control Roompanels L, K, H, F and G for _ 15 min. (Note 4)AND EITHER:A significant transient is in progress, Table S-2ORCompensatory indications are unavailable (Plant Process Computer, SPDS)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeTable S-2 Significant Transients" Turbine runback > 25% thermal reactor power* Electric load rejection > 25% full electrical load* Reactor scram* ECCS injection* Thermal power oscillations > 10%Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificPlant Process Computer and SPDS are considered compensatory indication (ref. 1).Significant transients are listed in Table S-2.GenericThis EAL is intended to recognize the difficulty associated with monitoring changing plantconditions without the use of a major portion of the annunciation or indication equipment during asignificant transient.Page 179EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety systemannunciators or indicators are lost, there is an increased risk that a degraded plant condition couldgo undetected. It is not intended that plant personnel perform a detailed count of theinstrumentation lost but use the value as a judgment threshold for determining the severity of theplant conditions. It is also not intended that the Shift Manager be tasked with making a judgmentdecision as to whether additional personnel are required to provide increased monitoring of systemoperation.It is further recognized that most plant designs provide redundant safety system indication poweredfrom separate uninterruptible power supplies. While failure of a large portion of annunciators ismore likely than a failure of a large portion of indications, the concern is included in this EAL due todifficulty associated with assessment of plant conditions. The loss of specific, or several, safetysystem indicators should remain a function of that specific system or component operability status.This will be addressed by the specific Technical Specification. The initiation of a TechnicalSpecification imposed plant shutdown related to the instrument loss will be reported via 10 CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the UE is basedon EAL SU4.1.Annunciators or indicators for this EAL include those identified in the Abnormal OperatingProcedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, process,and/or effluent rad monitors, etc.)."Compensatory indications" in this context includes computer based information such as PlantProcess Computer and SPDS.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor thetransient in progress due to a concurrent loss of compensatory indications with a significanttransient in progress during the loss of annunciation or indication.NMP1 Basis Reference(s):1. N1-OP-42 Process Computer/SPDS2. NEI 99-01 IC SA4Page 180 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 5 -InstrumentationInitiating Condition: UNPLANNED loss of safety system annunciation or indication inthe Control Room for >_ 15 min.EAL:SU5.1 Unusual EventUNPLANNED loss of > approximately 75% of annunciation or indication on Control Roompanels L, K, H, F and G for _ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeMode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificNoneGenericThis EAL is intended to recognize the difficulty associated with monitoring changing plantconditions without the use of a major portion of the annunciation or indication equipment.Recognition of the availability of computer based indication equipment is considered."Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety systemannunciators or indicators are lost, there is an increased risk that a degraded plant condition couldgo undetected. It is not intended that plant personnel perform a detailed count of theinstrumentation lost but use the value as a judgment threshold for determining the severity of theplant conditions.It is further recognized that plant design provides redundant safety system indication powered fromseparate uninterruptible power supplies. While failure of a large portion of annunciators is morelikely than a failure of a large portion of indications, the concern is included in this EAL due todifficulty associated with assessment of plant conditions. The loss of specific, or several, safetysystem indicators should remain a function of that specific system or component operability status.This will be addressed by the specific Technical Specification. The initiation of a TechnicalSpecification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the UE is basedon EAL SU4.1.Page 181 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisAnnunciators or indicators for this EAL include those identified in the Abnormal OperatingProcedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, process,and/or effluent rad monitors, etc.).Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.This UE will be escalated to an Alert based on a concurrent loss of compensatory indications or if asignificant transient is in progress during the loss of annunciation or indication.NMP1 Basis Reference(s):1. N1-OP-42 Process Computer/SPDS2. NEI 99-01 IC SU3Page 182EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory:Subcategory:S -System Malfunction6 -CommunicationsInitiating Condition: Loss of all onsite or offsite communications capabilitiesEAL:SU6.1 Unusual EventLoss of all Table S-3 onsite (internal) communication methods affecting the ability toperform routine operationsORLoss of all Table S-3 offsite (external) communication methods affecting the ability toperform offsite notificationsTable S-3 Communications SystemsSystem Onsite Offsite(internal) (external)PBX (normal dial telephones)GaitronicsHand-Held Portable Radio (station radio)Control Room installed satellite phones (non portable)ENSRECSXXXXXXXMode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificOnsite/offsite communications systems are listed in Table S-3 (ref. 1, 2).This EAL is the hot condition equivalent of the cold condition EAL CU6.1.GenericThe purpose of this EAL is to recognize a loss of communications capability that either defeats theplant operations staff ability to perform routine tasks necessary for plant operations or the ability tocommunicate issues with off-site authorities.Page 183EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisThe loss of off-site communications ability is expected to be significantly more comprehensive thanthe condition addressed by 10 CFR 50.72.The availability of one method of ordinary off-site communications is sufficient to inform federal,state, and local authorities of plant problems. This EAL is intended to be used only whenextraordinary means (e.g., relaying of information from non-routine radio transmissions, individualsbeing sent to off-site locations, etc.) are being used to make communications possible.NMP1 Basis Reference(s):1. Nine Mile Point Nuclear Station Site Emergency Plan, Section 7.22. USAR 2.4.5 Lighting and Communication3. NEI 99-01 IC SU6Page 184 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 7 -Fuel Clad DegradationInitiating Condition: Fuel clad degradationEAL:SU7.1 Unusual EventReactor coolant activity > 4 pCi/gm 1-131 EquivalentMode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificThis EAL addresses reactor coolant samples exceeding Technical Specification 3.2.4(ref. 1). A reactor coolant sample analysis with specific activity in excess of the TechnicalSpecification limit of 4 pCi/gm 1-131 Equivalent is indicative of a degradation of the fuelclad, and is a precursor of more serious problems. This activity level for which operation isallowed to continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to accommodate short duration Iodine spikesfollowing changes in thermal power.GenericThis EAL is included because it is a precursor of more serious conditions and, as result, isconsidered to be a potential degradation of the level of safety of the plant.Escalation of this EAL to the Alert level is via the EALs in Category F.This threshold addresses coolant samples exceeding coolant technical specifications for transientiodine spiking limits.NMP1 Basis Reference(s):1. Technical Specification 3.2.4 Reactor Coolant System -RCS Specific Activity2. NEI 99-01 IC SU4Page 185 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: S -System MalfunctionSubcategory: 7 -Fuel Clad DegradationInitiating Condition: Fuel clad degradationEAL:SU7.2 Unusual EventOffgas radiation monitor RN-12A or RN-12B > hi-hi alarm for > 15 min.Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificElevated offgas radiation activity represents a potential degradation in the level of safety ofthe plant and a potential precursor of more serious problems. The Technical Specificationallowable limit is an offgas level not to exceed 500,000 pCi/sec (recombiner dischargegross noble gases beta and/or gamma) (ref. 1).The hi-hi alarm setpoint has been conservatively selected because it is operationallysignificant and is readily recognizable by Control Room operating staff (ref. 2). 15 minutesis allotted for operator action to reduce the offgas radiation levels and exclude transientconditions (ref. 3). The high offgas radiation alarm is set using methodology outlined in theODCM (ref. 1).The offgas system automatically isolates (BV 77-03) when both RN-12A and RN-12Balarm.GenericThis EAL is included because it is a precursor of more serious conditions and, as result, isconsidered to be a potential degradation of the level of safety of the plant.Escalation of this EAL to the Alert level is via the EALs in Category F.This threshold addresses radiation monitor readings that provide indication of a degradation of fuelclad integrity.NMPI Basis Reference(s):1. ODCM 3.6.14 & 15Page 186 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis2. Ni-ARP-H!, Annunciator Hi 73. NI-SOP-25.2 Fuel Failure or High Activity in RX Coolant or Off Gas4. NEI 99-01 IC SU4Page 187EPMP-EPP-0101Rev 00 Draft A Attachment I -Emergency Action Level Technical BasisCategory:Subcategory:S -System Malfunction8 -RCS LeakageInitiating Condition: RCS leakageEAL:SU8.1 Unusual EventUnidentified drywell leakage > 10 gpmORIdentified reactor coolant drywell leakage > 25 gpmMode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificElevated RCS leakage may be detected by the following annunciators (ref. 1, 2):* H2 1 DRYWELL FLOOR DRAIN LEVEL-HIGH* H2 7 DRYWELL WATER LEAK DETECTION SYSOnce elevated drywell leakage is detected operators will monitor and record Drywell Floorand Equipment Drain Tank leakage to determine drywell unidentified and identifiedleakage values (ref. 3, 4).GenericThis EAL is included as a UE because it may be a precursor of more serious conditions and, asresult, is considered to be a potential degradation of the level of safety of the plant. The 10 gpmvalue for the unidentified or pressure boundary leakage was selected as it is observable withnormal control room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances).Relief valve normal operation should be excluded from this EAL. However, a relief valve thatoperates and fails to close per design should be considered applicable to this EAL if the relief valvecannot be isolated.The EAL for identified leakage is set at a higher value due to the lesser significance of identifiedleakage in comparison to unidentified or pressure boundary leakage. In either case, escalation ofthis EAL to the Alert level is via EALs in Category F.NMP1 Basis Reference(s):Page 188EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis1.2.3.45.N1-ARP-H2 Annunciator H2 1 DRYWELL FLOOR DRAIN LEVEL-HIGHN1-ARP-H2 Annunciator H2 7 DRYWELL WATER LEAK DETECTION SYSS-ODP-OPS-0110 Containment Leakage EvaluationN1-OP-08, Primary Containment Area Cooling SystemNEI 99-01 IC SU5Page 189EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory F -Fission Product Barrier DegradationEAL Group: Hot Conditions (RCS temperature > 2120F);EALs in this category are applicable only inone or more hot operating modes.EALs in this category represent threats to the defense in depth design concept thatprecludes the release of highly radioactive fission products to the environment. Thisconcept relies on multiple physical barriers any one of which, if maintained INTACT,precludes the release of significant amounts of radioactive fission products to theenvironment. The primary fission product barriers are:A. Fuel Clad (FC): Zirconium tubes which house the ceramic uranium oxide pelletsalong with the end plugs which are welded into each end of the fuel rods comprisethe FC barrier.B. Reactor Coolant System (RCS): The reactor vessel shell, vessel head, CRDhousings, vessel nozzles and penetrations, and all primary systems directlyconnected to the RPV up to the outermost Primary Containment isolation valvecomprise the RCS barrier.C. Containment (PC): The drywell, the suppression chamber/pool, their respectiveinterconnecting paths, and other connections up to and including the outermostcontainment isolation valves comprise the Primary Containment barrier.The EALs in this category require evaluation of the loss and potential loss thresholds listedin the fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "PotentialLoss" signify the relative damage and threat of damage to the barrier. "Loss" means thebarrier no longer assures containment of radioactive materials. "Potential Loss" meansintegrity of the barrier is threatened and could be lost if conditions continue to degrade.The number of barriers that are lost or potentially lost and the following criteria determinethe appropriate emergency classification level:Unusual Event:Any loss or any potential loss of ContainmentAlert:Any loss or any potential loss of either Fuel Clad or RCSSite Area Emergency:Loss or potential loss of any two barriersGeneral Emergency:Loss of any two barriers and loss or potential loss of the third barrierPage 190 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisThe logic used for Category F EALs reflects the following considerations:" The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than theContainment Barrier. UE EALs associated with RCS and Fuel Clad Barriers areaddressed under Category S." At the Site Area Emergency level, there must be some ability to dynamically assesshow far present conditions are from the threshold for a General Emergency. Forexample, if Fuel Clad and RCS Barrier "Loss" thresholds existed, that, in addition tooff-site dose assessments, would require continual assessments of radioactiveinventory and containment integrity. Alternatively, if both Fuel Clad and RCS Barrier"Potential Loss" thresholds existed, the ED would have more assurance that therewas no immediate need to escalate to a General Emergency.* The ability to escalate to higher emergency classification levels as an eventdeteriorates must be maintained. For example, RCS leakage steadily increasingwould represent an increasing risk to public health and safety." The Containment Barrier should not be declared lost or potentially lost based onexceeding Technical Specification action statement criteria, unless there 'is an eventin progress requiring mitigation by the Containment barrier.Page 191EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: Loss of ANY two barriers and loss or potential loss of the thirdbarrierEAL:FGI.1 General EmergencyLoss of ANY two fission product barriersANDLoss or potential loss of third fission product barrier (Table F-i)Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificFuel Clad, RCS and Containment comprise the fission product barriers. Table F-1(Attachment 2) lists the fission product barrier thresholds, bases and references.At the General Emergency classification level each barrier is weighted equally. A GeneralEmergency is therefore appropriate for any combination of the following conditions:" Loss of Fuel Clad, RCS and Containment barriers* Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier" Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier* Loss of Fuel Clad and Containment barriers with potential loss of RCS barrierGenericNoneNMP1 Basis Reference(s):1. NEI 99-01 IC FG1Page 192 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: Loss or potential loss of ANY two barriersEAL:FSI.1 Site Area EmergencyLoss or potential loss of ANY two fission product barriers (Table F-I)Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificFuel Clad, RCS and Containment comprise the fission product barriers. Table F-1(Attachment 2) lists the fission product barrier thresholds, bases and references.At the Site Area Emergency classification level, each barrier is weighted equally. A SiteArea Emergency is therefore appropriate for any combination of the following conditions:" One barrier loss and a second barrier loss (i.e., loss -loss)* One barrier loss and a second barrier potential loss (i.e., loss -potential loss)* One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss)At the Site Area Emergency classification level, the ability to dynamically assess theproximity of present conditions with respect to the threshold for a General Emergency isimportant. For example, the existence of Fuel Clad and RCS Barrier loss thresholds inaddition to offsite dose assessments would require continual assessments of radioactiveinventory and Containment integrity in anticipation of reaching a General Emergencyclassification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed,the Emergency Director would have greater assurance that escalation to a GeneralEmergency is less IMMINENT.GenericNoneNMP1 Basis Reference(s):Page 193 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical Basis1. NEI 99-01 IC FS1Page 194 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: ANY loss or ANY potential loss of EITHER Fuel Clad OR RCSEAL:FA1.1 AlertANY loss or ANY potential loss of EITHER Fuel Clad barrier OR RCS barrier (Table F-1)Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificFuel Clad, RCS and Containment comprise the fission product barriers. Table F-1(Attachment 2) lists the fission product barrier thresholds, bases and references.At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavilythan the Containment barrier. Unlike the Containment barrier, loss or potential loss ofeither the Fuel Clad or RCS barrier may result in the relocation of radioactive materials ordegradation of core cooling capability. Note that the loss or potential loss of Containmentbarrier in combination with loss or potential loss of either Fuel Clad or RCS barrier resultsin declaration of a Site Area Emergency under EAL FS1.GenericNoneNMP1 Basis Reference(s):1. NEI 99-01 IC FA1Page 195 EPMP-EPP-0101Rev 00 Draft A -Emergency Action Level Technical BasisCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: ANY loss or ANY potential loss of ContainmentEAL:FUI.1 Unusual EventANY loss or ANY potential loss of Containment barrier (Table F-I)Mode Applicability:1 -Power Operation, 2 -Hot ShutdownBasis:Plant-SpecificFuel Clad, RCS and Containment comprise the fission product barriers. Table F-1(Attachment 2) lists the fission product barrier thresholds, bases and references.Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier.Unlike the Fuel Clad and RCS barriers, the loss of either of which results in, an Alert (EALFA1.1), loss of the Containment barrier in and of itself does not result in the relocation ofradioactive materials or the potential for degradation of core cooling capability. However,loss or potential loss of the Containment barrier in combination with the loss or potentialloss of either the Fuel Clad or RCS barrier results in declaration of a Site Area Emergencyunder EAL FS1.1.GenericNoneNMP1 Basis Reference(s):1. NEI 99-01 IC FUlPage 196 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisIntroductionTable F-1 lists the threshold conditions that define the Loss and Potential Loss of the threefission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The tableis structured so that each of the three barriers occupies adjacent columns. Each fissionproduct barrier column is further divided into two columns; one for Loss thresholds and onefor Potential Loss thresholds.The first column of the table (to the left of the Fuel Clad Loss column) lists the categories(types) of fission product barrier thresholds. The fission product barrier categories are:A. RPV water levelB. Primary Containment Pressure / TemperatureC. IsolationD. RadE. JudgmentEach category occupies a row in Table F-1 thus forming a matrix defined by the categoryrows and the Loss/Potential Loss columns. The intersection of each category row witheach Loss/Potential Loss column forms a cell in which one or more fission product barrierthresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/PotentialLoss, the word "None" is entered in the cell.Thresholds are assigned sequential numbers within each Loss and Potential Loss columnbeginning with number one. In this manner, a threshold can be identified by its categorytitle and number. For example, the first Fuel Clad barrier Loss in Category A would beassigned "FC Loss A.1 ," the third Containment barrier Potential Loss would be assigned"PC P-Loss B.3," etc.If a cell in Table F-1 contains more than one numbered threshold, each of the numberedthresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessaryto exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.Subdivision of Table F-1 by category facilitates association of plant conditions to theapplicable fission product barrier Loss and Potential Loss thresholds. This structurePage 197 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and Basispromotes a systematic approach to assessing the classification status of the fissionproduct barriers.When equipped with knowledge of plant conditions related to the fission product barriers,the EAL-user first scans down the category column of Table F-i, locates the likelycategory and then reads across the row of fission product barrier Loss and Potential Lossthresholds in that category to determine if any threshold has been exceeded. If a thresholdhas not been exceeded in that category row, the EAL-user proceeds to the next likelycategory and continues review of the row of thresholds in the new categoryIf the EAL-user determines that any threshold has been exceeded, by definition, the barrieris lost or potentially lost -even if multiple thresholds in the same barrier column areexceeded; only that one barrier is lost or potentially lost. The EAL-user must examine eachof the three fission product barriers to determine if other barrier thresholds in the categoryare lost or potentially lost. For example, if Primary Containment radiation is sufficientlyhigh, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Containmentbarrier can occur. Barrier Losses and Potential Losses are then applied to the algorithmsgiven in EALs FG1.1, FS1.1, FAI.1 and FUI.1 to determine the appropriate emergencyclassification.In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first,followed by the RCS barrier and finally the Containment barrier threshold bases. In eachbarrier, the bases are given according category Loss followed by category Potential Lossbeginning with Category A, then B.. .E.Page 198 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisTable F-I Fission Product Barrier MatrixFuel Clad Barrier Reactor Coolant System Barrier Containment BarrierCategory Loss Potential Loss Loss Potential Loss Loss Potential Loss1 RPV water level cannot berestored and maintained 1. RPV water level cannot beRPV 1. Primary Containment Flooding is above -84 in. following restored and maintained None None 1, Primary Containment Flooding isWPe required depressurization of the RPV or above -84 in. or RPV water requiredWater RPV water level cannot be level cannot be determinedLevel determined2. Torus pressure > 35 psig and1. Primary Containment pressure risingrise followed by a rapid 3. Explosive mixture exists insidePrimary UNPLANNED drop in Primary Primary Containment (> 6% H2 andContainm None None 2. Primary Containment pressure > Containment pressure n5% 02)enm 3.5 psig due to RCS leakage 2. Primary Containment pressure 4. Torus water temperature and RPVesu response not consistent with pressure cannot be maintainedPressure LOCA conditions below the Heat CapacityTemp. Temperature Limit (N1-EOP-4Figure M)3. Failure of all PrimaryContainment isolation valves inANY one line to close followingauto or manual initiation3. Release pathway exists outside 1 UNISOLABLE primary system ANDPrimary Containment resulting leakage outside Primary Direct downstream pathwayfrom isolation failure in ANY of Containment as indicated by outside Primary Containment andthe following (excluding normal exceeding EITHER: to the environment existsprocess system flowpaths froman UNISOLABLE system): Area temperature above ANY 4. Intentional Primary ContainmentNone None N1 -EOP-5 Detail T alarm venting per EOPs NoneIsolation setpoint 5. UNISOLABLE primary system* EC steam line OR leakage outside Primary* RWCU Area radiation above ANY Containment as indicated by* Feedwater N1-EOP-5 Detail R alarm exceeding EITHER:4. RPV blowdown is required setpoint Maximum safe general areatemperature of 1351FORMaximum safe area radiationof 8 Rfhr2. Drywavll radiation R/hrD None 5. Drywell radiation 80 R/hr None None 5. Drywell radiation > 4.0E4 R/hrRad 3. Reactor coolant activity> 300 pCi/gm 1-131 Equivalent4. ANY condition in the opinion of 2. ANY condition in the opinion of 6. ANY condition in the opinion of 2 ANY condition in the opinion of the 6. ANY condition in the opinion of 6. ANY condition in the opinion of theE the Emergency Director that the Emergency Director that the Emergency Director that Emergency Director that indicates the Emergency Director that Emergency Director that indicatesindicates loss of the Fuel Clad indicates potential loss of the indicates loss of the Reactor potential loss of the Reactor indicates loss of the Containment potential loss of the ContainmentJudgment barrier Fuel Clad barrer Coolant System barrier Coolant System barrier barrer barderPage 199EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: A. RPV Water LevelDegradation Threat: LossThreshold:1. Primary Containment Flooding is requiredBasis:Plant-SpecificRequirements for Primary Containment Flooding are established in EOP-2 Step L-18; EOP-3 Steps L-8, L-10 and L-13; and EOP-7 Overrides 3 and 18. These EOPs provideinstructions to ensure adequate core cooling by maintaining RPV water level aboveprescribed limits or operating sufficient RPV injection sources when level cannot bedetermined. SAP entry is required when (ref. 1, 2. 3, 4):RPV water level cannot be restored and maintained above -109 in. with insufficientCore Spray flow: The Minimum Steam Cooling RPV Water Level (MSCRWL) is thelowest RPV water level at which the covered portion of the reactor core willgenerate sufficient steam to preclude any clad temperature in the uncovered portionof the core from exceeding 1500'F. Core spray flow is insufficient if you cannotrestore and maintain both Core Spray loop flows .at or above 180 x 104 Ibm/hr.Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that the parameter cannot be restored and maintained above the limitmay be made at, before, or after the parameter actually decreases to this point.* RPV water level cannot be determined and it is determined that core damage isoccurring: When RPV water level cannot be determined, EOPs require RPVflooding strategies. RPV water level indication provides the primary means ofknowing if adequate core cooling is being maintained. When all means ofdetermining RPV water level are unavailable, reliance on alternate means ofassuring adequate core cooling must be attempted. The instructions in EOP-7specify these means, which include blowdown of the RPV and injection into thePage 200 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisRPV at a rate needed to flood to the elevation of the main steam lines or hold RPVpressure above the Minimum Steam Cooling Pressure (in ATWS events)This threshold is also a Potential Loss of the Containment barrier (PC P-Loss A.1). SinceSAP entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RCSLoss A.1). Primary Containment Flooding (SAP entry), therefore, represents a Loss of twobarriers and a Potential Loss of a third, which requires a General Emergency classification.GenericThis site specific value corresponds to the level used in EOPs to indicate challenge of core cooling.This is the minimum value to assure core cooling without further degradation of the clad.NMP1 Basis Reference(s):1. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document2. N1-EOP-2 RPV Control3. N1-EOP-3 Failure to Scram4. N1-EOP-7 RPV Flooding5. NEI 99-01 FC Loss 2Page 201EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: LossThreshold:NonePage 202EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: C. IsolationDegradation Threat: LossThreshold:NonePage 203EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: D. RadDegradation Threat: LossThreshold:2. Drywell radiation _: 3,000 R/hrBasis:Plant-SpecificIt is important to recognize that the radiation monitor may be sensitive to shine from theRPV or RCS piping (caused by lower than normal RPV water level for example). TheDrywell High Range Radiation Monitors are the following (ref. 1):" RAM 201.7-36 Located: Az 3400, El 263' 6"" RAM 201.7-37 Located: Az 3100, El 301' 0"The Drywell High Range Radiation Monitors have a range of 1 EO to 1 E8 R/hr on recorderRR 201.7-36C pens 1 and 2 (ref. 1).The threshold value (3,090 R/hr rounded to 3,000 R/hr) was calculated assuming theinstantaneous release and dispersal of the reactor coolant noble gas and iodine inventoryassociated with a concentration of 300 pCi/gm 1-131 Equivalent (or approximately 5% cladfailure) into the drywell atmosphere (ref. 2).GenericThe 3,000 R/hr reading is a value which indicates the release of reactor coolant, with elevatedactivity indicative of fuel damage, into the drywell.Reactor coolant concentrations of this magnitude are several times larger than the maximumconcentrations (including iodine spiking) allowed within technical specifications and are thereforeindicative of fuel damage.This value is higher than that specified for RCS barrier Loss threshold D.5. Thus, this thresholdindicates a loss of both Fuel Clad barrier and RCS barrier that appropriately escalates theemergency classification level to a Site Area Emergency.There is no Potential Loss threshold associated with this item.Page 204 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMPI Basis Reference(s):1. N1-RSP-1OC The Use and Routine Calibration of the General Atomic High RangeGamma Radiation Monitoring System2. Calculation 1H21C003, Rev. 03. NEI 99-01 FC Loss 4Page 205EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:Category:Fuel CladD. RadDegradation Threat: LossThreshold:L 3. Reactor coolant activity > 300 pCi/gm 1-131 EquivalentBasis:Plant-SpecificNoneGenericThe site specific value corresponds to 300 pCi/gm 1-131 Equivalent. Assessment by the EAL TaskForce indicates that 300 piCi/gm 1-131 Equivalent coolant activity is well above that expected foriodine spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivityindicates significant clad damage and thus the Fuel Clad Barrier is considered lost.There is no Potential Loss threshold associated with this item.NMP1 Basis Reference(s):1. NEI 99-01 FC Loss 1Page 206EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: E. JudgmentDegradation Threat: LossThreshold:4. ANY condition in the opinion of the Emergency Director that indicates loss of the FuelClad barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the Fuel Clad barrier is lost. Such a determination should include IMMINENTbarrier degradation, barrier monitoring capability and dominant accident sequences.* IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to recognition of the inability to reach safety acceptance criteriabefore completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericThese This threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the Fuel Clad barrier is lost. In addition, the inability to monitor the barriershould also be incorporated in this threshold as a factor in Emergency Director judgment that thebarrier may be considered lost.Page 207 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP1 Basis Reference(s):1. NEI 99-01 FC Loss 6Page 208EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: A. RPV Water LevelDegradation Threat: Potential LossThreshold:1. RPV water level cannot be restored and maintained above -84 in. followingdepressurization of the RPV or RPV water level cannot be determinedBasis:Plant-SpecificAn RPV water level instrument reading of -84 in. (rounded from 84 7/16") indicates RPVwater level is at the top of active fuel. When RPV water level is at or above the top ofactive fuel, the core is completely submerged. Core submergence is the most desirablemeans of core cooling. When RPV water level is below the top of active fuel followingdepressurization of the RPV (automatically, manually or by failure of the RCS barrier), theuncovered portion of the core must be cooled by less reliable means (i.e., spray cooling).lIfcore uncovery is threatened, the EOPs specify alternate, more extreme, RPV water levelcontrol measures in order to restore and maintain adequate core cooling (ref. 1).Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that RPV water level cannot be restored and maintained above the top ofactive fuel may be made at, before, or after RPV water level actually decreases to thispoint. (ref. 1)When RPV water level cannot be determined, EOPs require RPV flooding strategies. RPVwater level indication provides the primary means of knowing if adequate core cooling isbeing maintained. When all means of determining RPV water level are unavailable, thefuel clad barrier is threatened and reliance on alternate means of assuring adequate corecooling must be attempted. The instructions in EOP-7 specify these means, which includeemergency depressurization of the RPV and injection into the RPV at a rate needed toflood to the elevation of the main steam lines or hold RPV pressure above the MinimumSteam Cooling Pressure (in ATWS events). (ref. 2) If RPV water level cannot bePage 209 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and Basisdetermined with respect to the top of active fuel, a potential loss of the fuel clad barrierexists.Note that EOP-3 may require intentional uncovery of the core and control of RPV waterlevel between -84 in. and -109 in., the Minimum Steam Cooling RPV Water Level(MSCRWL) (ref. 3). Under these conditions, a high-power ATWS event exists and requiresat least a Site Area Emergency classification in accordance with the ATWS/CriticalityEALs.GenericThe site specific RPV water level threshold is the same as the RCS barrier Loss threshold A.1 andcorresponds to the RPV water level at the top of the active fuel. Thus, this threshold indicates aPotential Loss of the Fuel Clad barrier and a Loss of RCS barrier that appropriately escalates theemergency classification level to a Site Area Emergency. This threshold is considered to beexceeded when, as specified in the site specific EOPs, that RPV water cannot be restored andmaintained above the specified level following depressurization of the RPV (either manually,automatically or by failure of the RCS barrier).NMP1 Basis Reference(s):1. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document2. N1-EOP-7 RPV Flooding3. N1-EOP-3 Failure to Scram4. NEI 99-01 FC Potential Loss 2Page 210EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: Potential LossThreshold:NonePage 211EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: C. IsolationDegradation Threat: Potential LossThreshold:NonePage 212EPMP-EPP-01 01Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:Category:Degradation Threat:Threshold:Fuel CladD. RadPotential LossNonePage 213EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: E. JudgmentDegradation Threat: Potential LossThreshold:2. ANY condition in the opinion of the Emergency Director that indicates potential loss ofthe Fuel Clad barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the Fuel Clad barrier is potentially lost. Such a determination should includeIMMINENT barrier degradation, barrier monitoring capability and dominant accidentsequences." IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to recognition of the inability to reach safety acceptance criteriabefore completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the Fuel Clad barrier is potentially lost. In addition, the inability to monitor thebarrier should also be incorporated in this threshold as a factor in Emergency Director judgmentthat the barrier may be considered potentially lost.Page 214 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP1 Basis Reference(s):1. NEI 99-01 FC Potential Loss 6Page 215EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: A. RPV Water LevelDegradation Threat: LossThreshold:1. RPV water level cannot be restored and maintained above -84 in. or RPV water levelcannot be determinedBasis:Plant-SpecificAn RPV water level instrument reading of -84 in. (rounded from 84 7/16") indicates RPVwater level is at the top of active fuel (ref. 1). The top of the active fuel is significantly lowerthan the normal operating RPV water level control band. To reach this level, RPV inventoryloss would have previously required isolation of the RCS and Containment (PC) barriers,and initiation of all ECCS. If RPV water level cannot be maintained above the top of activefuel, ECCS and other sources of RPV injection have been ineffective or incapable ofreversing the decreasing level trend. The cause of the loss of RPV inventory is thereforeassumed to be a Loss of Coolant Accident (LOCA). By definition, a LOCA event is a Lossof the RCS barrier.Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that RPV water level cannot be restored and maintained above the top ofactive fuel may be made at, before, or after RPV water level actually decreases to thispoint. (ref. 1)When RPV water level cannot be determined, EOPs require RPV flooding strategies. TheRPV flooding instructions in EOP-7 first specify blowdown of the RPV (ref. 2), which isdefined to be a Loss of the RCS barrier (RCS Loss C.4).Note that EOP-3 may require intentional uncovery of the core and control of RPV waterlevel between -84 in. and -109 in., the Minimum Steam Cooling RPV Water Level(MSCRWL) (ref. 3). Under these conditions, a high-power ATWS event exists and requiresat least a Site Area Emergency classification in accordance with the ATWS/CriticalityEALs.Page 216 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisGenericThe Loss threshold RPV water level of -84 in. corresponds to the level that is used in EOPs toindicate challenge of core cooling.The threshold value is the same as Fuel Clad Barrier Potential Loss threshold A.1 and correspondsto a challenge to core cooling. Thus, this threshold indicates a Loss of RCS barrier and PotentialLoss of Fuel Clad barrier that appropriately escalates the emergency classification level to a SiteArea Emergency.Unlike the Fuel Clad barrier RPV water level Potential Loss threshold (top of the active fuel), theadditional requirement that the RPV be depressurized is not associated with the RCS barrierPotential Loss. The significant loss of inventory that must occur to determine that RPV water levelcannot be restored and maintained above the threshold is, by itself, a very strong indication thatthe RCS barrier is no longer capable of retaining sufficient inventory to keep the core submerged,and thus represents a Loss of the RCS Barrier.There is no Potential Loss threshold associated with this item.NMP1 Basis Reference(s):1. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document2. N1-EOP-7 RPV Flooding3. N1-EOP-3 Failure to Scram4. NEI 99-01 RCS Loss 2Page 217EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: LossThreshold:2. Primary Containment pressure > 3.5 psig due to RCS leakageBasis:Plant-SpecificThe drywell high pressure scram setpoint is an entry condition to the EOP flowcharts:EOP-2, RPV Control, and EOP-4, Primary Containment Control (ref. 1, 2). Normal PrimaryCQntainment (PC) pressure control functions such as operation of drywell cooling andventing through RBEVS are specified in EOP-4 in advance of less desirable but moreeffective functions such as operation of drywell or suppression chamber sprays.In the NMP1 design basis, Primary Containment pressures above the drywell highpressure scram setpoint are assumed to be the result of a high-energy release into thecontainment for which normal pressure control systems are inadequate or incapable ofreversing the increasing pressure trend. Pressures of this magnitude, however, can becaused by non-LOCA events such as a loss of drywell cooling or inability to controlPrimary Containment vent/purge (ref. 3).The threshold phrase "...due to RCS leakage" focuses the barrier failure on the RCSinstead of the non-LOCA malfunctions that may adversely affect Primary Containmentpressure. Primary Containment pressure greater than 3.5 psig with corollary indications(e.g., elevated drywell temperature, indications of loss of RCS inventory) should, therefore,be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressuregreater than 3.5 psig should not be considered an RCS barrier loss.GenericThe Primary Containment pressure of 3.5 psig is based on the drywell high pressure set pointwhich indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.There is no Potential Loss threshold associated with this item.Page 218 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMPI Basis Reference(s):1. N1-EOP-2 RPV RPV Control2. N1-EOP-4 Primary Containment Control3. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document4. NEI 99-01 RCS Loss 1Page 219EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: C. IsolationDegradation Threat: LossThreshold:3. Release pathway exists outside Primary Containment resulting from isolation failure inANY of the following systems (excluding normal process system flowpaths from anUNISOLABLE system):" Main steam line" EC steam line* RWCU* FeedwaterBasis:Plant-SpecificThe conditions of this threshold include required containment isolation failures allowing aflow path to the environment. A release pathway outside Primary Containment exists whenflow is not prevented by downstream isolations. Emergency declaration under thisthreshold would not be required in the case of a failure of both isolation valves to close butno downstream flowpath exists. Similarly, if the emergency response requires the normalprocess flow of a system outside Primary Containment (e.g., EOP requirement to bypassMSIV low RPV water level interlocks and maintain the main condenser as a heat sinkusing main turbine bypass valves), the threshold is not met. The combination of thesethreshold conditions represent the loss of both the RCS and Containment (see PC LossC.3) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or PotentialLoss of any two barriers). (ref. 1-3)Even though RWCU and Feedwater systems do not contain steam, they are included inthe list because an UNISOLABLE break could result in the high-pressure discharge of fluidthat is flashed to steam from relatively large volume systems directly connected to theRCS.GenericPage 220 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisAn UNISOLABLE MSL break is a breach of the RCS barrier. Thus, this threshold is included forconsistency with the Alert emergency classification level.Other large high-energy line breaks such as EC steam line, Feedwater or RWCU that areUNISOLABLE also represent a significant loss of the RCS barrier and should be considered asMSL breaks for purposes of classification.NMPI Basis Reference(s):1. USAR Section VIII.A Protective Systems2. USAR Section V.E Emergency Cooling System3. USAR Section X.B Reactor Cleanup System4. NEI 99-01 RCS Loss 3APage 221EPMP-EPP-0101Rev 00 Draft A -Fission Product 1affier Loss I P6tehtial Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: C. IsolationDegradation Threat: LossThreshold:4. RPV blowdown is requiredBasis:Plant-SpecificRPV blowdown (Emergency RPV Depressurization) is specified in the EOP flowcharts(EOP-8 RPV Blowdown) when symbols containing the phrase "BLOW DOWN" arereached. The requirements for emergency RPV depressurization appear in the followingEOPs (ref. 1-7):" EOP-2 RPV Control* EOP-3 Failure to Scram* EOP-4 Primary Containment Control-" EOP-4.2 Hydrogen Control* EOP-5 Secondary Containment Control0 EOP-6 Radioactivity Release Control" EOP-9 Steam CoolingRPV blowdown (Emergency RPV depressurization) is also performed upon entry to EOP-7(ref. 8).GenericPlant symptoms requiring Emergency RPV Depressurization (RPV blowdown) per the EOPflowcharts are indicative of a loss of the RCS barrier. If Emergency RPV depressurization isrequired, the plant operators are directed to open electromatic relief valves (ERVs) and keep themopen. Even though the RCS is being vented into the suppression pool, a loss of the RCS should beconsidered to exist due to the diminished effectiveness of the RCS pressure barrier to a release offission products beyond its boundary.Page 222 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMPI Basis Reference(s):1. N1-EOP-2 RPV Control2. N1-EOP-3 Failure to Scram3. N1-EOP-4 Primary Containment Control4. N1 -EOP-4.2 Hydrogen Control5. N1-EOP-5 Secondary Containment Control6. N1-EOP-6 Radioactivity Release Control7. N1-EOP-9 Steam Cooling8. N1-EOP-7 RPV Flooding9. NEI 99-01 RCS Loss 3BPage 223EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: D. RadDegradation Threat: LossThreshold:5. Drywell radiation _ 80 R/hrBasis:Plant-SpecificIt is important to recognize that the radiation monitor may be sensitive to shine from theRPV or RCS piping (caused by lower than normal RPV water level for example). TheDrywell High Range Radiation Monitors are the following (ref. 1):" RAM 201.7-36 Located: Az 340°, El 263' 6"* RAM 201.7-37 Located: Az 3100, El 301' 0"The Drywell High Range Radiation Monitors have a range of 1 EO to 1 E8 R/hr on recorderRR 201.7-36C pens 1 and 2 (ref. 1).The threshold value was calculated assuming the instantaneous release and dispersal ofthe reactor coolant noble gas and iodine inventory associated with normal operatingconcentrations (i.e., Technical Specification coolant activity limit of 4 pCi/gm 1-131Equivalent) into the drywell atmosphere (ref. 2). The reading is less than that specified forthe Fuel Clad Loss because no damage to the fuel clad is assumed. Only leakage fromthe RCS is assumed in this RCS Loss. The referenced calculation resulted in a thresholdvalue of 88.5 R/hr. A value of 80 R/hr is selected because it is observable on existinginstrumentation.GenericThe 80 R/hr reading is a value which indicates the release of reactor coolant to the PrimaryContainment.This reading will be less than that specified for Fuel Clad barrier Loss threshold D.2. Thus, thisthreshold would be indicative of a RCS leak only. If the radiation monitor reading increased to thatvalue specified by Fuel Clad Barrier threshold, fuel damage would also be indicated.Page 224 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisThere is no Potential Loss threshold associated with this item.NMP1 Basis Reference(s):1. Nl-RSP-10C The Use and Routine Calibration of the General Atomic High RangeGamma Radiation Monitoring System2. Calculation 1H21C003, Rev. 0 CCN 008846 Rev. 03. NEI 99-01 RCS Loss 4Page 225EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: E. JudgmentDegradation Threat: LossThreshold:6. ANY condition in the opinion of the Emergency Director that indicates loss of the RCSbarrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the RCS barrier is lost. Such a determination should include IMMINENTbarrier degradation, barrier monitoring capability and dominant accident sequences.* IMMINENT barrier deqradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to the recognition of the inability to reach safety acceptancecriteria before completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the RCS barrier is lost. In addition, the inability to monitor the barrier shouldalso be incorporated in this threshold as a factor in Emergency Director judgment that the barriermay be considered lost.Page 226 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP1 Basis Reference(s):1. NEI 99-01 RCS Loss 6Page 227EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: A. RPV Water LevelDegradation Threat: Potential LossThreshold:NonePage 228EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier. Loss /Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: B. Primary Containment Pressure I TemperatureDegradation Threat: Potential LossThreshold:NonePage 229EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: C. IsolationDegradation Threat: Potential LossThreshold:1. UNISOLABLE primary system leakage outside Primary Containment as indicated byexceeding EITHER:ANY N1-EOP-5 Detail T area temperature alarm setpointORANY N1-EOP-5 Detail R area radiation alarm setpointBasis:Plant-SpecificThe presence of elevated general area temperatures or radiation levels in the ReactorBuilding (RB) may be indicative of UNISOLABLE primary system leakage outside thePrimary Containment. When parameters reach the threshold level, equipment failure ormisoperation may be occurring. Elevated parameters may also adversely affect the abilityto gain access to or operate equipment within the affected area. (ref. 1, 2)In general, multiple indications should be used to determine if a primary system isdischarging outside Primary Containment. For example, a high area radiation conditiondoes not necessarily indicate that a primary system is discharging into the secondarycontainment since this may be caused by radiation shine from nearby steam lines or themovement of radioactive materials. Conversely, a high area radiation condition inconjunction with other indications (e.g. room flooding, high area temperatures, reports ofsteam in the secondary containment, an unexpected rise in feedwater flowrate, orunexpected main turbine control valve closure) may indicate that a primary system isdischarging into the secondary containment.GenericEOP-5 temperature alarm setpoints or area radiation alarm setpoints in the areas of the mainsteam line tunnel, main turbine generator, RCIC, etc., indicate a direct path from the RCS to areasoutside Primary Containment.Page 230 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisThe indicators reaching the threshold barriers and confirmed to be caused by RCS leakagewarrant an Alert classification. An UNISOLABLE leak which is indicated by a high alarm setpointescalates to a Site Area Emergency when combined with Containment Barrier Loss threshold C.5(after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is alsoexceeded.NMP1 Basis Reference(s):1. N1-EOP-5 Secondary Containment Control2. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document3. NEI 99-01 RCS Potential Loss 3BPage 231 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:Category:Degradation Threat:Threshold:Reactor Coolant SystemD. RadPotential LossNonePage 232EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: E. JudgmentDegradation Threat: Potential LossThreshold:2. ANY condition in the opinion of the Emergency Director that indicates potential loss ofthe RCS barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the RCS barrier is potentially lost. Such a determination should includeIMMINENT barrier degradation, barrier monitoring capability and dominant accidentsequences.* IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to the inability to reach final safety acceptance criteria beforecompleting all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the RCS barrier is potentially lost. In addition, the inability to monitor thebarrier should also be incorporated in this threshold as a factor in Emergency Director judgmentthat the barrier may be considered potentially lost.Page 233 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMPI1 Basis Reference(s):1. NEI 99-01 RCS Potential Loss 6Page 234EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: A. RPV Water LevelDegradation Threat: LossThreshold:NonePage 235EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:ContainmentB. Primary Containment Pressure / TemperatureCategory:Degradation Threat: LossThreshold:1. Primary Containment pressure rise followed by a rapid UNPLANNED drop in PrimaryContainment pressureBasis:Plant-SpecificNoneGenericRapid UNPLANNED loss of pressure (i.e., not attributable to drywell spray or condensation effects)following an initial pressure increase from a high energy line break indicates a loss of containmentintegrity. Primary Containment pressure should increase as a result of mass and energy releaseinto containment from a LOCA. Thus, Primary Containment pressure not increasing under theseconditions indicates a loss of containment integrity.This indicator relies on operator recognition of an unexpected response for the condition andtherefore does not have a specific value associated with it. The unexpected response is importantbecause it is the indicator for a containment bypass condition.NMPI Basis Reference(s):1. NEI 99-01 CMT Loss 1APage 236 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: LossThreshold:2. Primary Containment pressure response not consistent with LOCA conditionsBasis:Plant-SpecificUSAR Sections VI.A.2.2 and VI.B.1.2 provide a summary of Primary Containmentpressure response for the design basis loss of coolant accident and the conditionsresulting in the release of RCS inventory to the containment (ref. 1, 2). The maximumcalculated drywell pressure is 34 psig (unless the large rupture was preceded by a smallbreak that prepurged the drywell of nitrogen in which case the peak pressure could reach50 psig). These pressures are well below the design allowable drywell pressure of 62psig. (ref. 1,2)Due to conservatisms in LOCA analyses, actual pressure response is expected to be lessthan the analyzed response. LOCA conditions are manifested on Control Roominstrumentation by drywell pressure rising with torus pressure following and eventuallyequalizing (around 22 psig for the DBA LOCA) (ref. 2).GenericRapid UNPLANNED loss of pressure (i.e., not attributable to drywell spray or condensation effects)following an initial pressure increase from a high energy line break indicates a loss of containmentintegrity. Primary Containment pressure should increase as a result of mass and energy releaseinto containment from a LOCA. Thus, Primary Containment pressure not increasing under theseconditions indicates a loss of containment integrity.This indicator relies on operator recognition of an unexpected response for the condition andtherefore does not have a specific value associated with it. The unexpected response is importantbecause it is the indicator for a containment bypass condition.NMPI Basis Reference(s):1. USAR Section VI.A.2.2 Loss-of-Coolant-Accident2. USAR Section VI.B.1.2 Design Basis Accident (DBA)Page 237 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and Basis3. NEI 99-01 CMT Loss 1BPage 238EPMP-EPP-01 01Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: C. IsolationDegradation Threat: LossThreshold:3. Failure of all Primary Containment isolation valves in ANY one line to close followingauto or manual initiationANDDirect downstream pathway outside Primary Containment and to the environmentexistsBasis:Plant-SpecificThis threshold addresses failure of open isolation devices which should close upon receiptof a manual or automatic Primary Containment isolation signal resulting in a significantradiological release pathway directly to the environment. The concern is the UNISOLABLEopen pathway to the environment. A failure of the ability to isolate any one. line indicates abreach of Primary Containment integrity.As stated above, the adjective "Direct" modifies "pathway" to discriminate against releasepaths through interfacing liquid systems. Leakage into a closed system is to be consideredonly if the closed system is breached and thereby creates a significant pathway to theenvironment. Examples include UNISOLABLE Main steam line or Feedwater line breaks,UNISOLABLE RWCU system breaks, and UNISOLABLE Primary Containmentatmosphere vent paths. If the main condenser is available with an UNISOLABLE mainsteam line, there may be releases through the steam jet air ejectors and gland sealexhausters. These pathways are monitored, however, and do not meet the intent of anonisolable release path to the environment. These minor releases are assessed usingthe Category R EALs.The existence of an in-line charcoal filter (RBEVS) does not make a release path indirectsince the filter is not effective at removing fission noble gases. Typical filters have anefficiency of 95-99% removal of iodine. Given the magnitude of the core inventory ofiodine, significant releases could still occur. In addition, since the fission product releasePage 239 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and Basiswould be driven by boiling in the reactor vessel, the high humidity in the release streamcan be expected to render the filters ineffective in a short period.The threshold is met if the breach is not isolable from the Control Room or an attempt forisolation from the Control Room has been made and was unsuccessful. An attempt forisolation from the Control Room should be made prior to the emergency classification. Ifoperator actions from the Control Room are successful, this threshold is not applicable.Credit is not given for operator actions taken in-plant (outside the Control Room) to isolatethe breach.EOP-4, Primary Containment Control, and EOP-4.2, Hydrogen Control, may specifyPrimary Containment venting and intentional bypassing of the containment isolation valvelogic even if offsite radioactivity release rate limits are exceeded (ref. 1, 2). Under theseconditions with a VALID containment isolation signal, the Containment barrier should beconsidered lost.GenericThese thresholds address incomplete containment isolation that allows direct release to theenvironment.The use of the modifier "direct" in defining the release path discriminates against release pathsthrough interfacing liquid systems. The existence of an in-line charcoal filter does not make arelease path indirect since the filter is not effective at removing fission product noble gases. Typicalfilters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory ofiodine, significant releases could still occur. In addition, since the fission product release would bedriven by boiling in the reactor vessel, the high humidity in the release stream can be expected torender the filters ineffective in a short period.NMPI Basis Reference(s):1. N1-EOP-4 Primary Containment Control2. N1-EOP-4.2 Hydrogen Control3. NEI 99-01 CMT Loss 3APage 240 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: C. IsolationDegradation Threat: LossThreshold:4. Intentional Primary Containment venting per EOPsBasis:Plant-SpecificEOP-4, Primary Containment Control, and EOP-4.2, Hydrogen Control, may specifyPrimary Containment venting and intentional bypassing of the containment isolation valvelogic, even if offsite radioactivity release rate limits are exceeded (ref. 1, 2). The thresholdis met when the operator begins venting the Primary Containment in accordance withEOP-4.1 Primary Containment Venting , not when actions are taken to bypass interlocksprior to opening the vent valves (ref. 3). Purge and vent actions specified in EOP-1Attachment 10 to control Primary Containment pressure below the drywell high pressurescram setpoint by venting through RBEVS do not meet this threshold because such actionis only permitted if offsite radioactivity release rates will remain below the ODCM limits (ref.1,2).GenericThese thresholds address incomplete containment isolation that allows direct release to theenvironment.Site specific EOPs may direct containment isolation valve logic(s) to be intentionally bypassed,regardless of radioactivity release rates. Under these conditions with a VALID containmentisolation signal, the containment should also be considered lost if containment venting is actuallyperformed.Intentional venting of Primary Containment for Primary Containment pressure or combustible gascontrol per EOPs to the secondary containment and/or the environment is considered a loss ofcontainment. Containment venting for pressure when not in an accident situation should not beconsidered.NMP1 Basis Reference(s):1. N1-EOP-4 Primary Containment Control2. N1-EOP-4.2 Hydrogen ControlPage 241 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and Basis3. N1-EOP-4.1 Primary Containment Venting4. NEI 99-01 CMT Loss 3BPage 242 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: C. IsolationDegradation Threat: LossThreshold:5. UNISOLABLE primary system leakage outside Primary Containment as indicated byexceeding EITHER:Maximum safe general area temperature of 1351FORMaximum safe area radiation of 8 R/hrBasis:Plant-SpecificThe presence of elevated general area temperatures or radiation levels in the ReactorBuilding (RB) may be indicative of UNISOLABLE primary system leakage outside thePrimary Containment. The EOP maximum safe values define this Containment barrierthreshold because they are indicative of problems in the secondary containment that arespreading and pose a threat to achieving a safe plant shutdown. This threshold addressesproblematic discharges outside Primary Containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concerncorrespond to the areas addressed in EOP-5 Detail S (ref. 1).A "Maximum Safe Value" is the highest value at which equipment necessary for the safeshutdown of the plant will operate and personnel can perform any actions necessary forthe safe shutdown of the plant (ref. 2).The maximum safe value for temperature is 1350F.The maximum safe value for radiation is 8 R/hr.In general, multiple indications should be used to determine if a primary system isdischarging outside Primary Containment. For example, a high area radiation conditiondoes not necessarily indicate that a primary system is discharging into the secondarycontainment since this may be caused by radiation shine from nearby steam lines or themovement of radioactive materials. Conversely, a high area radiation condition inPage 243 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and Basisconjunction with other indications (e.g. room flooding, high area temperatures, reports ofsteam in the secondary containment, an unexpected rise in feedwater flowrate, orunexpected main turbine control valve closure) may indicate that a primary system isdischarging into the secondary containment.GenericThis threshold addresses incomplete containment isolation that allows direct release to theenvironment.In addition, The presence of area radiation or temperature Maximum Safe Values indicatingUNISOLABLE primary system leakage outside the Primary Containment are addressed after acontainment isolation. The indicators should be confirmed to be caused by RCS leakage.There is no Potential Loss threshold associated with this item.NMP1 Reference(s):1. N1-EOP-5 Secondary Containment Control2. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document3. NEI 99-01 CMT Loss 30Page 244EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: D. RadDegradation Threat: LossThreshold:[NonePage 245EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: E. JudgmentDegradation Threat: LossThreshold:6. ANY condition in the opinion of the Emergency Director that indicates loss of theContainment barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the Containment barrier is lost. Such a determination should includeIMMINENT barrier degradation, barrier monitoring capability and dominant accidentsequences.* IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to recognition of the inability to reach safety acceptance criteriabefore completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericPage 246 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the Containment barrier is lost. In addition, the inability to monitor the barriershould also be incorporated in this threshold as a factor in Emergency Director judgment that thebarrier may be considered lost.The Containment barrier should not be declared lost based on exceeding Technical Specificationaction statement criteria, unless there is an event in progress requiring mitigation by theContainment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Cladand/or RCS) the Containment barrier status is addressed by Technical Specifications.NMPI Basis Reference(s):1. NEI 99-01 CMT Loss 6Page 247 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: A. RPV Water LevelDegradation Threat: Potential LossThreshold:1. Primary Containment Flooding is requiredBasis:Plant-SpecificRequirements for Primary Containment Flooding are established in EOP-2 Step L-18; EOP-3 Steps L-8, L-10 and L-13; and EOP-7 Override 3 and 18. These EOPs provideinstructions to ensure adequate core cooling by maintaining RPV water level aboveprescribed limits or operating sufficient RPV injection sources when level cannot bedetermined. SAP entry is required when (ref. 1, 2. 3, 4):RPV water level cannot be restored and maintained above -109 in. with insufficientCore Spray flow: The Minimum Steam Cooling RPV Water Level (MSCRWL) is thelowest RPV water level at which the covered portion of the reactor core willgenerate sufficient steam to preclude any clad temperature in the uncovered portionof the core from exceeding 15000F. Core spray flow is insufficient if you cannotrestore and maintain both Core Spray loop flows at or above 180 x 104 Ibm/hr.Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that the parameter cannot be restored and maintained above the limitmay be made at, before, or after the parameter actually decreases to this point.RPV water level cannot be determined and it is determined that core damage isoccurring: When RPV water level cannot be determined, EOPs require RPVflooding strategies. RPV water level indication provides the primary means ofknowing if adequate core cooling is being maintained. When all means ofdetermining RPV water level are unavailable, reliance on alternate means ofassuring adequate core cooling must be attempted. The instructions in EOP-7specify these means, which include blowdown of the RPV and injection into thePage 248 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisRPV at a rate needed to flood to the elevation of the main steam lines or hold RPVpressure above the Minimum Steam Cooling Pressure (in ATWS events)This threshold is also a Loss of the Fuel Clad barrier (FC Loss A.1). Since PrimaryContainment Flooding occurs after core uncovery has occurred a Loss of the RCS barrierexists (RCS Loss A.1). Primary Containment Flooding (SAP entry), therefore, represents aLoss of two barriers and a Potential Loss of a third, which requires a General Emergencyclassification.GenericThere is no Loss threshold associated with this item.The potential loss requirement for drywell flooding indicates adequate core cooling cannot beestablished and maintained and that core melt is possible. Entry into Primary ContainmentFlooding procedures (SAPs) is a logical escalation in response to the inability to maintain adequatecore cooling.The condition in this potential loss threshold represents a potential core melt sequence which, ifnot corrected, could lead to vessel failure and increased potential for containment failure. Inconjunction with Reactor Vessel water level "Loss" thresholds in the Fuel Clad and RCS barriercolumns, this threshold will result in the declaration of a General Emergency -- loss of two barriersand the potential loss of a third.NMPI Basis Reference(s):1. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document2. N1-EOP-2 RPV Control3. N1-EOP-3 Failure to Scram4. 'N1-EOP-7 RPV Flooding5. NEI 99-01 CMT Potential Loss 2Page 249EPMP-EPP-01 01Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: Potential LossThreshold:2. Torus pressure > 35 psig and risingBasis:Plant-SpecificThe internal design pressure of the primary containment is identified by two pressures, adrywell pressure of 62 psig and a torus pressure of 35 psig. The more limiting of the twopressures defines this potential loss threshold. If this threshold is exceeded, a challenge tothe Primary Containment structure has occurred because assumptions used in theaccident analysis are no longer VALID and an unanalyzed condition exists (ref. 1). Thisconstitutes a Potential Loss of the Containment barrier even if a containment breach hasnot occurred.GenericThe torus pressure of 35 psig is based on the torus internal design pressure.NMP1 Basis Reference(s):1. USAR Section Vt.B.1.2 Design Basis Accident (DBA)2. NEI 99-01 CMT Potential Loss 1APage 250 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: Potential LossThreshold:3. Explosive mixture exists inside Primary Containment (>: 6% H2 and -> 5% 02)Basis:Plant-SpecificExplosive (deflagration) mixtures in the Primary Containment are assumed to be elevatedconcentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generationfor development of EOPs/SAPs indicates that any hydrogen concentration above minimumdetectable is not to be expected within the short term. Post-LOCA hydrogen generationprimarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogenconcentrations that rapidly develop are most likely caused by metal-water reaction. Ametal-water reaction is indicative of an accident more severe than accidents considered inthe plant design basis and would be indicative, therefore, of a potential threat to PrimaryContainment integrity. Hydrogen concentration of approximately 6% is considered theglobal deflagration concentration limit (ref. 1).Except for brief periods during plant startup and shutdown, oxygen concentration in thePrimary Containment is maintained at insignificant levels by nitrogen inertion. Thespecified values for this Potential Loss threshold are the minimum global deflagrationconcentration limits (6% hydrogen and 5% oxygen, ref. 1) and readily recognizablebecause 6% hydrogen is well above the EOP-4.2 entry condition (ref. 2). The minimumglobal deflagration hydrogen/oxygen concentrations (6%/5%, respectively) requireintentional Primary Containment venting, which is defined to be a Loss of Containment(PC Loss C.4).If the hydrogen or oxygen monitor is unavailable, sampling and analysis may determinegas concentrations. The validity of sample results must be judged based upon plantconditions, since drawing and analyzing samples may take some time. If sample resultsPage 251 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and Basiscannot be relied upon and hydrogen concentrations cannot be determined by any othermeans, the concentrations must be considered "unknown." The monitors should not beconsidered "unavailable" until an attempt has been made to place them in service. (ref. 1)GenericBWRs specifically define the limits associated with explosive mixtures in terms of deflagrationconcentrations of hydrogen and oxygen.NMP1 Basis Reference(s):1. NER-1 M-095, NMP1 Emergency Operating Procedures (EOP) Basis Document2. N1-EOP-4.2 Hydrogen Control3. NEI 99-01 CMT Potential Loss 1 BPage 252EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:ContainmentCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: Potential LossThreshold:4. Torus water temperature and RPV pressure cannot be maintained below the HeatCapacity Temperature Limit (N1-EOP-4 Figure M)Basis:Plant-SpecificThe Heat Capacity Temperature Limit (HCTL) is given in N1-EOP-4 Figure M. Thisthreshold is met when N1-EOP-4 Step TT-5 is reached and RPV blowdown is required(ref. 1).GenericThe Heat Capacity Temperature Limit (HCTL) is the highest torus water temperature from whichEmergency RPV Depressurization will not raise:" Torus temperature above the design value (2050F),OR* Torus pressure above Primary Containment Pressure Limit, before the rate of energytransfer from the RPV to the containment is greater than the capacity of the containmentvent.The HCTL is a function of RPV pressure and suppression pool water level. It is utilized to precludefailure of the containment and equipment in the containment necessary for the safe shutdown ofthe plant and therefore, the inability to maintain plant parameters below the limit constitutes apotential loss of Containment.NMP1 Basis Reference(s):1. N1-EOP-4 Primary Containment Control2. NEI 99-01 CMT Potential Loss 1CPage 253EPMP:EPP-0101Rev 00 Draft A -Fission Product Barrier Loss I Potential Loss Matrix and BasisBarrier: ContainmentCategory: D. RadDegradation Threat: Potential LossThreshold:S5. Drywell radiation __4.0E4 RhBasis:Plant-SpecificIt is important to recognize that the radiation monitor may be sensitive to shine from theRPV or RCS piping (caused by lower than normal RPV water level for example). TheDrywell High Range Radiation Monitors are the following (ref. 1):" RAM 201.7-36 Located: Az 3400, El 263' 6"" RAM 201.7-37 Located: Az 3100, El 301' 0"The Drywell High Range Radiation Monitors have a range of 1 E0 to 1 E8 R/hr on recorderRR 201.7-36C pens 1 and 2 (ref. 1).The threshold value (4.1 E4 R/hr rounded to 4.0E4 R/hr) was calculated assuming theinstantaneous release and dispersal of the reactor coolant noble gas and iodine inventoryassociated with 20% fuel clad damage into the drywell atmosphere (ref. 2).GenericThe 4.0 E4 R/hr reading is a value that indicates significant fuel damage well in excess of thatrequired for loss of RCS and Fuel Clad.Regardless of whether containment is challenged, this amount of activity in containment, ifreleased, could have such severe consequences that it is prudent to treat this as a potential loss ofcontainment, such that a General Emergency declaration is warranted.There is no Loss threshold associated with this item.NMPI Basis Reference(s):1. N1-RSP-10C The Use and Routine Calibration of the General Atomic High RangeGamma Radiation Monitoring System2. Calculation 1H21C003, Rev. 0 CCN 008830, Rev. 03. NEI 99-01 CMT Potential Loss 4Page 254 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: E. JudgmentDegradation Threat: Potential LossThreshold:6. ANY condition in the opinion of the Emergency Director that indicates potential loss ofthe Containment barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the Containment barrier is potentially lost. Such a determination shouldinclude IMMINENT barrier degradation, barrier monitoring capability and dominantaccident sequences." IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to recognition of the inability to reach safety acceptance criteriabefore completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericPage 255 EPMP-EPP-0101Rev 00 Draft A -Fission Product Barrier Loss / Potential Loss Matrix and BasisThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the Containment barrier is potentially lost. In addition, the inability to monitorthe barrier should also be incorporated in this threshold as a factor in Emergency Directorjudgment that the barrier may be considered potentially lost.The Containment barrier should not be declared potentially lost based on exceeding TechnicalSpecification action statement criteria, unless there is an event in progress requiring mitigation bythe Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Cladand/or RCS) the Containment barrier status is addressed by Technical Specifications.NMPI Basis Reference(s):1. NEI 99-01 CMT Potential Loss 6Page 256EPMP-EPP-0101Rev 00 Draft A NINE MILE POINT NUCLEAR STATIONEMERGENCY PLAN MAINTENANCE PROCEDUREEPMP-EPP-01 02REVISION 00 (Draft RAI 1 12)UNIT 2 EMERGENCY CLASSIFICATION TECHNICAL BASESTECHNICAL SPECIFICATION REQUIREDApproved by:J. KaminskiDirector Emergency PlanningTHIS IS A COMPLETE REVISIONDateEffective Date:PERIODIC REVIEW DUE DATE:

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SECTION1.02.03.04.05.06.06.1Table of ContentsTITLE PAGEP U R P O S E ..............................................................................................D IS C U S S IO N .....................................................................................2 .1 B ackground ................................................................................2.2 Fission Product Barriers .................................2.3 Emergency Classification Based on Fission ProductB arrier D egradation .....................................................................2.4 EA L Relationship to EO Ps ...........................................................2.5 Symptom-Based vs. Event-Based Approach ...............................2.6 EA L O rganization ........................................................................2.7 Technical Bases Inform ation .......................................................2.8 O perating M ode Applicability .......................................................2.9 Validation of Indications, Reports and Conditions .......................2.10 Planned vs. UNPLANNED Events ...............................................2.11 Classifying Transient Events ......................................................2.12 Multiple Simultaneous Events and IMMINENT EAL Thresholds..2.13 Emergency Classification Level DowngradingR E F E R E N C E S .......................................................................................3.1 D evelo pm ental .............................................................................3 .2 Im plem enting ...............................................................................3.3 C om m itm ents .........................................................................D E F IN IT IO N S .........................................................................................NMP2-TO-NEI 99-01 EAL CROSSREFERENCE ..................................A TTA C H M E N T S ....................................................................................Attachment 1 -Emergency Action Level Technical Bases ....................Cateqory R Abnormal Radiation Levels / Radiological Effluents ..........R G 1 .1 ...............................................................................R G 1 .2 ...............................................................................R G 1 .3 ...............................................................................R S I .1 ................. ........................................................ ..RS1.2.. ...................................RS1.3....................................RA1.1.....................................RA1.2 ................ ....... .............RA1.3.. ...................................RU1.1............................... ......RU1.2....................................R U 1 .3 ................................................................................Page iv EPMP-EPP-01 02Rev 00 (Draft A)

SECTION TITLECategory RCategory HCategory ETable of ContentsPAGE(cont'd)R A 2 .1 ................................................................................R A 2 .2 ................................................................................R U 2 .1 ................................................................................R U 2 .2 ................................................................................R A 3 ................................................................................Hazards and Other Conditions Affecting Plant Safety ......H A 1 .1 ...............................................................................H A 1 .2 ...............................................................................H A 1 .3 ...............................................................................H A 1 .4 ...............................................................................H A 1 .5 ...............................................................................H A 1 .6 ...............................................................................H U I .1 ...............................................................................HU1.2 ..........................................................................H U 1 .3 ...............................................................................H U 1 .4 ...............................................................................H U 1 .5 ...............................................................................H A 2 .1 ...............................................................................H U 2 .1 ...............................................................................H U 2 .2 ...............................................................................H A 3 ...............................................................................H U 3 .........................................................HU3.2 .......................................................... ........HG4.1 ........................................................... ......HG4.2 ....................................................... ........HS4.1 ......................................................... ........H A 4 .1 ...............................................................................H U 4 .1 ...............................................................................H S 5 .1 ...............................................................................H G 6 .1 ..............................................................................H S 6 .1 ...............................................................................H A 5 .1 ...............................................................................H A 6 .1 ...............................................................................H U 6 .1 ...............................................................................IS F S I .................................................................................E U 1 .1 ...............................................................................Page vEPMP-EPP-0102Rev 00 (Draft A)

Table of ContentsSECTION TITLECategory CCategory SPAGECold Shutdown / Refueling System Malfunction ...............C A l .1 ................................................................................C U I .1 ................................................................................C U 2 .1 ................................................................................C G 3 .1 ..............................................................................C G 3 .2 ..............................................................................C S 3 .1 ...............................................................................C S 3 .2 ...............................................................................C S 3 .3 ...............................................................................C A 3 .1 ................................................................................CU3.1 .................................................................CU3.2 .......................................................... ........CU3.3 .......................................................... ..........CA4.1 ......................................................... .......CU4.1 ......................................................... .........CU4.2 ........................................................ ........CU5.1 ......................................................... ......C U 6 .1 ...............................................................................System Malfunction ..........................................................S G 1 .1 ...............................................................................S S 1 .1 ...............................................................................S A 1 .1 ...............................................................................S U 1 .1 ................................................................................S S 2 .1 ...............................................................................S G 3 .1 ...............................................................................S S 3 .1 ...............................................................................SA3.1 ..........................................S U 3 .1 ...............................................................................S U 4 .1 ...............................................................................S S 5 .1 ...............................................................................S A 5 .1 ...............................................................................S U 5 .1 ...............................................................................S U 6 .1 ...............................................................................S U 7 .1 ...............................................................................S U 7 .2 ...............................................................................S U 8 .1 ...............................................................................Page viEPMP-EPP-0102Rev 00 (Draft A)

Table of ContentsSECTION TITLE PAGECatecqorV F Fission Product Barrier Degradation ................................F G 1 .1 ...............................................................................F S 1 .1 ...............................................................................F A I .1 ...............................................................................F U I .1 ...............................................................................6.2 Attachment 2 -Fission Product Barrier Loss / Potential LossMatrix and Bases ...................................................................................FC Loss A.1 .....................................................................FC Loss D.2 ....................................................................FC Loss D.3 .....................................................................FC Loss EA .....................................................................FC Potential Loss A.1 ......................................................FC Potential Loss E.2 ......................................................RCS Loss A.1 ..................................................................RCS Loss B.2 ..................................................................RCS Loss C.3 ..................................................................RCS Loss C0 ..................................................................RCS Loss D.5 ..................................................................RCS Loss E.6 ..................................................................RCS Potential Loss C.1 ...................................................RCS Potential Loss C.2 ...................................................RCS Potential Loss E.3 ...................................................PC Loss B.1 ....................................................................PC Loss B.2 .....................................................................PC Loss C.3 .....................................................................PC Loss C0 .....................................................................PC Loss C.5 ....................................PC Loss E.6 .....................................................................PC Potential Loss A.1 ......................................................PC Potential Loss B.2 ......................................................PC Potential Loss B.3 ......................................................PC Potential Loss BA ......................................................PC Potential Loss C.5 ......................................................PC Potential Loss E.6 ......................................................Page vii EPMP-EPP-01 02Rev 00 (Draft A)

ABBEVIATIONS / ACRONYMSAC ................................................................................................................. Alternating CurrentAPRM ............................................................................................. Average Power Range MeterATW S ................................................................................. Anticipated Transient W ithout ScramBW R .......................................................................................................... Boiling W ater ReactorCDE .................................................................................................. Com m itted Dose EquivalentCFR ................................................................................................. Code of Federal RegulationsDC .......................................................................................................................... Direct CurrentEAL ......................................................................................................... Emergency Action LevelECCS ........................................................................................ Emergency Core Cooling SystemED ................................................................................................................. Emergency Directorel ..................................................................................................................................... elevationEOF .............................................................................................. Emergency Operations FacilityEOP .......................................................................................... Emergency Operating ProcedureEPA .......................................................................................... Environmental Protection AgencyEPMP .......................................................................... Emergency Plan Maintenance ProcedureEPRI ......................................................................................... Electric Power Research InstituteFAA ............................................................................................. Federal Aviation Adm inistrationFBI ............................................................................................... Federal Bureau of InvestigationFEMA ........................................................................... Federal Emergency Management AgencyGE ................................................................................................................ General EmergencyGTS ............................................................................................ Standby Gas Treatm ent SystemHCTL ......................................................................................... Heat Capacity Tem perature Lim itHOO .................................... Headquarters (NRC) Operations OfficerHPCS ................................................................................................... High Pressure Core SprayIC ..................................................................................................................... Initiating ConditionISFSI ................................................ INDEPENDENT SPENT FUEL STORAGE INSTALLATIONJAFNPP ........................................................... Jam es A. FitzPatrick Nuclear Power PlantLCO ............................................................................................. Lim iting Condition of OperationLOCA ..................................................................................................... Loss of Coolant AccidentMSIV .................................................................................................. Main Steam Isolation ValveMSL ................................................................................................................... Main Steam Linem R .......................................................................................................................... m illiRoentgenMSCP .......... .......................................................................... Minim um Steam Cooling PressureMSCRW L ................................................................... Minim um Steam Cooling RPV W ater LevelMSIV .................................................................................................. Main Steam Isolation ValveMSL ................................................................................................................... Main Steam LineNEI ......................................................................................................... Nuclear Energy InstituteNESP .............................................................................. National Environmental Studies ProjectNRC ........................................................................................... Nuclear Regulatory Com m issionPage viii EPMP-EPP-0102Rev 00 (Draft A)

ACRONYMS & ABBREVIATIONS (continued)NORAD ............................................................... North American Aerospace Defense CommandNUMARC ............................................................... Nuclear Management and Resources CouncilOBE .................................................................................................. Operating Basis EarthquakeODCM ....................................................................................... Off-site Dose Calculation ManualORO ............................................................................................ Off-site Response OrganizationPAG ................................................................................................... Protective Action GuidelinePC ..................................................................................... ......................... Primary ContainmentPGCC ..................................................................................... Power Generator Control ComplexPSIG ........................................................................................... Pounds per Square Inch GaugeR ................................................................................................................................... R o e n tg e nRB ...................................................................................................................... Reactor BuildingRCIC ............................................................................................. Reactor Core Isolation CoolingRCS ....................................................................................................... Reactor Coolant SystemRE ................................................................................................................... Radiation ElementRem ..................................................................................................... Roentgen Equivalent ManRHR ......................................................................................................... Residual Heat RemovalRMS ............................................................................................... Radiation Monitoring SystemRPS .................................................................................................... Reactor Protection SystemRPV ....................................................................................................... Reactor Pressure VesselRW CU ..................................................................................................... Reactor W ater CleanupSAE ............................................................................................................ Site Area EmergencySC .......................................................................................................... Secondary ContainmentSPDS ....................................................................................... Safety Parameter Display SystemTEDE .......................................................................................... Total Effective Dose EquivalentTSC ...................................................................................................... Technical Support CenterUE ......................................................................................................................... Unusual EventUSAR .......................................................................................... Updated Safety Analysis ReportPage ix EPMP-EPP-0102Rev 00 (Draft A) 1.0 PURPOSEThis document provides an explanation and rationale for each Emergency Action Level(EAL) included in the EAL Upgrade Project for Nine Mile Point Nuclear Station Unit 2(NMP2). It should be used to facilitate review of the NMP2 EALs and provide historicaldocumentation for future reference. Decision-makers responsible for implementation ofEPIP-EPP-02, "Classification of Emergency Conditions at Unit 2," and the EmergencyAction Level Matrices, may use this document as a technical reference in support of EALinterpretation. This information may assist the Emergency Director in makingclassifications, particularly those involving judgment or multiple events. The basisinformation may also be useful in training, for explaining event classifications to offsiteofficials, and facilitates regulatory review and approval of the classification scheme.The expectation is that emergency classifications are to be made as soon as conditionsare present and recognizable for the classification, but within 15 minutes in all cases ofconditions present. Use of this document for assistance is not intended to delay theemergency classification.2.0 DISCUSSION2.1 BackgroundEALs are the plant-specific indications, conditions or instrument readings that are utilizedto classify emergency conditions defined in the Nine Mile Point Site Emergency Plan.In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development ofEmergency Action Levels" as an alternative to NUREG-0654 EAL guidance.NEI 99-01 (NUMARC/NESP-007) Revision 4 was subsequently issued for industryimplementation. Enhancements over earlier revisions included:* Consolidating the system malfunction initiating conditions and example emergencyaction levels which address conditions that may be postulated to occur during plantshutdown conditions." Initiating conditions and example emergency action levels that fully addressconditions that may be postulated to occur at permanently Defueled Stations andINDEPENDENT SPENT FUEL STORAGE INSTALLATIONs (ISFSIs).Page 1 EPMP-EPP-0102Rev 00 (Draft A)

, Simplifying the fission product barrier EAL threshold for a Site Area Emergency.Subsequently, Revision 5 of NEI 99-01 has been issued which incorporates resolutions tonumerous implementation issues including the NRC EAL FAQs. Using NEI 99-01 Revision5 Final, February 2008 (ADAMS Accession Number ML080450149), NMP2 conducted anEAL implementation upgrade project that produced the EALs discussed herein.2.2 Fission Product BarriersMany of the EALs derived from the NEI methodology are fission product barrier based.That is, the conditions that define the EALs are based upon loss or potential loss of one ormore of the three fission product barriers. "Loss" and "Potential Loss" signify the relativedamage and threat of damage to the barrier. "Loss" means the barrier no longer assurescontainment of radioactive materials; "potential loss" implies an increased probability ofbarrier loss and decreased certainty of maintaining the barrier.The primary fission product barriers are:A. Fuel Clad (FC): Zirconium tubes which house the ceramic uranium oxide pelletsalong with the end plugs which are welded into each end of the fuel rods comprisethe FC barrier.B. Reactor Coolant System (RCS): The reactor vessel shell, vessel head, CRDhousings, vessel nozzles and penetrations, and all primary systems directlyconnected to the RPV up to the outermost Primary Containment isolation valvecomprise the RCS barrier.C. Containment (PC): The drywell, the suppression chamber/pool, their respectiveinterconnecting paths, and other connections up to and including the outermostcontainment isolation valves comprise the Primary Containment barrier.2.3 Emergency Classification Based on Fission Product Barrier DegradationThe following criteria are the bases for event classification related to fission product barrierloss or potential loss:Unusual Event:Any loss or any potential loss of ContainmentAlert:Any loss or any potential loss of either Fuel Clad or RCSSite Area Emergency:Page 2 EPMP-EPP-0102Rev 00 (Draft A)

Loss or potential loss of any two barriersGeneral Emergency:Loss of any two barriers and loss or potential loss of third barrier2.4 EAL Relationship to EOPsWhere possible, the EALs have been made consistent with and utilize the conditionsdefined in the NMP2 Emergency Operating Procedures (EOPs). While the symptoms thatdrive operator actions specified in the EOPs are not indicative of all possible conditionswhich warrant emergency classification, they define the symptoms, independent ofinitiating events, for which reactor plant safety and/or fission product barrier integrity arethreatened. When these symptoms are clearly representative of one of the NEI InitiatingConditions, they have been utilized as an EAL. This permits rapid classification ofemergency situations based on plant conditions without the need for additional evaluationor event diagnosis. Although some of the EALs presented here are based on conditionsdefined in the EOPs, classification of emergencies using these EALs is not dependentupon EOP entry or execution. The EALs can be utilized independently or in conjunctionwith the EOPs.2.5 Symptom-Based vs. Event-Based ApproachTo the extent possible, the EALs are symptom-based. That is, the action level threshold isdefined by values of key plant operating parameters that identify emergency or potentialemergency conditions. This approach is appropriate because it allows the full scope ofvariations in the types of events to be classified as emergencies. However, a purelysymptom-based approach is not sufficient to address all events for which emergencyclassification is appropriate. Particular events to which no predetermined symptoms can beascribed have also been utilized as EALs since they may be indicative of potentially moreserious conditions not yet fully realized.2.6 EAL OrganizationThe NMP2 EAL scheme includes the following features:* Division of the EAL set into three broad groups:Page 3 EPMP-EPP-0102Rev 00 (Draft A) o EALs applicable under all plant operating modes -This group would bereviewed by the EAL-user any time emergency classification is considered.o EALs applicable only under hot operating modes -This group would only bereviewed by the EAL-user when the plant is in Hot Shutdown, Startup orPower Operation mode.o EALs applicable only under cold operating modes -This group would only bereviewed by the EAL-user when the plant is in Cold Shutdown, Refuel orDefueled mode.The purpose of the groups is to avoid review of hot condition EALs when the plant isin a cold condition and avoid review of cold condition EALs when the plant is in ahot condition. This approach significantly minimizes the total number of EALs thatmust be reviewed by the EAL-user for a given plant condition, reduces EAL-userreading burden and, thereby, speeds identification of the EAL that applies to theemergency.Within each of the above three groups, assignment of EALs tocategories/subcategories -Category and subcategory titles are selected torepresent conditions that are operationally significant to the EAL-user.Subcategories are used as necessary to further divide the EALs of a category intological sets of possible emergency classification thresholds. The NMP2 EALcategories/subcategories and their relationship to NEI 99-01 Rev. 5 RecognitionCategories are listed below.Page 4 EPMP-EPP-01 02Rev 00 (Draft A)

EAL Groups, Categories and SubcategoriesEAL Group/Category EAL SubcategoryAny Operating Mode:R -Abnormal Radiation Levels / 1 -Offsite Rad ConditionsRadiological Effluents 2 -Onsite Rad Conditions & Spent Fuel Events3 -CR/CAS RadH -Hazards and Other Conditions 1 -Natural or Destructive PhenomenaAffecting Plant Safety 2 -FIRE or EXPLOSION3 -Hazardous Gas4 -Security5 -Control Room Evacuation6 -JudgmentE- ISFSI NoneCold Conditions:C -Cold Shutdown I Refueling System 1 -Loss of AC PowerMalfunction 2 -Loss of DC Power3 -RPV Level4 -RCS Temperature5 -Inadvertent Criticality6 -CommunicationsHot Conditions:S -System Malfunction 1 -Loss of AC Power2 -Loss of DC Power3 -Criticality & RPS Failure4 -Inability to Reach or Maintain Shutdown Conditions5 -Instrumentation6 -Communications7 -Fuel Clad Degradation8 -RCS LeakageF -Fission Product Barrier Degradation NonePage 5 EPMP-EPP-0102Rev 00 (Draft A)

The primary tool for determining the emergency classification level is the EALClassification Matrix. The user of the EAL Classification Matrix may (but is not required to)consult the EAL Technical Bases Document in order to obtain additional informationconcerning the EALs under classification consideration. The user should consult Sections2.7 and 2.8, and Attachments 1 and 2 of this document for such information.2.7 Technical Bases InformationEAL technical bases are provided in Attachment 1 for each EAL according to EAL group(Any, Hot, Cold), EAL category (R, H, E, C, S and F) and EAL subcategory. A summaryexplanation of each category and subcategory is given at the beginning of the technicalbases discussions of the EALs included in the category. For each EAL, the followinginformation is provided:Category Letter & TitleSubcategory Number & TitleInitiatinq Condition (IC)Site-specific description of the generic IC given in NEI 99-01 Rev. 5.EAL Identifier (enclosed in rectangle)Each EAL is assigned a unique identifier to support accurate communication of theemergency classification to onsite and offsite personnel. Four characters define eachEAL identifier:1. First character (letter): Corresponds to the EAL category as described above (R,H, E, C, S or F)2. Second character (letter): The emergency classification (G, S, A or U)G = General EmergencyS = Site Area EmergencyA = AlertU = Unusual Event3. Third character (number): Subcategory number within the given category.Subcategories are sequentially numbered beginning with the number one (1). Ifa category does not have a subcategory, this character is assigned the numberone (1).Page 6 EPMP-EPP-0102Rev 00 (Draft A)

4. Fourth character (number): The numerical sequence of the EAL within the EALsubcategory. If the subcategory has only one EAL, it is given the number one(1).Classification (enclosed in rectangqle):Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)EAL (enclosed in rectangqle)Wording enclosed in the rectangle appears as it is displayed in the EAL ClassificationMatrix. Selected terms are highlighted for emphasis:" Bold, uppercase print is assigned to: "ANY," EAL identifiers, and logic termssuch as AND, OR, EITHER, etc. (When used as conjunctions, the words "and"and "or" are not highlighted.)* Bold, mixed case print is assigned to: "all," "only," "both," table titles andcolumn headings, numbers following the word "ANY," and negative terms (e.g.,"not," "cannot," etc.)* Uppercase print is assigned to acronyms, abbreviations, and terms defined inSection 4.0.Mode ApplicabilityOne or more of the following plant operating conditions comprise the mode to whicheach EAL is applicable: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -ColdShutdown, 5 -Refuel, D -Defueled, or All. (See Section 2.8 for operating modedefinitions.)Basis:A Generic basis section provides a description of the rationale for the EAL as providedin NEI 99-01 Rev. 5. This is followed by a Plant-Specific basis section that providesNMP2-relevant information concerning the EAL.NMP2 Basis Reference(s):Site-specific source documentation from which the EAL is derived2.8 Operating Mode Applicability (Technical Specifications Table 1.1-1)1 Power OperationPage 7 EPMP-EPP-0102Rev 00 (Draft A)

Reactor mode switch is in RUN2 StartupThe mode switch is in STARTUP/HOT STANDBY or REFUEL with all reactor vesselhead closure bolts fully tensioned3 Hot ShutdownThe mode switch is in SHUTDOWN, average reactor coolant temperature is >2000F, and all reactor vessel head closure bolts are fully tensioned4 Cold ShutdownThe mode switch is in SHUTDOWN, average reactor coolant temperature is _2000F, and all reactor vessel head closure bolts are fully tensioned5 RefuelThe mode switch is in SHUTDOWN or REFUEL, and one or more reactor vesselhead closure bolts are less than fully tensionedD DefueledAll reactor fuel is removed from the RPV (full core off load during refueling orextended outage)The plant operating mode that exists at the time that the event occurs (prior to anyprotective system or operator action is initiated in response to the condition) should becompared to the mode applicability of the EALs. If a lower or higher plant operating modeis reached before the emergency classification is made, the declaration shall be based onthe mode that existed at the time the event occurred.2.9 Validation of Indications, Reports and ConditionsAll emergency classifications shall be based upon VALID indications, reports or conditions.An indication, report, or condition, is considered to be VALID when it is verified by (1) aninstrument channel check, or (2) indications on related or redundant indicators, or (3) bydirect observation by plant personnel, such that doubt related to the indicator's operability,the condition's existence, or the report's accuracy is removed. Implicit in this definition isthe need for timely assessment.2.10 Planned vs. UNPLANNED EventsPage 8 EPMP-EPP-0102Rev 00 (Draft A)

Planned evolutions involve preplanning to address the limitations imposed by thecondition, the performance of required surveillance testing, and the implementation ofspecific controls prior to knowingly entering the condition in accordance with the specificrequirements of the site's Technical Specifications. Activities which cause the site tooperate beyond that allowed by the site's Technical Specifications, planned orUNPLANNED, may result in an EAL threshold being met or exceeded. Planned evolutionsto test, manipulate, repair, perform maintenance or modifications to systems andequipment that result in an EAL value being met or exceeded are not subject toclassification and activation requirements as long as the evolution proceeds as plannedand is within the operational limitations imposed by the specific operating license.However, these conditions may be subject to the reporting requirements of 10 CFR 50.72.2.11 Classifying Transient EventsFor some events, the condition may be corrected before a declaration has been made.The key consideration in this situation is to determine whether or not further plant damageoccurred while the corrective actions were being taken. In some situations, this can bereadily determined, in other situations, further analyses may be necessary (e.g., coolantradiochemistry following an ATWS event, plant structural examination following anearthquake, etc.). Classify the event as indicated and terminate the emergency onceassessment shows that there were no consequences from the event and other terminationcriteria are met.Existing guidance for classifying transient events addresses the period of time of eventrecognition and .classification (15 minutes). However, in cases when EAL declarationcriteria may be met momentarily during the normal expected response of the plant,declaration requirements should not be considered to be met when the conditions are apart of the designed plant response, or result from appropriate Operator actions.There may be cases in which a plant condition that exceeded an EAL was not recognizedat the time of occurrence but is identified well after the condition has occurred (e.g., as aresult of routine log or record review), and the condition no longer exists. In these cases,an emergency should not be declared. Reporting requirements of 10 CFR 50.72 areapplicable and the guidance of NUREG-1 022, Event Reporting Guidelines 10 CFR 50.72and 50.73, should be applied.Page 9 EPMP-EPP-0102Rev 00 (Draft A) 2.12 Multiple Simultaneous Events and IMMINENT EAL ThresholdsWhen multiple simultaneous events occur, the emergency classification level is based onthe highest EAL reached. For example, two Alerts remain in the Alert category. Or, an Alertand a Site Area Emergency is a Site Area Emergency. Further guidance is provided inRIS 2007-02, Clarification of NRC Guidance for Emergency Notifications During QuicklyChanging Events.Since NMP2 is at a multi-unit site, emergency classification level upgrading must alsoconsider the effects of a loss of a common system on more than one unit (e.g., potentialfor radioactive release from more than one core).Although the majority of the EALs provide very specific thresholds, the Emergency Director(ED) must remain alert to events or conditions that lead to the conclusion that exceedingthe EAL threshold is IMMINENT. If, in the judgment of the ED, an IMMINENT situation is athand, the classification should be made as if the threshold has been exceeded. While thisis particularly prudent at the higher emergency classes (the early classification may permitmore effective implementation of protective measures), it is nonetheless applicable to allemergency classes.2.13 Emergency Classification Level DowngradingAnother important aspect of usable EAL guidance is the consideration of what to do whenthe risk posed by an emergency is clearly decreasing. A combination approach involvingrecovery from General Emergencies and some Site Area Emergencies and terminationfrom Unusual Events, Alerts, and certain Site Area Emergencies causing no long termplant damage appears to be the best choice. Downgrading to lower emergencyclassification levels adds notifications but may have merit under certain circumstances.Page 10 EPMP-EPP-0102Rev 00 (Draft A) 3.0 REFERENCES3.1 Developmental3.1.1 NEI 99-01 Rev. 5 Final, Methodology for Development of EmergencyAction Levels, February 2008, ADAMS Accession Number ML0804501493.1.2 NRC Regulatory Issue Summary (RIS) 2003-18, Supplement 2, Use ofNuclear Energy Institute (NEI) 99-01, Methodology for Development ofEmergency Action Levels Revision 4, Dated January 2003 (December 12,2005)3.1.3 RIS 2007-02 Clarification of NRC Guidance for Emergency NotificationsDuring Quickly Changing Events3.1.4 Nine Mile Point Site Emergency Plan3.2 Implementing3.2.1 EPIP-EPP-02 Classification of Emergency Conditions at Unit 23.2.2 EAL Comparison Matrix3.3 CommitmentsNonePage 11 EPMP-EPP-0102Rev 00 (Draft A) 4.0 DEFINITIONS (ref. 3.1.1 except as noted)AFFECTING SAFE SHUTDOWNEvent in progress has adversely affected functions that are necessary to bring the plant toand maintain it in the applicable hot or cold shutdown condition. Plant conditionapplicability is determined by Technical Specification LCOs in effect.Example 1: Event causes damage that results in entry into an LCO that requires theplant to be placed in hot shutdown. Hot shutdown is achievable, but cold shutdown isnot. This event is not "AFFECTING SAFE SHUTDOWN."Example 2: Event causes damage that results in entry into an LCO that requires theplant to be placed in cold shutdown. Hot shutdown is achievable, but cold shutdown isnot. This event is "AFFECTING SAFE SHUTDOWN."AIRLINER/LARGE AIRCRAFTAny size or type of aircraft with the potential for causing significant damage to the plant(refer to the Security Plan for a more detailed definition).BOMBRefers to an explosive device suspected of having sufficient force to damage plantsystems or structures.CIVIL DISTURBANCEA group of people violently protesting station operations or activities at the site.CONFINEMENT BOUNDARYThe barrier(s) between areas containing radioactive substances and the environment.CONTAINMENT CLOSUREThe procedurally defined actions taken to secure containment (primary or secondary) andits associated structures, systems, and components as a functional barrier to fissionproduct release under existing plant conditions.EXPLOSIONA rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energizedequipment that imparts energy of sufficient force to potentially damage permanentstructures, systems, or components.EXTORTIONAn attempt to cause an action at the station by threat of force.FIRECombustion characterized by heat and light. Sources of smoke such as slipping drive beltsor overheated electrical equipment do not constitute FIREs. Observation of flame ispreferred but is not required if large quantities of smoke and heat are observed.HOSTAGEA person(s) held as leverage against the station to ensure that demands will be met by thePage 12 EPMP-EPP-0102Rev 00 (Draft A) station.HOSTILE ACTIONAn act toward NMP2 or its personnel that includes the use of violent force to destroyequipment, take HOSTAGEs, and/or intimidate the licensee to achieve an end. Thisincludes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, orother devices used to deliver destructive force. Other acts that satisfy the overall intentmay be included.HOSTILE ACTION should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on NMP2. Non-terrorism-based EALsshould be used to address such activities, (e.g., violent acts between individuals in theowner controlled area).HOSTILE FORCEOne or more individuals who are engaged in a determined assault, overtly or by stealthand deception, equipped with suitable weapons capable of killing, maiming, or causingdestruction.IMMINENTMitigation actions have been ineffective, additional actions are not expected to besuccessful, and trended information indicates that the event or condition will occur. WhereIMMINENT timeframes are specified, they shall apply.INTACTThe RCS should be considered INTACT when the RCS pressure boundary is in its normalcondition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).INTRUSIONThe act of entering without authorization. Discovery of a BOMB in a specified area isindication of INTRUSION into that area by a HOSTILE FORCE.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)A complex that is designed and constructed for the interim storage of spent nuclear fueland other radioactive materials associated with spent fuel storage.NORMAL LEVELSAs applied to radiological IC/EALs, the highest reading in the past twenty-four hoursexcluding the current peak value.NORMAL PLANT OPERATIONSActivities at the plant site associated with routine testing, maintenance, or equipmentoperations, in accordance with normal operating or administrative procedures. Entry intoabnormal or emergency operating procedures, or deviation from normal security orradiological controls posture, is a departure from NORMAL PLANT OPERATIONS.PROJECTILEAn object directed toward NMP2 that could cause concern for its continued operability,reliability, or personnel safety.Page 13 EPMP-EPP-0102Rev 00 (Draft A)

PROTECTED AREAThe area which normally encompasses all controlled areas within the securityPROTECTED AREA fence. NMP1 and NMP2 share a common PROTECTED AREAborder. NMP1 and NMP2 PROTECTED AREA boundaries are illustrated in USAR Figure1.2-1.SABOTAGEDeliberate damage, mis-alignment, or mis-operation of plant equipment with the intent torender the equipment inoperable. Equipment found tampered with or damaged due tomalicious mischief may not meet the definition of SABOTAGE until this determination ismade by security supervision.SAFETY-RELATED STRUCTUREs, SYSTEMs and COMPONENTs (as defined in1 OCFR50.2)Those structures, systems and components that are relied upon to remain functionalduring and following design basis events to assure:(1) The integrity of the reactor coolant pressure boundary;(2) The capability to shut down the reactor and maintain it in a safe shutdown condition;(3) The capability to prevent or mitigate the consequences of accidents which couldresult in potential offsite exposures.SECURITY CONDITIONAny security event as listed in the approved security contingency plan that constitutes athreat/compromise to site security, threat/risk to site personnel, or a potential degradationto the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILEACTION.SITE BOUNDARYPer ODCM Figure D 1.0-1, the line around the Nine Mile Point Nuclear Station beyondwhich the land is not owned, leased or otherwise controlled by the owners and operators ofNine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant.STRIKE ACTIONWork stoppage within the PROTECTED AREA by a body of workers to enforcecompliance with demands made on NMP2. The STRIKE ACTION must threaten to interruptNORMAL PLANT OPERATIONS.UNISOLABLEA breach or leak that cannot be promptly isolated.UNPLANNEDA parameter change or an event, the reasons for which may be known or unknown, that isnot the result of an intended evolution or expected plant response to a transient.VALIDAn indication, report, or condition, is considered to be VALID when it is verified by (1) aninstrument channel check, or (2) indications on related or redundant indicators, or (3) byPage 14 EPMP-EPP-0102Rev. 00 (Draft A) direct observation by plant personnel, such that doubt related to the indicator's operability,the condition's existence, or the report's accuracy is removed. Implicit in this definition isthe need for timely assessment.VISIBLE DAMAGEDamage to equipment or structure that is readily observable without measurements,testing, or analysis. Damage is sufficient to cause concern regarding the continuedoperability or reliability of affected SAFETY-RELATED STRUCTURE, SYSTEM orCOMPONENT. Example damage includes: deformation due to heat or impact, denting,penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping,scratches) should not be included.VITAL AREAAny areas, normally within the NMP2 PROTECTED AREA, that contains equipment,systems, components, or material, the failure, destruction, or release of which coulddirectly or indirectly endanger the public health and safety by exposure to radiation.Page 15EPMP-EPP-0102Rev 00 (Draft A) 5.0 NMP2-TO-NEI 99-01 EAL CROSSREFERENCEThis cross-reference is provided to facilitate'association and location of a NMP2 EALwithin the NEI 99-01 IC/EAL identification scheme. Further information regarding thedevelopment of the NMP2 EALs based on the NEI guidance can be found in the EALComparison Matrix.NMP2 NEI 99-01EAL 11C ExampleEALRG1.1 AG1 1RG1.2 AG1 2RG1.3 AG1 4RS1.1 AS1 1RS1.2 AS1 2RS1.3 AS1 4RA1.1 AA1 1RA1.2 AA1 2RA1.3 AA1 3RU1.1 AU1 1RU1.2 AU1 2RU1.3 AU1 3RA2.1 AA2 2RA2.2 AA2 1RU2.1 AU2 1RU2.2 AU2 2RA3.1 AA3 1HA1.1 HA1 1HA1.2 HA1 2Page 16EPMP-EPP-0102Rev 00 (Draft A)

NMP2 NEI 99-01EAL IC ExampleEALHA1.3 HA1 3HA1.4 HA1 4HA1.5 HA1 6HA1.6 HA1 5HU1.1 HU1 1HU1.2 HU1 2HU1.3 HU1 3HU1.4 HU1 4HU1.5 HU1 5HA2.1 HA2 1HU2.1 HU2 1HU2.2 HU2 2HA3.1 HA3 1HU3.1 HU3 1HU3.2 HU3 2HG4.1 HG1 1HG4.2 HG1 2HS4.1 HS4 1HA4.1 HA4 1,2HU4.1 HU4 1,2,3HS5.1 HS2 1HA5.1 HA5 1HG6.1 HG2 1HS6.1 HS3 1HA6.1 HA6 1Page 17EPMP-EPP-0102Rev 00 (Draft A)

NMP2 NEI 99-01EAL IC ExampleEALHU6.1 HU5 1EU1.1 E-HU1 1CA1.1 CA3 1CU1.1 CU3 1CU2.1 CU7 1CG3.1 CG1 1CG3.2 CG1 2CS3.1 CS1 1CS3.2 CS1 2CS3.3 CS1 3CA3.1 CAl 1,2CU3.1 CUI 1CU3.2 CU2 1CU3.3 CU2 2CA4.1 CA4 1,2CU4.1 CU4 1CU4.2 CU4 2CU5.1 CU8 1CU6.1 CU6 1,2SG1.1 SG1 1SS1.1 SS1 1SAI.1 SA5 1SU1.l SUl 1SS2.1 SS3 1SG3.1 SG2 1Page 18EPMP-EPP-01 02Rev 00 (Draft A)

NMP2 NEI 99-01EAL IC ExampleEALSS3.1 SS2 1SA3.1 SA2 1SU3.1 SU8 1SU4.1 SU2 1SS5.1 SS6 1SA5.1 SA4 1SU5.1 SU3 1SU6.1 SU6 1,2SU7.1 SU4 2SU7.2 SU4 1SU8.1 SU5 1,2FGI.1 FG1 1FS1.1 FS1 1FA1.1 FA1 1FU1.1 FU1 1Page 19EPMP-EPP-0102Rev 00 (Draft A) 6.0 ATTACHMENTS6.16.2Attachment 1, Emergency Action Level Technical BasesAttachment 2, Fission Product Barrier Loss / Potential Loss Matrix and BasisPage 20 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory R -Abnormal Radiation Levels / Radiological EffluentsEAL Group: ANY (EALs in this category are applicable toany plant condition, hot or cold.)Many EALs are based on actual or potential degradation of fission product barriersbecause of the elevated potential for offsite radioactivity release. Degradation of fissionproduct barriers though is not always apparent via non-radiological symptoms. Therefore,direct indication of elevated radiological effluents or area radiation levels are appropriatesymptoms for emergency classification.At lower levels, abnormal radioactivity releases may be indicative of a failure ofcontainment systems or precursors to more significant releases. At higher release rates,offsite radiological conditions may result which require offsite protective actions. Elevatedarea radiation levels in plant may also be indicative of the failure of containment systemsor preclude access to plant vital equipment necessary to ensure plant safety.Events of this category pertain to the following subcategories:1. Offsite Rad ConditionsDirect indication of effluent radiation monitoring systems provides a rapid assessmentmechanism to determine releases in excess of classifiable limits. Projected offsitedoses, actual offsite field measurements or measured release rates via samplingindicate doses or dose rates above classifiable limits.2. Onsite Rad Conditions & Spent Fuel EventsSustained general area radiation levels in excess of those indicating loss of control ofradioactive materials or those levels which may preclude access to vital plant areasalso warrant emergency classification.3. CR/CAS RadSustained general area radiation levels which may preclude access to areas requiringcontinuous occupancy also warrant emergency classification.Page 21 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: Offsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity > 1,000 mRem TEDE or 5,000 mRem thyroidCDE for the actual or projected duration of the release using actualmeteorologyEAL:RGI1 General EmergencyANY monitor reading > Table R-1 "GE" column for > 15 min. (Note 1)* Do not delay declaration awaiting dose assessment results0 If dose assessment results are available, declaration should be based on doseassessment instead of radiation monitor values (see EAL RG1.2)Note 1: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition will likely exceed the applicable timeTable R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousRadwaste/RB Vent Effluent 5.5E+7 pCi/s 5.5E+6 pCi/s 200 x Alarm 2 x AlarmMain Stack Effluent 1.OE+10 pCi/s 1.0E+9 pCi/s 200 x Alarm 2 x AlarmLiquidService Water Effluent N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High(red)Cooling Tower Blowdown' N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Mode Applicability:AllBasis:Plant-SpecificThe DRAGON computer code has been used to determine the threshold values in TableR-1 for the GE classification level. The methodology develops an isotopic concentration inthe secondary containment that, when released through the Radwaste/RB Vent or theMain Stack, achieves 1,000 mRem TEDE or 5,000 mRem thyroid CDE at the SITEBOUNDARY. The nuclide inventory in the secondary containment was artificially createdby postulating a source term in secondary containment based on main steam designPage 22 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesisotopic distribution and adjusting the release rate from secondary containment until eitherthe whole body or child thyroid dose limit at the SITE BOUNDARY is reached. Thisisotopic distribution is not intended to specify a particular accident as the initiating event.Values have been calculated for the GEMs noble gas channel only since this is the readingthat is readily available to the operator. Realistic, accident atmospheric dispersion (X/Q)factors have been applied. (ref. 1)The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 2).Liquid effluent radiation monitors are not addressed in Table R-1 at the Site AreaEmergency and General Emergency levels because the dose assessment code used tocalculate these Table R-1 readings only considers a release through the Radwaste/RBVent or the Main Stack.A radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.Releases of this magnitude are associated with the failure of plant systems needed for theprotection of the public and likely involve fuel damage.The monitor list in Table R-1 includes effluent monitors on all potential release pathways.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementingprocedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when theclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP2 Basis Reference(s):1. Calculation PR-C-24-X2. NMP2 Offsite Dose Calculation Manual Figure D.1.0-13. NEI 99-01 IC AG1Page 23 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R- Abnormal Radiation Levels/ Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: Offsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity > 1,000 mRem TEDE or 5,000 mRem thyroidCDE for the actual or projected duration of the release using actualmeteorologyEAL:RGI.2 General EmergencyDose assessment using actual meteorology indicates doses > 1,000 mRem TEDE or5,000 mRem thyroid CDE at or beyond the SITE BOUNDARYMode Applicability:AllBasis:Plant-SpecificThe 1,000 mRem TEDE dose is set at 100% of the EPA PAG, while the 5,000 mRemthyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDEand thyroid CDE.Dose assessment is performed in accordance with EPIP-EPP-08 "Offsite DoseAssessment and PAR" (ref. 1)The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 2).GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.Releases of this magnitude are associated with the failure of plant systems needed for theprotection of the public and likely involve fuel damage.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementingprocedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when theclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP2 Basis Reference(s):Page 24 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases1. EPIP-EPP-08 Offsite Dose Assessment and PAR2. NMP2 Offsite Dose Calculation Manual Figure D.1.0-13. NEI 99-01 IC AG1Page 25EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical BasesR -Abnormal Radiation Levels / Radiological EffluentsCategory:Subcategory: 1 -Offsite Rad ConditionsInitiating Condition: Offsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity > 1,000 mRem TEDE or 5,000 mRem thyroidCDE for the actual or projected duration of the release using actualmeteorologyEAL:RGI.3 General EmergencyField survey results indicate closed window dose rates > 1,000 mRem/hr expected tocontinue for __ 60 min. at or beyond the SITE BOUNDARY (Note 1)ORAnalyses of field survey samples indicate thyroid CDE > 5,000 mRem for 1 hr of inhalationat or beyond the SITE BOUNDARY (Note 1)Note 1: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition will likely exceed the applicable timeMode Applicability:AllBasis:Plant-SpecificReal time field surveys and sample analysis is performed by offsite field monitoring teamsper EPIP-EPP-07, "Downwind Radiological Monitoring" (ref. 1) and assessed forradiological dose consequences per EPIP-EPP-08 "Offsite Dose Assessment and PAR"(ref. 2).The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 3).GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary.Releases of this magnitude are associated with the failure of plant systems needed for theprotection of the public and likely involve fuel damage.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted. For thisreason, emergency implementing procedures should call for the timely performance of doseassessments using actual meteorology and release information. If the results of these dosePage 26EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesassessments are available when the classification is made (e.g., initiated at a lower classificationlevel), the dose assessment results override the monitor reading EAL.NMP2 Basis Reference(s):1. EPIP-EPP-07 Downwind Radiological Monitoring2. EPIP-EPP-08 Offsite Dose Assessment and PAR3. NMP2 Offsite Dose Calculation Manual Figure D.1.0-14. NEI 99-01 IC AG1Page 27 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: Offsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity exceeds 100 mRem TEDE or 500 mRemthyroid CDE for the actual or projected duration of the releaseusing actual meteorologyEAL:RSI.1 Site Area EmergencyANY monitor reading > Table R-1 "SAE" column for > 15 min. (Note 1)" Do not delay declaration awaiting dose assessment results* If dose assessment results are available, declaration should be based on doseassessment instead of radiation monitor values (see EAL RS1.2)Note 1: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition will likely exceed the applicable timeTable R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousRadwaste/RB Vent Effluent 5.5E+7 pCi/s 5.5E+6 pCi/s 200 x Alarm 2 x AlarmMain Stack Effluent 1.OE+10 pCi/s 1.OE+9 pCi/s 200 x Alarm 2 x AlarmLiquidService Water Effluent N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High(red)Cooling Tower Blowdown N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Mode Applicability:AllBasis:Plant-SpecificThe DRAGON computer code has been used to determine the threshold values in TableR-1 for the SAE classification level. The methodology develops an isotopic concentrationin the secondary containment that, when released through the Radwaste/RB Vent or theMain Stack, achieves 100 mRem TEDE or 500 mRem thyroid CDE at the SITEBOUNDARY. The nuclide inventory in the secondary containment was artificially createdby postulating a source term in secondary containment based on main steam designPage 28 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesisotopic distribution and adjusting the release rate from secondary containment until eitherthe whole body or child thyroid dose limit at the SITE BOUNDARY is reached. Thisisotopic distribution is not intended to specify a particular accident as the initiating event.Values have been calculated for the GEMs noble gas channel only since this is the readingthat is readily available to the operator. Realistic, accident atmospheric dispersion (X/Q)factors have been applied. (ref. 1)The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 2).Liquid effluent radiation monitors are not addressed in Table R-1 at the Site AreaEmergency and General Emergency levels because the dose assessment code used tocalculate these Table R-1 readings only considers a release through the Radwaste/RBVent or the Main Stack.A radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude areassociated with the failure of plant systems needed for the protection of the public.The site specific monitor list in Table R-1 includes effluent monitors on all potential releasepathways.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementingprocedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when theclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP2 Basis Reference(s):1. Calculation PR-C-24-X2. NMP2 Offsite Dose Calculation Manual Figure D.1.0-13. NEI 99-01 IC AS1Page 29 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: Offsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity exceeds 100 mRem TEDE or 500 mRemthyroid CDE for the actual or projected duration of the releaseusing actual meteorologyEAL:RSI.2 Site Area EmergencyDose assessment using actual meteorology indicates doses > 100 mRem TEDE or500 mRem thyroid CDE at or beyond the SITE BOUNDARYMode Applicability:AllBasis:Plant-SpecificThe 100 mRem TEDE dose is set at 10% of the EPA PAG, while the 500 mRem thyroidCDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE andthyroid CDE.Dose assessment is performed in accordance with EPIP-EPP-08 "Offsite DoseAssessment and PAR" (ref. 1)The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 2).GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude areassociated with the failure of plant systems needed for the protection of the public.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementingprocedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when theclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP2 Basis Reference(s):1. EPIP-EPP-08 Offsite Dose Assessment and PARPage 30 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases2. NMP2 Offsite Dose Calculation. Manual Figure D.1.0-13. NEI 99-01 IC AS1Page 31EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:R -Abnormal Radiation Levels / Radiological Effluents1 -Offsite Rad ConditionsOffsite dose resulting from an actual or IMMINENT release ofgaseous radioactivity exceeds 100 mRem TEDE or 500 mRemthyroid CDE for the actual or projected duration of the releaseusing actual meteorologyEAL:RS1.3 Site Area EmergencyField survey results indicate closed window dose rates > 100 mRem/hr expected tocontinue for _ 60 min. at or beyond the SITE BOUNDARY (Note 1)ORAnalyses of field survey samples indicate thyroid CDE > 500 mRem for 1 hr of inhalation ator beyond the SITE BOUNDARY (Note 1)Note 1: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition will likely exceed the applicable timeMode Applicability:AllBasis:Plant-SpecificReal time field surveys and sample analysis is performed by offsite field monitoring teamsper EPIP-EPP-07, "Downwind Radiological Monitoring" (ref. 1) and assessed forradiological dose consequences per EPIP-EPP-08 "Offsite Dose Assessment and PAR"(ref. 2).The SITE BOUNDARY is the line beyond which the land is not owned, leased, norotherwise controlled by Constellation (ref. 3).GenericThis EAL addresses radioactivity releases that result in doses at or beyond the SITE BOUNDARYthat exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude areassociated with the failure of plant systems needed for the protection of the public.Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not,the results from these assessments may indicate that the classification is not warranted, or mayindicate that a higher classification is warranted. For this reason, emergency implementingprocedures should call for the timely performance of dose assessments using actual meteorologyand release information. If the results of these dose assessments are available when thePage 32EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesclassification is made (e.g., initiated at a lower classification level), the dose assessment resultsoverride the monitor reading EAL.NMP2 Basis Reference(s):1. EPIP-EPP-07 Downwind Radiological Monitoring2. EPIP-EPP-08 Offsite Dose Assessment and PAR3. NMP2 Offsite Dose Calculation Manual Figure D.1.0-14. NEI 99-01 IC AS1Page 33EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 200 times the ODCM for 15 minutes or longerEAL:RAI.1 AlertANY gaseous monitor reading > Table R-1 "Alert" column for - 15 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Table R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousRadwaste/RB Vent Effluent 5.5E+7 pCi/s 5.5E+6 pCi/s 200 x Alarm 2 x AlarmMain Stack Effluent 1.0E+10 pCi/s 1.0E+9 pCi/s 200 x Alarm 2 x AlarmLiquidService Water Effluent N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High(red)Cooling Tower Blowdown N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Mode Applicability: AllBasis:Plant-SpecificThe value shown for each monitor in Table R-1 is two hundred times the high (red) alarmsetpoint for the Digital Radiation Monitoring System (DRMS). The DRMS high (red) alarmsetpoints for the listed monitors are conservatively set to ensure ODCM radioactivityrelease limits are not exceeded (ref. 1). Instrumentation that may be used to assess thisEAL is listed below (ref. 2):* Radwaste/Reactor Building Vent Effluent Monitoring Systemmonitor: 2RMS-PNL180Crecorder: 2RMS-RR170/180Page 34 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesannunciator: 851248* Main Stack Effluent Monitoring Systemmonitor: 2RMS-PNL1 70Crecorder: 2RMS-RR170/180annunciator: 851256A radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses an actual or substantial potential decrease in the level of safety of the plant asindicated by a radiological release that exceeds regulatory commitments for an extended period oftime.Nuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 200 x DRMS high (red) multiples are specified only to distinguish between non-emergencyconditions. While these multiples obviously correspond to an off-site dose or dose rate, theemphasis in classifying these events is the degradation in the level of safety of the Releasesshould not be prorated or averaged. For example, a release exceeding 600x ODCM for 5 minutesdoes not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL is intended for sites that have established effluent monitoring on non-routine releasepathways for which a discharge permit would not normally be prepared.NMP2 Basis Reference(s):1. NMP2 Off-Site Dose Calculation Manual Sections D.3.1.1, D.3.2.1, D.3.3.1, D.3.3.22. N2-OP-79 Radiation Monitoring System3. NEI 99-01 IC AA1Page 35 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 200 times the ODCM for 15 minutes or longerEAL:RA1.2 AlertANY liquid monitor reading > Table R-1 "Alert" column for - 15 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Table R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousRadwaste/RB Vent Effluent 5.5E+7 pCi/s 5.5E+6 pCi/s 200 x Alarm 2 x AlarmMain Stack Effluent 1.0E+10 pCi/s 1.0E+9 pCi/s 200 x Alarm 2 x AlarmLiquidService Water Effluent N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High(red)Cooling Tower Blowdown N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Mode Applicability:AllBasis:Plant-SpecificThe value shown for each monitor in Table R-1 is two hundred times the high (red) alarmsetpoint for the Digital Radiation Monitoring System (DRMS). The DRMS high (red) alarmsetpoints for the listed monitors are conservatively set to ensure ODCM radioactivityrelease limits are not exceeded (ref. 1). Instrumentation that may be used to assess thisEAL is listed below (ref. 2):* Service Water Effluent Loop A/B Radiationmonitor: 2SWP*RE146A/BPage 36 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesrecorder: 2SWP*RR146A/Bannunciator: 851258Cooling Tower Blowdown Linemonitor: CWS-RE 157annunciator: 851258The designation "N/A" in Table R-1 indicates that the listed instrument range is insufficientto indicate the specified value and therefore no value is used.A radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses an actual or substantial potential decrease in the level of safety of the plant asindicated by a radiological release that exceeds regulatory commitments for an extended period oftime.Nuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 200 x DRMS high (red) multiples are specified only to distinguish between non-emergencyconditions. While these multiples obviously correspond to an off-site dose or dose rate, theemphasis in classifying these events is the degradation in the level of safety of the plant, not themagnitude of the associated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 600x ODCM for 5minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiationmonitor readings to exceed the threshold identified in the EAL established by the radioactivitydischarge permit. This value may be associated with a planned batch release, or a continuousrelease path.Page 37 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesNMP2 Basis Reference(s):1. NMP2 Off-Site Dose Calculation Manual Sections D.3.1.1, D.3.2.1, D.3.3.1, D.3.3.22. N2-OP-79 Radiation Monitoring System3. NEI 99-01 IC AA1Page 38EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 200 times the ODCM for 15 minutes or longerEAL:RA1.3 AlertConfirmed sample analyses for gaseous or liquid releases indicate concentrations orrelease rates > 200 x ODCM limits for > 15 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Mode Applicability:AllBasis:Plant-SpecificConfirmed sample analyses in excess of two hundred times the site Offsite DoseCalculation Manual (ODCM) limits that continue for 15 minutes or longer represent anuncontrolled situation and hence, a potential degradation in the level of safety. This eventescalates from the Unusual Event by raising the magnitude of the release by a factor of100 over the Unusual Event level (i.e., 200 times ODCM). Prorating the 500 mRem/yrbasis of the 10 CFR 20 non-occupational MPC limits for both time (8766 hr/yr) and the 200multiplier, the associated Exclusion Area Boundary dose rate would be approximately 10mRem/hr. If sample analysis indicates the threshold is met and nothing is done within 15minutes to effect a release reduction, the ED can conclude that the EAL threshold is metwithout second sample results.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses an actual or substantial potential decrease in the level of safety of the plant asindicated by a radiological release that exceeds regulatory commitments for an extended period oftime.Page 39 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesNuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 200 x ODCM limit are specified only to distinguish between non-emergency conditions. Whilethese multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifyingthese events is the degradation in the level of safety of the plant, not the magnitude of theassociated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 600x ODCM for5 minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses uncontrolled releases that are detected by sample analyses, particularly onunmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage.NMP2 Basis Reference(s):1. NMP2 Off-Site Dose Calculation Manual2. NEI 99-01 IC AA1Page 40EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:R -Abnormal Radiation Levels / Radiological Effluents1 -Offsite Rad ConditionsANY release of gaseous or liquid radioactivity to the environment> 2 times the ODCM for 60 minutes or longerEAL:RUI.1 Unusual EventANY gaseous monitor reading > Table R-1 "UE" column for > 60 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Table R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousRadwaste/RB Vent Effluent 5.5E+7 pCi/s 5.5E+6 pci/s 200 x Alarm 2 x AlarmMain Stack Effluent 1.0E+10 pCi/s 1.0E+9 pCi/s 200 x Alarm 2 x AlarmLiquidService Water Effluent N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High(red)Cooling Tower Blowdown N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Mode Applicability:AllBasis:Plant-SpecificThe value shown for each monitor in Table R-1 is two times the high (red) alarm setpointfor the Digital Radiation Monitoring System (DRMS). The DRMS high (red) alarm setpointsfor the listed monitors are conservatively set to ensure ODCM radioactivity release limitsare not exceeded (ref. 1). Instrumentation that may be used to assess this EAL is listedbelow (ref. 2):Radwaste/Reactor Building Vent Effluent Monitoring Systemmonitor: 2RMS-PNL180CPage 41EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesrecorder: 2RMS-RR170/180annunciator: 851248* Main Stack Effluent Monitoring Systemmonitor: 2RMS-PNL170Crecorder: 2RMS-RR1 70/180annunciator: 851256A radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses a potential decrease in the level of safety of the plant as indicated by aradiological release that exceeds regulatory commitments for an extended period of time.Nuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 2 x DRMS (red) multiples are specified only to distinguish between non-emergency conditions.While these multiples obviously correspond to an off-site dose or dose rate, the emphasis inclassifying these events is the degradation in the level of safety of the plant, not the magnitude ofthe associated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 4x ODCM for 30minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiationmonitor readings to exceed the threshold identified in the IC.This EAL is intended for sites that have established effluent monitoring on non-routine releasepathways for which a discharge permit would not normally be prepared.NMP2 Basis Reference(s):Page 42 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases1. NMP2 Off-Site Dose Calculation Manual Sections D.3.1.1, D.3.2.1, D.3.3.1, D.3.3.22. N2-OP-79 Radiation Monitoring System3. NEI 99-01 IC AU1Page 43EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 2 times the ODCM for 60 minutes or longerEAL:RU1.2 Unusual EventANY liquid monitor reading > Table R-1 "UE" column for > 60 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Table R-1 Effluent Monitor Classification ThresholdsMonitor GE SAE Alert UEGaseousRadwaste/RB Vent Effluent 5.5E+7 pCi/s 5.5E+6 pCi/s 200 x Alarm 2 x AlarmMain Stack Effluent 1.0E+10 pCi/s 1.0E+9 pCi/s 200 x Alarm 2 x AlarmLiquidService Water Effluent N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Liquid RadWaste Effluent N/A N/A N/A 2 x DRMS High(red)Cooling Tower Blowdown N/A N/A 200 x DRMS High(red) 2 x DRMS High(red)Mode Applicability:AllBasis:Plant-SpecificThe value shown for each monitor in Table R-1 is two times the high (red) alarm setpointfor the Digital Radiation Monitoring System (DRMS). The DRMS high (red) alarm setpointsfor the listed monitors are conservatively set to ensure ODCM radioactivity release limitsare not exceeded (ref. 1). Instrumentation that may be used to assess this EAL is listedbelow (ref. 2):* Service Water Effluent Loop A/B Radiationmonitor: 2SWP*RE146A/Brecorder: 2SWP*RR146A/BPage 44EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesannunciator: 851258* Liquid Effluent Linemonitor: LWS-RE206annunciator: 851258" Cooling Tower Blowdown Linemonitor: CWS-RE 157annunciator: 851258A radiation monitor reading is VALID when a release path is established. If the releasepath to the environment has been isolated, the radiation monitor reading is not VALID forclassification.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This IC addresses a potential decrease in the level of safety of the plant as indicated by aradiological release that exceeds regulatory commitments for an extended period of time.Nuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 2 x ODCM limit multiples are specified only to distinguish between non-emergency conditions.While these multiples obviously correspond to an off-site dose or dose rate, the emphasis inclassifying these events is the degradation in the level of safety of the plant, not the magnitude ofthe associated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 4x ODCM for 30minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiationmonitor readings to exceed the threshold identified in the EAL established by the radioactivitydischarge permit. This value may be associated with a planned batch release, or a continuousrelease path.NMP2 Basis Reference(s):Page 45 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases1. NMP2 Off-Site Dose Calculation Manual Sections D.3.1.1, D.3.2.1, D.3.3.1, D.3.3.22. N2-OP-79 Radiation Monitoring System3. NEI 99-01 IC AU1Page 46EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 1 -Offsite Rad ConditionsInitiating Condition: ANY release of gaseous or liquid radioactivity to the environment> 2 times the ODCM for 60 minutes or longerEAL:RU1.3 Unusual EventConfirmed sample analyses for gaseous or liquid releases indicate concentrations orrelease rates > 2 x ODCM limits for -> 60 min. (Note 2)Note 2: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the release duration has exceeded, or will likely exceed, the applicable time. In the absenceof data to the contrary, assume that the release duration has exceeded the applicable time if an ongoingrelease is detected and the release start time is unknown.Mode Applicability:AllBasis:Plant-SpecificReleases in excess of two times the site Offsite Dose Calculation Manual (ODCM) (ref. 1)instantaneous limits that continue for 60 minutes or longer represent an uncontrolledsituation and hence, a potential degradation in the level of safety. The final integrated dose(which is very low in the Unusual Event emergency class) is not ihe primary concern here;it is the degradation in plant control implied by the fact that the release was not isolatedwithin 60 minutes. Therefore, it is not intended that the release be averaged over 60minutes. For example, a release of 4 times the ODCM limit for 30 minutes does not exceedthis initiating condition. Further, the ED should not wait until 60 minutes has elapsed, butshould declare the event as soon as it is determined that the release duration has or willlikely exceed 60 minutes.GenericThe Emergency Director should not wait until the applicable time has elapsed, but should declarethe event as soon as it is determined that the condition will likely exceed the applicable time.This EAL addresses a potential decrease in the level of safety of the plant as indicated by aradiological release that exceeds regulatory commitments for an extended period of time.Page 47 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesNuclear power plants incorporate features intended to control the release of radioactive effluents tothe environment. Further, there are administrative controls established to prevent unintentionalreleases, or control and monitor intentional releases. The occurrence of extended, uncontrolledradioactive releases to the environment is indicative of a degradation in these features and/orcontrols.The 2 x ODCM limit multiples are specified only to distinguish between non-emergency conditions.While these multiples obviously correspond to an off-site dose or dose rate, the emphasis inclassifying these events is the degradation in the level of safety of the plant, not the magnitude ofthe associated dose or dose rate.Releases should not be prorated or averaged. For example, a release exceeding 4x ODCM for 30minutes does not meet the threshold.This EAL includes any release for which a radioactivity discharge permit was not prepared, or arelease that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarmsetpoints, etc.) on the applicable permit.This EAL addresses uncontrolled releases that are detected by sample analyses, particularly onunmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakagein water systems, etc.NMP2 Basis Reference(s):1. NMP2 Off-Site Dose Calculation Manual2. NEI 99-01 IC AU1Page 48 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:R -Abnormal Radiation Levels / Radiological Effluents2 -Onsite Rad Conditions & Spent Fuel EventsDamage to irradiated fuel or loss of water level that has resulted orwill result in the uncovering of irradiated fuel outside the ReactorVesselEAL:RA2.1 AlertAlarm on ANY of the following radiation monitors due to damage to irradiated fuel or lossof water level:* 2RMS-RE111* 2RMS-RE112* 2RMS-RE113" 2RMS-RE114* 2RMS-RE140" 2HVR*RE14A" 2HVR*RE14BMode Applicability:AllBasis:Plant-SpecificThis EAL is defined by the specific areas where irradiated fuel is located such as thereactor cavity, RPV or Spent Fuel Pool.The bases for the area radiation high alarms and the Above Refuel Floor HVAC Exhaust(2HVR*RE14A/B) high alarms are a spent fuel handling accident and are, therefore,appropriate for this EAL.Elevated readings on the ventilation monitors may also be indication of a radioactivityrelease from the fuel, confirming that damage has occurred. However, elevatedbackground at the monitor due to water level lowering may mask elevated ventilationexhaust airborne activity and needs to be considered.However, while radiation monitors may detect a rise in dose rate due to a drop in the waterlevel, it might not be a reliable indication of whether or not the fuel is covered. Forexample, the monitor could in fact be properly responding to a known event involvingPage 49EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basestransfer or relocation of a source stored in or near the Spent Fuel Pool or responding to aplanned evolution such as removal of the RPV head. Interpretation of these EALthresholds requires some understanding of the actual radiological conditions present in thevicinity of the monitors.GenericThis EAL addresses increases in radiation dose rates within plant buildings, and may be aprecursor to a radioactivity release to the environment. These events represent a loss of controlover radioactive material and represent an actual or substantial potential degradation in the level ofsafety of the plant.This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage.Increased ventilation monitor readings may be indication of a radioactivity release from the fuel,confirming that damage has occurred. Increased background at the ventilation monitor due towater level decrease may mask increased ventilation exhaust airborne activity and needs to beconsidered.While a radiation monitor could detect an increase in dose rate due to a drop in the water level, itmight not be a reliable indication of whether or not the fuel is covered.Escalation of this emergency classification level, if appropriate, would be based on RS1.1, RS1.2,RS1.3, RG1.1, RG1.2 or RG1.3.NMP2 Basis Reference(s):1. N2-SOP-39 Refuel Floor Events2. N2-ARP-01 Annunciator Response Procedures for annunciator 8512543. NEI 99-01 IC AA2Page 50EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 2 -Onsite Rad Conditions & Spent Fuel EventsInitiating Condition: Damage to irradiated fuel or loss of water level that has resulted orwill result in the uncovering of irradiated fuel outside the ReactorVesselEAL:RA2.2 AlertA water level drop in a reactor refueling pathway that will result in irradiated fuel becominguncoveredMode Applicability:AllBasis:Plant-SpecificThe reactor cavity and Spent Fuel Pool comprise the reactor refueling pathway (ref. 1).The movement of irradiated fuel assemblies requires a minimum water level of 22 ft 3 in.above the RPV flange and the top of spent fuel in the SFP. During refueling activities, thismaintains sufficient water level in the reactor cavity and SFP to retain iodine fissionproduct activity in the water in the event of a fuel handling accident (ref. 2, 3).Allowing level to decrease could result in spent fuel being uncovered, reducing spent fueldecay heat removal and creating an extremely hazardous radiation environment.There is no indication that water level in the spent fuel pool has dropped to the level of thefuel other than by visual observation by personnel on the refueling floor. N2-SOP-39,Refuel Floor Events, provides appropriate instructions to report a visual observation ofirradiated fuel uncovery (ref. 4).GenericThis event represents a loss of control over radioactive material and represents an actual orsubstantial potential degradation in the level of safety of the plant.Escalation of this emergency classification level, if appropriate, would be based on RS1.1, RS1.2,RSl.3, RG1.1, RG1.2 or RG1.3.NMP2 Basis Reference(s):Page 51 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases1.2.3.4.5.USAR Section 9.1.2Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.7.6Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.9.6N2-SOP-39, Refuel Floor EventsNEI 99-01 IC AA2Page 52 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:EAL:R -Abnormal Radiation Levels / Radiological Effluents2 -Onsite Rad Conditions & Spent Fuel EventsUNPLANNED rise in plant radiation levelsRU2.1 Unusual EventUNPLANNED water level drop in a reactor refueling pathway as indicated by inability torestore and maintain SFP level > low water level alarm (Note 3)ANDArea radiation monitor reading rise on ANY of the following:" 2RMS-RE111" 2RMS-RE112" 2RMS-RE113" 2RMS-RE114" 2RMS-RE140Note 3: If loss of water level in the refueling pathway occurs while in Mode 4, 5 or D, consider classification underEALs CU3.1, CU3.2 or CU3.3Mode Applicability:AllBasis:Plant-SpecificThe reactor cavity and Spent Fuel Pool (SFP) comprise the reactor refueling pathway (ref.1).The SFP is normally filled to a level of 352 ft 10 in. Level switches 2SFC*LS55A and B areset at 2 inches below the normal water level (or 352 ft 8 in.) and activate annunciators873317 and 875117 in the Control Room. (ref. 2, 3)The phrase "... inability to restore and maintain level >..." allows the operator to visuallyobserve the low water level condition, if possible, and to attempt water level restorationactions as long as water level remains above the top of irradiated fuel. Water levelrestoration operations are performed in accordance with N2-OP-38 (ref. 4).Technical Specifications requires that:* SFP water level be maintained 22 ft 3 in. above irradiated fuel seated in the storagePage 53 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesracks during movement of irradiated fuel assemblies in the SFP (ref. 5).* RPV water level be maintained 22 ft 3 in. above the top of the RPV flange duringmovement of irradiated fuel assemblies in the RPV (ref. 6).The listed Area radiation monitors are located in the proximity of where spent fuel may belocated and have been selected to be indicative of a decrease in radiation shielding due todecreasing refueling pathway water level (ref. 1). While a radiation monitor could detect arise in dose due to a drop in the water level, it might not be a reliable indication, in and ofitself, of whether or not the fuel is uncovered. For example, the reading on an arearadiation monitor located on the refuel bridge may rise due to planned evolutions such asRPV head lift or a fuel assembly being raised on fuel grapple. Elevated radiation monitorindications will need to be combined with another indicator (or personnel report) of waterloss.This event escalates to an Alert if irradiated fuel outside the RPV is uncovered.GenericThis EAL addresses increased radiation levels as a result of water level decreases aboveirradiated fuel or events that have resulted, or may result, in UNPLANNED increases in radiationdose rates within plant buildings. These radiation increases represent a loss of control overradioactive material and represent a potential degradation in the level of safety of the plant.The refueling pathway is a combination of cavities, tubes, canals and pools. While a radiationmonitor could detect an increase in dose rate due to a drop in the water level, it might not be areliable indication of whether or not the fuel is covered.For refueling events where the water level drops below the RPV flange classification would be viaEAL CU3.1, CU3.2 or CU3.3. This event escalates to an Alert per EAL RA2.1 if irradiated fueloutside the reactor vessel is uncovered. For events involving irradiated fuel in the reactor vessel,escalation would be via the Fission Product Barrier Table for events in operating modes 1-4.NMP2 Basis Reference(s):1. USAR Section 9.1.22. N2-ARP-01 Annunciator Response Procedures for annunciator 8733173. N2-ARP-01 Annunciator Response Procedures for annunciator 8751174. N2-OP-38 Spent Fuel Pool Cooling and Cleanup System5. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.7.66. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.9.67. N2-SOP-39 Refuel Floor Events8. NEI 99-01 IC AU2Page 54 EPMP-EPP-0102Rev 00 (Draft A)

Category:Subcategory:Attachment 1 -Emergency Action Level Technical BasesR -Radioactivity Release / Area Radiation2 -Onsite Rad Conditions & Spent Fuel EventsInitiating Condition: UNPLANNED rise in plant radiation levelsEAL:RU2.2 Unusual EventUNPLANNED area radiation readings rise by a factor of 1,000 over NORMAL LEVELSMode Applicability:AllBasis:Plant-SpecificAssessment of this EAL may be made with survey readings using portable instruments aswell as installed radiation monitors.GenericThis EAL addresses increased radiation levels as a result of water level decreases aboveirradiated fuel or events that have resulted, or may result, in UNPLANNED increases in radiationdose rates within plant buildings. These radiation increases represent a loss of control overradioactive material and represent a potential degradation in the level of safety of the plant.This EAL addresses increases in plant radiation levels that represent a loss of control ofradioactive material resulting in a potential degradation in the level of safety of the plant.This EAL excludes radiation level increases that result from planned activities such as use ofradiographic sources and movement of radioactive waste materials. A specific list of ARMs is notrequired as it would restrict the applicability of the threshold. The intent is to identify loss of controlof radioactive material in any monitored area.NMP2 Basis Reference(s):1. NEI 99-01 IC AU2Page 55EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: R -Abnormal Radiation Levels / Radiological EffluentsSubcategory: 3 -CR/CAS RadInitiating Condition: Rise in radiation levels within the facility that impedes operation ofsystems required to maintain plant safety functionsEAL:RA3.1 AlertDose rates > 15 mRem/hr in EITHER of the following areas requiring continuousoccupancy to maintain plant safety functions:Control RoomORCASMode Applicability:AllBasis:Plant-SpecificThe Control Room and Central Alarm Station (CAS) must be continuously occupied in allplant operating modes at NMP2. CAS is included in this EAL because of its importance topermitting access to areas required to assure safe plant operation.Area Radiation Monitor (ARM) 2RMS-RE129 monitors radiation levels in the Control Roomat 306' elevation. This is one of three Control Building ARMs that actuate Control Roomannunciator 851246, CONTROL BLDG AREA RADN MON ACTVATED, giving personnelsufficient warning of changing levels (ref. 1). There is no area radiation monitoring systemat NMP2 for the CAS. Abnormal radiation levels may be initially detected by routineradiological surveys.It is the impaired ability to operate the plant that results in the actual or potentialdegradation of the level of safety of the plant. The cause or magnitude of the increase inradiation levels is not a concern of this EAL. The Emergency Director must consider thesource or cause of the increased radiation levels and determine if any other EALs may beinvolved. For example, a dose rate of 15 mRem/hr in the Control Room may be a problemin itself. However, the increase may also be indicative of high dose rates in the primarycontainment due to a LOCA. In the latter case, a Site Area Emergency or a GeneralPage 56 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesEmergency may be indicated by other EAL categories.This EAL could result in declaration of an Alert at NMP2 due to a radioactivity release orradiation shine resulting from a major accident at the NMP1 or JAFNPP. Such adeclaration would be appropriate if the increase impairs safe plant operation.This EAL is not intended to apply to anticipated temporary radiation increases due toplanned events (e.g., radwaste container movement, depleted resin transfers, etc.).GenericThis EAL addresses increased radiation levels that: impact continued operation in areas requiringcontinuous occupancy to maintain safe operation or to perform a safe shutdown.The cause and/or magnitude of the increase in radiation levels is not a concern of this EAL. TheEmergency Director must consider the source or cause of the increased radiation levels anddetermine if any other EAL may be involved.Areas requiring continuous occupancy include the Control Room and any other control stationsthat are staffed continuously, such as the security alarm station CAS.NMP2 Basis Reference(s):1. N2-ARP-01 Annunciator Response Procedures for annunciator 8512462. NEI 99-01 IC AA3Page 57EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory H -Hazards and Other Conditions Affecting Plant SafetyEAL Group: ANY (EALs in this category are applicable toany plant condition, hot or cold.)Hazards are non-plant, system-related events that can directly or indirectly affect plantoperation, reactor plant safety or personnel safety.The events of this category pertain to the following subcategories:1. Natural or Destructive PhenomenaNatural events include hurricanes, earthquakes or tornados that have potential tocause plant structure or equipment damage of sufficient magnitude to threatenpersonnel or plant safety. Non-naturally occurring events that can cause damage toplant facilities and include aircraft crashes, missile impacts, etc.2. FIRE or EXPLOSIONFIREs can pose significant hazards to personnel and reactor safety. Appropriate forclassification are FIREs within the site PROTECTED AREA or which may affectoperability of equipment needed for safe shutdown3. Hazardous GasNon-naturally occurring events that can cause damage to plant facilities and includetoxic, asphyxiant, corrosive or flammable gas leaks.4. SecurityUnauthorized entry attempts into the PROTECTED AREA, BOMB threats, SABOTAGEattempts, and actual security compromises threatening loss of physical control of theplant.5. Control Room EvacuationEvents that are indicative of loss of Control Room habitability. If the Control Room mustbe evacuated, additional support for monitoring and controlling plant functions isnecessary through the emergency response facilities.6. JudQmentThe EALs defined in other categories specify the predetermined symptoms or eventsPage 58 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesthat are indicative of emergency or potential emergency conditions and thus warrantclassification. While these EALs have been developed to address the full spectrum ofpossible emergency conditions which may warrant classification and subsequentimplementation of the Emergency Plan, a provision for classification of emergenciesbased on operator/management experience and judgment is still necessary. The EALsof this category p rovide the Emergency Director the latitude to classify emergencyconditions consistent with the established classification criteria based upon EmergencyDirector judgment.Page 59EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant SafetyNatural or Destructive PhenomenaNatural or destructive phenomena affecting VITAL AREAsHAl.1 AlertSeismic event > OBE (0.075g)as indicated by EITHER:Computer Point ERSNC02, OBE DetectedORANY amber LED light lit at the Seismic Monitor Panel, Response SpectrumAnnunciatorANDEarthquake confirmed by ANY of the following:* Earthquake felt in plant" JAFNPP seismic instrumentation* Control Room indication of degraded performance of systems required for thesafe shutdown of the plantMode Applicability:AllBasis:Plant-SpecificThis EAL is based on the USAR design basis operating earthquake of 0.075g (ref. 1, 2).Seismic events of this magnitude can cause damage to plant safety functions.The method of detection relies on actuation of the NMP2 seismic monitor OBE alarmconfirmed by one or more indications such as shift operators on duty in the Control Roomdetermining that the ground motion was felt or degraded system performance.NMP2 seismic instrumentation actuates at 0.01g upon sensing any seismic activity (ref. 2).NMP1 and NMP2 share a common PROTECTED AREA border. Consideration should begiven to the opposite unit when classifying under this EAL.GenericThese EALs escalate from HU1.1 in that the occurrence of the event has resulted in VISIBLEPage 60EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical BasesDAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or hascaused damage to the safety systems in those structures evidenced by control room indications ofdegraded system response or performance. The occurrence of VISIBLE DAMAGE and/ordegraded system response is intended to discriminate against lesser events. The initial reportshould not be interpreted as mandating a lengthy damage assessment prior to classification. Noattempt is made in this EAL to assess the actual magnitude of the damage. The significance hereis not that a particular system or structure was damaged, but rather, that the event was of sufficientmagnitude to cause this degradation.Escalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.Seismic events of this magnitude can result in a VITAL AREA being subjected to forces beyonddesign limits, and thus damage may be assumed to have occurred to plant safety systems.NMP2 Basis Reference(s):1. USAR Section 3.7A.1.12. N2-SOP-90 Natural Events3. USAR Section 2.1.1.14. NEI 99-01 IC HA1Page 61 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety1 -Natural or Destructive PhenomenaNatural or destructive phenomena affecting VITAL AREAsHA1.2 AlertTornado strikingORSustained high winds > 90 mphresulting in EITHER:VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM orCOMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas" Reactor Building (including Primary Containment)" Control Room* Diesel Generator Engine and Board Rooms* Standby Switchgear and Battery Rooms" HPCS Switchgear and Battery Rooms* Remote Shutdown Rooms" Control Building HVAC Rooms* Service Water Pump Rooms* Electrical Protection Assembly Room" PGCC Relay RoomMode Applicability:AllBasis:Plant-SpecificAll Category 1 structures are designed for a wind velocity of 90 mph (ref. 1). This EAL isbased on the structural design basis of 90 mph or impact by tornado. Wind loads of thismagnitude can cause damage to safety functions.Page 62EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesWeather conditions are monitored at three locations:* The 200 foot high Primary OR Main Meteorological Tower located 0.6 miles west-southwest of NMP2" The 90 foot Backup Tower located east of JAFNPP" The 30 foot Inland Tower located at the Oswego County Airport near FultonMeteorological parameters such as wind speed are sent to the Control Rooms andTechnical Support Centers (TSC) at NMP1, NMP2, JAFNPP and the EmergencyOperations Facility (EOF). Data from sensors mounted on these towers are sent to bothdigital and analog systems for display, processing and storage. Wind speed and winddirection, as well as wind speed deviation and differential temperatures are monitored inNMP2 Control Room and recorded on strip chart recorders. (ref. 2)Wind speed can be measured up to 100 mph.Weather information may be obtained from (ref. 4):" National Weather Service: 716-565-9001 or 800-462-7751" Accu-Weather: 815-235-8650 or 814-237-5803The PROTECTED AREA Boundary is depicted in USAR Figure 1.2-1, Plot Plan (ref. 3).This threshold addresses events that may have resulted in a Safe Shutdown Area beingsubjected to forces beyond design limits and thus damage may be assumed to haveoccurred to plant safety systems. Safe Shutdown Areas are areas that house equipmentthe operation of which may be needed to ensure the reactor safely reaches and ismaintained in cold shutdown. Safe Shutdown Areas include structures that contain theequipment of concern. The Alert classification is appropriate if relevant plant parametersindicate that the performance of safety systems in the affected Safe Shutdown Areas hasbeen degraded. No attempt should be made to fully inventory the actual magnitude of thedamage or quantify the degradation of safety system performance prior to declaration ofan Alert under this threshold.Table H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 5).Page 63 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesNMP1 and NMP2 share a common PROTECTED AREA border. Consideration should begiven to the opposite unit when classifying under this EAL.GenericThis EAL escalates from HU1I.2 in that the occurrence of the event has resulted in VISIBLEDAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or hascaused damage to the safety systems in those structures evidenced by control room indications ofdegraded system response or performance. The occurrence of VISIBLE DAMAGE and/ordegraded system response is intended to discriminate against lesser events. The initial reportshould not be interpreted as mandating a lengthy damage assessment prior to classification. Noattempt is made in this EAL to assess the actual magnitude of the damage. The significance hereis not that a particular system or structure was damaged, but rather, that the event was of sufficientmagnitude to cause this degradation.Escalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.This EAL is based on a tornado striking (touching down) or high winds that have caused VISIBLEDAMAGE to structures containing functions or systems required for safe shutdown of the plant.NMP2 Basis Reference(s):1. USAR Section 3.3.1.12. N2-OP-102 Meteorological Monitoring3. USAR Figure 1.2-14. N2-SOP-64 High Winds5. USAR 9B and Figure 9B.6-16. NEI 99-01 IC HA1Page 64EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HA1.3 AlertInternal floodingresulting in EITHER:An electrical shock hazard that precludes access to operate or monitor ANY SAFETY-RELATED STRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas* Reactor Building (including Primary Containment)* Control Room* Diesel Generator Engine and Board Rooms" Standby Switchgear and Battery Rooms* HPCS Switchgear and Battery Rooms* Remote Shutdown Rooms* Control Building HVAC Rooms" Service Water Pump Rooms* Electrical Protection Assembly Room* PGCC Relay RoomMode Applicability:AllBasis:Plant-SpecificThis threshold addresses the affect of flooding caused by internal events such ascomponent failures, Circulating, Component Cooling or Service Water line ruptures,equipment misalignment, FIRE suppression system actuation, and outage activitymishaps.Table H-1 Safe Shutdown Areas include all structures containing Category I equipmentPage 65EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesand systems needed for safe shutdown (ref. 1).Uncontrolled internal flooding that has degraded safety-related equipment or created asafety hazard precluding access necessary for the safe operation or monitoring of safetyequipment warrants declaration of an Alert.GenericEscalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.This EAL addresses the effect of internal flooding caused by events such as component failures,equipment misalignment, or outage activity mishaps. It is based on the degraded performance ofsystems, or has created industrial safety hazards (e.g., electrical shock) that preclude necessaryaccess to operate or monitor safety equipment. The inability to access, operate or monitor safetyequipment represents an actual or substantial potential degradation of the level of safety of theplant.Flooding as used in this EAL describes a condition where water is entering the room faster thaninstalled equipment is capable of removal, resulting in a rise of water level within the room.Classification of this EAL should not be delayed while corrective actions are being taken to isolatethe water source.NMP2 Basis Reference(s):1. USAR 9B and Figure 9B.6-12. NEI 99-01 IC HA1Page 66 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HA1.4 AlertTurbine failure-generated PROJECTILEsresulting in EITHER:VISIBLE DAMAGE to or penetration of ANY SAFETY-RELATED STRUCTURE,SYSTEM or COMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas* Reactor Building (including Primary Containment)" Control Room* Diesel Generator Engine and Board Rooms* Standby Switchgear and Battery Rooms* HPCS Switchgear and Battery Rooms* Remote Shutdown Rooms" Control Building HVAC Rooms" Service Water Pump Rooms" Electrical Protection Assembly Room" PGCC Relay RoomMode Applicability:AllBasis:Plant-SpecificThe turbine generator stores large amounts of rotational kinetic energy in its rotor. In theunlikely event of a major mechanical failure, this energy may be transformed into bothrotational and translational energy of rotor fragments. These fragments may impact thesurrounding stationary parts. If the energy-absorbing capability of these stationary turbinegenerator parts is insufficient, external PROJECTILEs will be released. These ejectedPage 67 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesPROJECTILEs may impact various plant structures, including those housing safety relatedequipment.Table H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1).GenericThis EAL escalates from HU1.4 in that the occurrence of the event has resulted in VISIBLEDAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or hascaused damage to the safety systems in those structures evidenced by control room indications ofdegraded system response or performance. The occurrence of VISIBLE DAMAGE and/ordegraded system response is intended to discriminate against lesser events. The initial reportshould not be interpreted as mandating a lengthy damage assessment prior to classification. Noattempt is made in this EAL to assess the actual magnitude of the damage. The significance hereis not that a particular system or structure was damaged, but rather, that the event was of sufficientmagnitude to cause this degradation.Escalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.This EAL addresses the threat to safety related equipment imposed by PROJECTILEs generatedby main turbine rotating component failures. Therefore, this EAL is consistent with the definition ofan Alert in that the potential exists for actual or substantial potential degradation of the level ofsafety of the plant.NMP2 Basis Reference(s):1. USAR 9B and Figure 9B.6-12. NEI 99-01 IC HA1Page 68EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting VITAL AREAsEAL:HA1.5 AlertLake water level > 254 ftORIntake water level < 233 ftMode Applicability:AllBasis:Plant-SpecificThis threshold covers high and low water level conditions that may have resulted in a plantVITAL AREA being subjected to levels beyond design limits, and thus damage may beassumed to have occurred to plant safety systems.The high lake level is based upon the maximum probable flood level (ref. 1).The low forebay water level corresponds to the minimum intake bay water level whichprovides adequate submergence to the service water pumps (ref. 2, 3).GenlericIThisJEAL addresses other site specific phenomena that result in VISIBLE DAMAGE to VITALAREAs or results in indication of damage to SAFETY STRUCTURES, SYSTEMS, orCOMPONENTS containing functions and systems required for safe shutdown of the plant that canalso be precursors of more serious events.NMP2 Basis Reference(s):1. USAR Section 2.4.5.22. USAR Section 2.4.1.13. USAR Section 9.2.5.3.14. N2-OSP-LOG-WO01, Weekly Checks5. NEI 99-01 IC HA1Page 69 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety1 -Natural or Destructive PhenomenaNatural or destructive phenomena affecting VITAL AREAsHA1.6 AlertVehicle crashresulting in EITHER:VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM orCOMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas* Reactor Building (including Primary Containment)" Control Room" Diesel Generator Engine and Board Rooms" Standby Switchgear and Battery Rooms" HPCS Switchgear and Battery Rooms* Remote Shutdown Rooms" Control Building HVAC Rooms* Service Water Pump Rooms* Electrical Protection Assembly Room" PGCC Relay RoomMode Applicability:AllBasis:iPlaht-SpecificThis EAL is intended to address crashes of vehicle types large enough to cause significantdamage to plant structures containing functions and systems required for safe shutdown ofthe plant. Vehicle types include automobiles, aircraft, trucks, cranes, forklifts, waterbornecraft, etc.Table H-1 Safe Shutdown Areas include all structures containing Category I equipmentPage 70EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesand systems needed for safe shutdown (ref. 1).GenericThe occurrence of VISIBLE DAMAGE and/or degraded system response is intended todiscriminate against lesser events. The initial report should not be interpreted as mandating alengthy damage assessment prior to classification. No attempt is made in this EAL to assess theactual magnitude of the damage. The significance here is not that a particular system or structurewas ,damaged, but rather, that the event was of sufficient magnitude to cause this degradation.IEscalation of this emergency classification level, if appropriate, would be based on SystemMalfunction EALs.IThis, EAL addresses vehicle crashes within the PROTECTED AREA that results in VISIBLEDAMAGE to VITAL AREAs or indication of damage to SAFETY STRUCTURES, SYSTEMS, orCOMPONENTS containing functions and systems required for safe shutdown of the plant.NMP2 Basis Reference(s):1. USAR 9B and Figure 9B.6-12. NEI 99-01 IC HA1Page 71 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety1 -Natural or Destructive PhenomenaNatural or destructive phenomena affecting the PROTECTEDAREAEAL:HUI.1 Unusual EventSeismic event identified by ANY two of the following:* Annunciator 842121 SEISMIC ACCELERATION EXCEEDED indicates seismic eventdetected* Confirmation of earthquake received on NMP-1 or JAFNPP seismic instrumentation* Earthquake felt in plantMode Applicability:All'Basis:Plant-SpecificThe' NMP2 seismic instrumentation actuates at 0.01 g causing (ref. 1-4):* Power to remote acceleration sensor units* Activation of MRS1 recorders* EVENT alarm light on PWRS1 to light* EVENT INDICATOR on PWRS1 to turn from black to white* Annunciator 842121 on panel 2CEC-PNL842 to be receivedAnnunciator 842121 provides the most direct indication in the Control Room that a seismicIevent has occurred. The EVENT alarm light and EVENT INDICATOR are located on2CES-PNL889 in the relay room (ref. 4). Other methods are indication received from NMP-1 or JAFNPP instrumentation.Evaluation of the magnitude of the event will require evaluation of data recorded by theSeismic Monitoring Recorders.NMP1 and NMP2 share a common PROTECTED AREA border. Consideration should begiven to the opposite unit when classifying under this EAL.Page 72 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesGenericThis:EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be ofconcern to plant operators.Damage may be caused to some portions of the site, but should not affect ability of safetyfunctions to operate.As defined in the EPRI-sponsored Guidelines for Nuclear Plant Response to an Earthquake, datedOctober 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) thevibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based ona consensus of control room operators on duty at the time, and (b) for plants with operable seismicinstrumentation, the seismic switches of the plant are activated.NMP2 Basis Reference(s):1. USAR Section 3.72. Technical Requirements Manual Section 3.3.7.23. N2-OP-90 Seismic Monitor4. N,2-SOP-90 Natural Events5. USAR Section 2.1.1.16. NEI 99-01 IC HU1Page 73EPMP-EPP-0102Rev 00 (Draft A)

Attachment,1 -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting the PROTECTEDAREAEAL:HU1.2 Unusual EventTornado striking within PROTECTED AREA boundaryORSustained high winds > 90 mphMode Applicability:AllBasis:Plant-SpecificAll Category 1 safe shutdown structures are designed for a wind velocity of 90 mph, 30feet above ground using a gust factor of 1.1 (ref. 1).Weather conditions are monitored at three locations:* The 200 foot high Primary OR Main Meteorological Tower located 0.6 miles west-southwest of NMP2* The 90 foot Backup Tower located east of JAFNPP" The 30 foot Inland Tower located at the Oswego County Airport near FultonMeteorological parameters such as wind speed are sent to the Control Rooms andTechnical Support Centers (TSC) at NMP1, NMP2, JAFNPP and the EmergencyOperations Facility (EOF). Data from sensors mounted on these towers are sent to bothdigital and analog systems for display, processing and storage. Wind speed and winddirection, as well as wind speed deviation and differential temperatures are monitored inNMP2 Control Room and recorded on strip chart recorders. (ref. 2)Wind speed can be measured up to 100 mph.Weather information may be obtained from (ref. 3):Page 74 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases" National Weather Service: 716-565-9001 or 800-462-7751" Accu-Weather: 815-235-8650 or 814-237-5803NMP1 and NMP2 share a common PROTECTED AREA border. Consideration should begiven to the opposite unit when classifying under this EAL.GenericThis EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be ofconcern to plant operators.This EAL is based on a tornado striking (touching down) or high winds within the PROTECTEDAREA.Escalation of this emergency classification level, if appropriate, would be based on VISIBLEDAMAGE, or by other in plant conditions, via EAL HA1.2.NMP2 Basis Reference(s):1. USAR Section 3.3.1.12. N2-OP-102 Meteorological Monitoring3. N2-SOP-90 Natural Events4. NEI 99-01 IC HU1Page 75 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety1 -Natural or Destructive PhenomenaNatural or destructive phenomena affecting the PROTECTEDAREAEAL:HU1.3 Unusual EventInternal flooding that has the potential to affect ANY SAFETY-RELATED STRUCTURE,SYSTEM or COMPONENT required by Technical Specifications for the current operatingmode in ANY Table H-1 areaTable H-1 Safe Shutdown Areas* Reactor Building (including Primary Containment)* Control Room* Diesel Generator Engine and Board Rooms* Standby Switchgear and Battery Rooms* HPCS Switchgear and Battery Rooms* Remote Shutdown Rooms* Control Building HVAC Rooms* Service Water Pump Rooms* Electrical Protection Assembly Room* PGCC Relay RoomMode Applicability:AllBasis:Plant-SpecificThis threshold addresses the affect of flooding caused by internal events such ascomponent failures, Circulating, Component Cooling or Service Water line ruptures,equipment misalignment, FIRE suppression system actuation, and outage activitymishaps.Table H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1).Flooding as used in this EAL describes a condition where water is entering the room fasterPage 76EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesthan installed equipment is capable of removal, resulting in a rise of water level within theroom. Classification of this EAL should not be delayed while corrective actions are beingtaken to isolate the water source.GenericThis EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be ofconcern to plant operators.This EAL addresses the effect of internal flooding caused by events such as component failures,equipment misalignment, or outage activity mishaps.Escalation of this emergency classification level, if appropriate, would be based VISIBLE DAMAGEvia EAL HA1.3, or by other plant conditions.NMP2 Basis Reference(s):1. USAR 9B and Figure 9B.6-12. NEI 99-01 IC HU1Page 77EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting the PROTECTEDAREAEAL:HUI.4 Unusual EventTurbine failure resulting in ANY of the following:* Casing penetration" Damage to turbine seals" Damage to generator sealsMode Applicability:AllBasis:Plant-SpecificThe turbine generator stores large amounts of rotational kinetic energy in its rotor. In theunlikely event of a major mechanical failure, this energy may be transformed into bothrotational and translational energy of rotor fragments. These fragments may impact thesurrounding stationary parts. If the energy-absorbing capability of these stationary turbinegenerator parts is insufficient, external PROJECTILEs will be released. These ejectedPROJECTILEs may impact various plant structures, including those housing safety relatedequipment.In the event of PROJECTILE ejection, the probability of a strike on a plant region is afunction of the energy and direction of an ejected PROJECTILE and of the orientation ofthe turbine with respect to the plant region.Failure of turbine or generator seals may be indicated by a loss of seal oil pressure or lossof condenser vacuum (ref. 2, 3).GenericThese EALs are categorized on the basis of the occurrence of an event of sufficient magnitude tobe of concern to plant operators.This EAL addresses main turbine rotating component failures of sufficient magnitude to causePage 78 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesobservable damage to the turbine casing or to the seals of the turbine generator. Generator sealdamage observed after generator purge does not meet the intent of this EAL because it did notimpact normal operation of the plant.Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases(hydrogen cooling) to the plant environs. Actual FIRES and flammable gas build up areappropriately classified via EAL HU2.1 and EAL HU3.1.This EAL is consistent with the definition of a UE while maintaining the anticipatory nature desiredand recognizing the risk to non-safety related equipment.Escalation of this emergency classification level, if appropriate, would be to EAL HA1.4 based ondamage done by PROJECTILES generated by the failure or in conjunction with a steam generatortube rupture. These latter events would be classified by the Category R EALs or Category F EALs.NMP2 Basis Reference(s):1. N2-OP-21 Main Turbine System2. N2-SOP-21 Turbine Trip3. N2-ARP-01 Annunciator Response Procedures for annunciator 8511024. N2-ARP-01 Annunciator Response Procedures for annunciator 8511405. N2-SOP-09 Loss of Condenser Vacuum6. NEI 99-01 IC HU1Page 79EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 1 -Natural or Destructive PhenomenaInitiating Condition: Natural or destructive phenomena affecting the PROTECTEDAREAEAL:HUI.5 Unusual EventLake water level > 248.2 ftORIntake water level < 237 ftMode Applicability:AllBasis:Plant-SpecificThis threshold addresses high and low lake water level conditions that could be aprecursor of more serious events.The high lake level is based upon the maximum attainable uncontrolled lake water level asspecified in the USAR. Dams on the St. Lawrence River, under the authority of theInternational St. Lawrence River Board of Control, are now used to regulate the lake level.The low limit is set for el 74.37 m (244 ft) on April 1 and is maintained at or above thatelevation during the entire navigation season (April 1 to November 30). The upper limit ofthe lake level is el 75.59 m (248.2 ft) (ref. 1).The low level is based on intake water level and corresponds to the design minimum lakelevel. The probable minimum low water level of Lake Ontario at the site has beendetermined to be 72.0 m (236.3 ft) resulting from a setdown caused by a ProbableMaximum Wind Storm concurrent with the lowest probable lake level. (ref. 2)GenericThis EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be ofconcern to-plant operators.This EAL addresses other site specific phenomena that can also be precursors of more seriousevents.Page 80 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesNMP2 Basis Reference(s):1. USAR Section 2.4.1.22. USAR Section 2.4.11.23. N2-OSP-LOG-WO01, Weekly Checks4. NEI 99-01 IC HU1Page 81EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 2 -FIRE or EXPLOSIONInitiating Condition: FIRE or EXPLOSION affecting the operability of plant safetysystems required to establish or maintain safe shutdownEAL:HA2.1 AlertFIRE or EXPLOSIONresulting in EITHER:VISIBLE DAMAGE to ANY SAFETY-RELATED STRUCTURE, SYSTEM orCOMPONENT within ANY Table H-1 areaORControl Room indication of degraded performance of ANY SAFETY-RELATEDSTRUCTURE, SYSTEM or COMPONENT within ANY Table H-1 areaTable H-1 Safe Shutdown Areas* Reactor Building (including Primary Containment)* Control Room* Diesel Generator Engine and Board Rooms* Standby Switchgear and Battery Rooms* HPCS Switchgear and Battery Rooms* Remote Shutdown Rooms* Control Building HVAC Rooms* Service Water Pump Rooms* Electrical Protection Assembly Room* PGCC Relay RoomMode Applicability:AllBasis:Plant-SpecificTable H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1).GenericVISIBLE DAMAGE is used to identify the magnitude of the FIRE or EXPLOSION and todiscriminate against minor FIREs and EXPLOSIONs.Page 82 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesThe reference to structures containing safety systems or components is included to discriminateagainst FIREs or EXPLOSIONs in areas having a low probability of affecting safe operation. Thesignificance here is not that a safety system was degraded but the fact that the FIRE orEXPLOSION was large enough to cause damage to these systems.The use of VISIBLE DAMAGE should not be interpreted as mandating a lengthy damageassessment prior to classification. The declaration of an Alert and the activation of the TechnicalSupport Center will provide the Emergency Director with the resources needed to perform detaileddamage assessments.The Emergency Director also needs to consider any security aspects of the EXPLOSION.Escalation of this emergency classification level, if appropriate, will be based on EALs in CategoryS, Category F or Category R.NMP2 Basis Reference(s):1. USAR 9B and USAR Figure 9B.6-12. NEI 99-01 IC HA2Page 83EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety2 -FIRE or EXPLOSIONFIRE within the PROTECTED AREA not extinguished within 15min. of detection or EXPLOSION within the PROTECTED AREAEAL:HU2.1 Unusual EventFIRE not extinguished within 15 min. of Control Room notification or verification of aControl Room FIRE alarm in ANY Table H-1 area or Turbine Building (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table H-1 Safe Shutdown Areas" Reactor Building (including Primary Containment)* Control Room* Diesel Generator Engine and Board Rooms* Standby Switchgear and Battery Rooms.* HPCS Switchgear and Battery Rooms* Remote Shutdown Rooms* Control Building HVAC Rooms* Service Water Pump Rooms* Electrical Protection Assembly Room" PGCC Relay RoomMode Applicability:AllBasis:Plant-SpecificTable H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1). The Turbine Building is included becauseit is immediately adjacent to one or more Table H-1 areas and a FIRE within the TurbineBuilding may potentially impact safe shutdown equipment should the FIRE not becontrolled.GenericPage 84EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesThis EAL addresses the magnitude and extent of FIREs that may be potentially significantprecursors of damage to safety systems. It addresses the FIRE, and not the degradation inperformance of affected systems that may result.As used here, detection is visual observation and either report by plant personnel or sensor alarmindication.The 15 minute time period begins with a credible notification that a FIRE is occurring, or indicationof a FIRE detection system alarm/actuation. Verification of a FIRE detection systemalarm/actuation includes actions that can be taken within the control room or other nearby sitespecific location to ensure that it is not spurious. An alarm is assumed to be an indication of a FIREunless it is disproved within the 15 minute period by personnel dispatched to the scene. In otherwords, a personnel report from the scene may be used to disprove a sensor alarm if receivedwithin 15 minutes of the alarm, but shall not be required to verify the alarm.The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIREsthat are readily extinguished (e.g., smoldering waste paper basket).NMP2 Basis Reference(s):1. USAR 9B and Figure 9B.6-12. NEI 99-01 IC HU2Page 85EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 2 -FIRE or EXPLOSIONInitiating Condition: FIRE within the PROTECTED AREA not extinguished within 15min. of detection or EXPLOSION within the PROTECTED AREAEAL:HU2.2 Unusual EventEXPLOSION of sufficient force to damage permanent structures or equipment within thePROTECTED AREAMode Applicability:AllBasis:Plant-SpecificWhile some EXPLOSIONs may also result in FIREs that exceed EAL HU2.1, no FIRE isnecessary to declare an emergency in the event of an EXPLOSION. If a FIRE also occursas a result or with an EXPLOSION, declare the Unusual Event based on the EXPLOSIONand monitor the progress of the FIRE for potential escalation due to FIRE damage.NMP1 and NMP2 share a common PROTECTED AREA border. NMP1 and NMP2PROTECTED AREA boundaries are illustrated in USAR Figure 1.2-1 (ref. 1).GenericThis EAL addresses the magnitude and extent of EXPLOSIONs that may be potentially significantprecursors of damage to safety systems. It addresses the EXPLOSION, and not the degradation inperformance of affected systems that may result.This EAL addresses only those EXPLOSIONs of sufficient force to damage permanent structuresor equipment within the PROTECTED AREA.No attempt is made to assess the actual magnitude of the damage. The occurrence of theEXPLOSION is sufficient for declaration.The Emergency director also needs to consider any security aspects of the EXPLOSION, ifapplicable.Escalation of this emergency classification level, if appropriate, would be based on EAL HA2.1.NMP2 Basis Reference(s):1. USAR Figure 1.2-12. NEI 99-01 IC HU2Page 86 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety3- Hazardous GasAccess to a VITAL AREA is prohibited due to toxic, corrosive,asphyxiant or flammable gases which jeopardize operation ofoperable equipment required to maintain safe operations or safelyshutdown the reactorEAL:HA3.1 AlertAccess to ANY Table H-1 area is prohibited due to toxic, corrosive, asphyxiant orflammable gases which jeopardize operation of systems required to maintain safeoperations or safely shutdown the reactor (Note 5)Note 5: If the equipment in the stated area was already inoperable, or out of service, before the event occurred, thenEAL HA3.1 should not be declared as it will have no adverse impact on the ability of the plant to safelyoperate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.Table H-1 Safe Shutdown Areas* Reactor Building (including Primary Containment)* Control Room* Diesel Generator Engine and Board Rooms* Standby Switchgear and Battery Rooms* HPCS Switchgear and Battery Rooms* Remote Shutdown Rooms* Control Building HVAC Rooms* Service Water Pump Rooms* Electrical Protection Assembly Room* PGCC Relay RoomMode Applicability:AllBasis:Plant-SpecificTable H-1 Safe Shutdown Areas include all structures containing Category I equipmentand systems needed for safe shutdown (ref. 1)..For areas that contain no safety-related structure, system or component that wouldpotentially be required to be operated or for which the structure, system or component wasPage 87EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesalready out of service or inoperable before the event, this EAL would not be applicable.For purposes of this EAL, any gas (CO2 included) is considered toxic when oxygenconcentrations in the affected areas have been or could be expected to be reduced to<19.5% or toxicity of the gas will be injurious to persons inhaling it. For discharges ofHalon, NMP's systems are designed for discharge concentration from 5% up to 6.5%. Inaccordartme with NFPA 12 A, Halon 1301 Fire Extinguishing Systems, exposures to levelsof up to 7% produce little if any noticeable effect (ref. 2).GenericGases in a Safe Shutdown AREA can affect the ability to safely operate or safely shutdown thereactor.The fact that SCBA may be worn does not eliminate the need to declare the event.Declaration should not be delayed for confirmation from atmospheric testing if the atmosphereposes an immediate threat to life and health or an immediate threat of severe exposure to gases.This could be based upon documented analysis, indication of personal ill effects from exposure, oroperating experience with the hazards.If the equipment in the stated area was already inoperable, or out of service, before the eventoccurred, then this EAL should not be declared as it will have no adverse impact on the ability ofthe plant to safely operate or safely shutdown beyond that already allowed by TechnicalSpecifications at the time of the event.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.Most commonly, asphyxiants work by merely displacing air in an enclosed environment. Thisreduces the concentration of oxygen below the normal level of around 19%, which can lead tobreathing difficulties, unconsciousness or even death.An uncontrolled release of flammable gasses within a facility structure has the potential to affectsafe operation of the plant by limiting either operator or equipment operations due to the potentialfor ignition and resulting equipment damage/personnel injury. Flammable gasses, such ashydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repairequipment/components (acetylene -used in. welding). This EAL assumes concentrations offlammable gasses which can ignite/support combustion.Escalation of this emergency classification level, if appropriate, will be based on EALs in CategoryS, Category F or Category R.NMP2 Basis Reference(s):1. USAR 9B and Figure 9B.6-12. NFPA 12 A Halon 1301 Fire Extinguishing Systems3. NEI 99-01 IC HA3Page 88 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Hazardous GasInitiating Condition: Release of toxic, corrosive, asphyxiant or flammable gasesdeemed detrimental to NORMAL PLANT OPERATIONSEAL:HU3.1 Unusual EventToxic, corrosive, asphyxiant or flammable gases in amounts that have or could adverselyaffect NORMAL PLANT OPERATIONSMode Applicability:AllBasis:Plant-SpecificNORMAL PLANT OPERATIONS is defined to mean activities at the plant site associatedwith routine testing, maintenance, or equipment operations, in accordance with normaloperating or administrative procedures. Entry into abnormal or emergency operatingprocedures, or deviation from normal security or radiological controls posture, is adeparture from NORMAL PLANT OPERATIONS.For purposes of this EAL, any gas (C02 included) is considered toxic when oxygenconcentrations in the affected areas have been or could be expected to be reduced to<19.5% or toxicity of the gas will be injurious to persons inhaling it. For discharges ofHalon, NMP's systems are designed for discharge concentration from 5% up to 6.5%. Inaccordance with NFPA 12 A, Halon 1301 Fire Extinguishing Systems, exposures to levelsof up to 7% produce little if any noticeable effect (ref. 1).GenericThis EAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficientquantity to affect NORMAL PLANT OPERATIONS.The fact that SCBA may be worn does not eliminate the need to declare the event.This EAL is not intended to require significant assessment or quantification. It assumes anuncontrolled process that has the potential to affect plant operations. This would preclude small orincidental releases, or releases that do not impact structures needed for plant operation.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.Page 89 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesMost commonly, asphyxiants work by merely displacing air in an enclosed environment. Thisreduces the concentration of oxygen below the normal level of around 19%, which can lead tobreathing difficulties, unconsciousness or even death.Escalation of this emergency classification level, if appropriate, would be based on EAL HA3.1.NMP2 Basis Reference(s):1. NFPA 12 A Halon 1301 Fire Extinguishing Systems2. NEI 99-01 IC HU3Page 90EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 3 -Hazardous GasInitiating Condition: Release of toxic, corrosive, asphyxiant or flammable gasesdeemed detrimental to NORMAL PLANT OPERATIONSEAL:HU3.2 Unusual EventRecommendation by local, county or state officials to evacuate or shelter site personnelbased on an offsite eventMode Applicability:AllBasis:Plant-SpecificA recommendation by offsite officials that a potential evacuation of site personnel may berequired based on an offsite event assumes that the plant lies within an evacuation areaestablished by offsite officials due to a release of toxic, corrosive, asphyxiant or flammablegas. In this case, it can be assumed that an actual or potential release of such hazardousgas is anticipated to enter the PROTECTED AREA in amounts that could affect the healthof plant personnel or NORMAL PLANT OPERATIONS.GenericEscalation of this emergency classification level, if appropriate, would be based on EAL HA3.1.NMP2 Basis Reference(s):1. NEI 99-01 IC HU3Page 91 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:H -Hazards and Other Conditions Affecting Plant Safety4 -SecurityHOSTILE ACTION resulting in loss of physical control of thefacilityEAL:HG4.1 General EmergencyA HOSTILE ACTION has occurred such that plant personnel are unable to operateequipment required to maintain safety functionsMode Applicability:AllBasis:Plant-SpecificSafety functions include:* Reactivity control -ability to shut down the reactor and keep it shutdown" RPV level control -ability to cool the core* Decay heat removal -ability to maintain a heat sinkGenericThis EAL encompasses conditions under which a HOSTILE ACTION has resulted in a loss ofphysical control of VITAL AREAs (containing vital equipment or controls of vital equipment)required to maintain safety functions and control of that equipment cannot be transferred to andoperated from another location.If control of the plant equipment necessary to maintain safety functions can be transferred toanother location, then the threshold is not met.NMP2 Basis Reference(s):1. NEI 99-01 IC HG1Page 92 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 4 -SecurityInitiating Condition: HOSTILE ACTION resulting in loss of physical control of the facilityEAL:HG4.2 General EmergencyA HOSTILE ACTION has caused failure of Spent Fuel Cooling systemsANDIMMINENT fuel damage is likelyMode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses failure of spent fuel cooling systems as a result of HOSTILE ACTION ifIMMINENT fuel damage is likely.NMP2 Basis Reference(s):1. NEI 99-01 IC HG1Page 93EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesH -Hazards and Other Conditions Affecting Plant Safety4 -SecurityCategory:Subcategory:Initiating Condition: HOSTILE ACTION within the PROTECTED AREAEAL:HS4.1 Site Area EmergencyA HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA asreported by the Security Site SupervisorMode Applicability:AllBasis:GenericThis condition represents an escalated threat to plant safety above that contained in the Alert inthat a HOSTILE FORCE has progressed from the Owner Controlled Area to the PROTECTEDAREA.This EAL addresses the contingency for a very rapid progression of events, such as thatexperienced on September 11, 2001. It is not premised solely on the potential for a radiologicalrelease. Rather the issue includes the need for rapid assistance due to the possibility for significantand indeterminate damage from additional air, land or water attack elements.The fact that the site is under serious attack with minimal time available for further, preparation oradditional assistance to arrive requires Offsite Response Organization (ORO) readiness andpreparation for the implementation of protective measures.This EAL addresses the potential for a very rapid progression of events due to a HOSTILEACTION. It is not intended to address incidents that are accidental events or acts of civildisobedience, such as small aircraft impact, hunters, or physical disputes between employeeswithin the PROTECTED AREA. Those events are adequately addressed by other EALs.Escalation of this emergency classification level, if appropriate, would be based on actual plantstatus after impact or progression of attack.NMP2 Basis Reference(s):1. NMP Site Security Plan2. NEI 99-01 IC HS4Page 94EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesH -Hazards and Other Conditions Affecting Plant Safety SecurityCategory:Subcategory:Initiating Condition: HOSTILE ACTION within the Owner Controlled Area or airborneattack threatEAL:HA4.1 AlertA HOSTILE ACTION is occurring or has occurred within the Owner Controlled Area asreported by the Security Site SupervisorORA validated notification from NRC of an AIRLINER attack threat within 30 min. of the siteMode Applicability:AllBasis:Plant-SpecificNoneGenericNote: Timely and accurate communication between the Security Site Supervisor and the ControlRoom is crucial for the implementation of effective Security EALs.This EAL addresses the contingency for a very rapid progression of events, such as thatexperienced on September 11, 2001. They are not premised solely on the potential for aradiological release. Rather the issue includes the need for rapid assistance due to the possibilityfor significant and indeterminate damage from additional air, land or water attack elements.The fact that the site is under serious attack or is an identified attack target with minimal timeavailable for further preparation or additional assistance to arrive requires a heightened state ofreadiness and implementation of protective measures that can be effective (such as on-siteevacuation, dispersal or sheltering).First ConditionThis condition addresses the potential for a very rapid progression of events due to a HOSTILEACTION. It is not intended to address incidents that are accidental events or acts of civildisobedience, such as small aircraft impact, hunters, or physical disputes between employeeswithin the Owner Controlled Area. Those events are adequately addressed by other EALs.Note that this condition is applicable for any HOSTILE ACTION occurring, or that has occurred, inthe Owner Controlled Area.Second ConditionPage 95EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesThis condition addresses the immediacy of an expected threat arrival or impact on the site within arelatively short time.The intent of this condition is to ensure that notifications for the AIRLINER attack threat are madein a timely manner and that Offsite Response Organizations (OROs) and plant personnel are at astate of heightened awareness regarding the credible threat. AIRLINER is meant to be a LARGEAIRCRAFT with the potential for causing significant damage to the plant.This condition is met when a plant receives information regarding an AIRLINER attack threat fromNRC and the AIRLINER is within 30 minutes of the plant. Only the plant to which the specific threatis made need declare the Alert.The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threatinvolves an AIRLINER (AIRLINER is meant to be a LARGE AIRCRAFT with the potential forcausing significant damage to the plant). The status and size of the plane may be provided byNORAD through the NRC.NMP2 Basis Reference(s):1. NMP Site Security Plan2. NEI 99-01 IC HA4Page 96 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesH -Hazards and Other Conditions Affecting Plant Safety4 -SecurityCategory:Subcategory:Initiating Condition: Confirmed SECURITY CONDITION or threat which indicates apotential degradation in the level of safety of the plantEAL:HU4.1 Unusual EventA SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by theSecurity Site SupervisorORA credible site-specific security threat notificationORA validated notification from NRC providing information of an aircraft threatMode Applicability:AllBasis:Plant-SpecificIf the Security Site Supervisor determines that a threat notification is credible, the SecuritySite Supervisor will notify the Operations Shift Manager that a "Credible Threat" conditionexists for NMP2. Generally, NMP2 Security Procedures address standard practices fordetermining credibility. The three main criteria for determining credibility are: technicalfeasibility, operational feasibility, and resolve. For NMP2, a validated notification deliveredby the FBI, the NRC or similar agency is treated as credible.GenericNote: Timely and accurate communication between Security the Site Supervisor and the ControlRoom is crucial for the implementation of effective Security EALs.Security events which do not represent a potential degradation in the level of safety of the plant arereported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. Security events assessed asHOSTILE ACTIONs are classifiable under EAL HA4.1, EAL HS4.1 and EAL HG4.1.A higher initial classification could be made based upon the nature and timing of the security threatand potential consequences. The licensee shall consider upgrading the emergency responsestatus and emergency classification level in accordance with the NMP Site Security and Plan.First ConditionPage 97EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesReference is made to security shift supervision because these individuals are the designatedpersonnel on-site qualified and trained to confirm that a security event is occurring or has occurred.Training on security event classification confirmation is closely controlled due to the strict secrecycontrols placed on the NMP Site Security Plan.This threshold is based on the NMP Site Security Plan. The NMP Site Security Plan is based onguidance provided by NEI 03-12.Second ConditionThis threshold is included to ensure that appropriate notifications for the security threat are made ina timely manner. This includes information of a credible threat. Only the plant to which the specificthreat is made need declare the Unusual Event.The determination of "credible" is made through use of information found in the NMP Site SecurityPlan.Third ConditionThe intent of this EAL is to ensure that notifications for the aircraft threat are made in a timelymanner and that Offsite Response Organizations and plant personnel are at a state of heightenedawareness regarding the credible threat. It is not the intent of this EAL to replace existing non-hostile related EALs involving aircraft.This EAL is met when a plant receives information regarding an aircraft threat from NRC.Validation is performed by calling the NRC or by other approved methods of authentication. Onlythe plant to which the specific threat is made need declare the Unusual Event.The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threatinvolves an AIRLINER (AIRLINER is meant to be a LARGE AIRCRAFT with the potential forcausing significant damage to the plant). The status and size of the plane may be provided byNORAD through the NRC.Escalation to Alert emergency classification level via EAL HA4.1 would be appropriate if the threatinvolves an AIRLINER within 30 minutes of the plant.NMP2 Basis Reference(s):1. NMP Site Security Plan2. NEI 99-01 IC HU4Page 98 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesH -Hazards and Other Conditions Affecting Plant Safety5 -Control Room EvacuationCategory:Subcategory:Initiating Condition: Control Room evacuation has been initiated and plant controlcannot be establishedEAL:HS5.1 Site Area EmergencyControl Room evacuation has been initiatedANDControl of the plant cannot be established within 15 min.Mode Applicability:AllBasis:Plant-SpecificN2-SOP-78, Control Room Evacuation, provides specific instructions for evacuating theControl Room/Building and establishing plant control in alternate locations.GenericThe intent of this EAL is to capture those events where control of the plant cannot be reestablishedin a timely manner. In this case, expeditious transfer of control of safety systems has not occurred(although fission product barrier damage may not yet be indicated).The intent of the EAL is to establish control of important plant equipment and knowledge ofimportant plant parameters in a timely manner. Primary emphasis should be placed on thosecomponents and instruments that supply protection for and information about safety functions.Typically, these safety functions are reactivity control (ability to reach and maintain reactorshutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintaina heat sink).The determination of whether or not control is established at the remote shutdown panel is basedon Emergency Director (ED) judgment. The Emergency Director is expected to make a reasonable,informed judgment within the site specific time for transfer that the licensee has control of the plantfrom the remote shutdown panel.Escalation of this emergency classification level, if appropriate, would be by EALs in Category F orCategory R.NMP2 Basis Reference(s):1. N2-SOP-78 Control Room Evacuation2. USAR Section 9B.8.2.2Page 99EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases3. NEI 99-01 IC HS2Page 100EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: H -Hazards and Other Conditions Affecting Plant SafetySubcategory: 5 -Control Room EvacuationInitiating Condition: Control Room evacuation has been initiatedEAL:HA5.1 AlertControl Room evacuation has been initiatedMode Applicability:AllBasis:Plant-SpecificN2-SOP-78, Control Room Evacuation, provides specific instructions for evacuating theControl Room/Building and establishing plant control in alternate locations.GenericWith the control room evacuated, additional support, monitoring and direction through theTechnical Support Center and/or other emergency response facilities may be necessary.Inability to establish plant control from outside the control room will escalate this event to a SiteArea Emergency.NMP2 Basis Reference(s):1. N2-SOP-78 Control Room Evacuation2. USAR Section 9B.8.2.23. NEI 99-01 IC HA5Page 101 EPMP-EPP-0102Rev 00 (Draft A)

Category:Subcategory:Attachment 1 -Emergency Action Level Technical BasesH -Hazards and Other Conditions Affecting Plant Safety6 -JudgmentInitiating Condition: Other conditions exist that in the judgment of the EmergencyDirector warrant declaration of a General EmergencyEAL:HG6.1 General EmergencyOther conditions exist which in the judgment of the Emergency Director indicate thatevents are in progress or have occurred which involve actual or IMMINENT substantialcore degradation or melting with potential for loss of containment integrity or HOSTILEACTION that results in an actual loss of physical control of the facility. Releases can bereasonably expected to exceed EPA Protective Action Guideline exposure levels (1,000mRem TEDE or 5,000 mRem thyroid CDE) offsite for more than the immediate site areaMode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDirector to fall under the emergency classification level description for General Emergency.NMP2 Basis Reference(s):1. NEI 99-01 IC HG2Page 102EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesH -Hazards and Other Conditions Affecting Plant Safety6 -JudgmentCategory:Subcategory:Initiating Condition: Other conditions existing that in the judgment of the EmergencyDirector warrant declaration of a Site Area EmergencyEAL:HS6.1 Site Area EmergencyOther conditions exist which in the judgment of the Emergency Director indicate thatevents are in progress or have occurred which involve actual or likely major failures ofplant functions needed for protection of the public or HOSTILE ACTION that results inintentional damage or malicious acts; (1) toward site personnel or equipment that couldlead to the likely failure of or; (2) that prevent effective access to equipment needed for theprotection of the public. ANY releases are not expected to result in exposure levels whichexceed EPA Protective Action Guideline exposure levels (1,000 mRem TEDE or 5,000mRem thyroid CDE) beyond the SITE BOUNDARYMode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDirector to fall under the emergency classification level description for Site Area Emergency.NMP2 Basis Reference(s):1. NEI 99-01 IC HS3Page 103EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:EAL:H -Hazards and Other Conditions Affecting Plant Safety6 -JudgmentOther conditions exist that in the judgment of the EmergencyDirector warrant declaration of an AlertHA6.1 AlertOther conditions exist which in the judgment of the Emergency Director indicate thatevents are in progress or have occurred which involve an actual or potential substantialdegradation of the level of safety of the plant or a security event that involves probable lifethreatening risk to site personnel or damage to site equipment because of HOSTILEACTION. ANY releases are expected to be limited to small fractions of the EPA ProtectiveAction Guideline exposure levels (1,000 mRem TEDE or 5,000 mRem thyroid CDE)Mode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDirector to fall under the Alert emergency classification level.NMP2 Basis Reference(s):1. NEI 99-01 IC HA6Page 104EPMP-EPP-01 02Rev 00 (Draft A)

Category:Subcategory:Attachment 1 -Emergency Action Level Technical BasesH -Hazards and Other Conditions Affecting Plant Safety6 -JudgmentInitiating Condition: Other conditions existing that in the judgment of the EmergencyDirector warrant declaration of a UEEAL:HU6.1 Unusual EventOther conditions exist which in the judgment of the Emergency Director indicate thatevents are in progress or have occurred which indicate a potential degradation of the levelof safety of the plant or indicate a security threat to facility protection has been initiated. Noreleases of radioactive material requiring offsite response or monitoring are expectedunless further degradation of safety systems occursMode Applicability:AllBasis:Plant-SpecificNoneGenericThis EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the EmergencyDirector to fall under the UE emergency classification level.NMP2 Basis Reference(s):1. NEI 99-01 IC HU5Page 105EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory E -INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)EAL Group: Not Applicable (the EAL in this category isapplicable independent of plant operatingmode)An INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) is a complex that isdesigned and constructed for the interim storage of spent nuclear fuel and otherradioactive materials associated with spent fuel storage. A significant amount of theradioactive material contained within a cask/canister must escape its packaging and enterthe biosphere for there to be a significant environmental effect resulting from an accidentinvolving the dry storage of spent nuclear fuel. Formal offsite planning is not requiredbecause the postulated worst-case accident involving an ISFSI has insignificantconsequences to the public health and safety.A Notification of Unusual Event is declared on the basis of the occurrence of an event ofsufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged orviolated. This includes classification based on a loaded fuel storage cask/canisterCONFINEMENT BOUNDARY loss leading to the degradation of the fuel during storage orposing an operational safety problem with respect to its removal from storage.A hostile security event that leads to a potential loss in the level of safety of the ISFSI is aclassifiable event under Security category EAL HA4.1.Minor surface damage that does not affect storage cask/canister boundary is excludedfrom the scope of these EALs.Page 106 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: E -ISFSISubcategory: Not ApplicableInitiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARYEAL:EUI.1 Unusual EventDamage to a loaded cask CONFINEMENT BOUNDARY as indicated by measured doserates > then ANY of the following:0 400 mRem/hr at 3 feet from the HSM surface0 100 mRem/hr outside HSM door on centerline* 20 mRem/hr end shield wall exteriorMode Applicability:AllBasis:Plant-SpecificThe NMP site ISFSI utilizes the NUHOMS Horizontal Modular Storage System.This EAL addresses any condition which indicates a loss of a cask CONFINEMENTBOUNDARY and thus a potential degradation in the level of safety of the ISFSI. The caskCONFINEMENT BOUNDARY is the NUHOMS 61BT Dry Shielded Canister (DSC). TheDSC is the pressure-retaining component of the storage system (ref. 1). Each loaded DSCis housed within a Horizontal Storage Module (HSM). Indication of a loss ofCONFINEMENT BOUNDARY is any increase in external HSM radiation levels in excess ofTechnical Specification limits (ref. 2).GenericAn UE in this EAL is categorized on the basis of the occurrence of an event of sufficient magnitudethat a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includesclassification based on a loaded fuel storage cask CONFINEMENT BOUNDARY loss leading tothe degradation of the fuel during storage or posing an operational safety problem with respect toits removal from storage.NMP2 Basis Reference(s):1. CDP No. Ni-07-092/N2-07-070 Nine Mile Point Nuclear Station -Conceptual Design,Independent Spent Fuel Storage InstallationPage 107 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases2. Transnuclear, Inc. Standardized NUHOMS Horizontal Modular Storage SystemCertificate of Compliance No. 1004, Attachment A Technical Specifications Section1.2.7 HSM Dose Rates with a Loaded 24P, 52B or 61 BT DSC3. NEI 99-01 IC E-HU1Page 108EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory C -Cold Shutdown / Refueling System MalfunctionEAL Group: Cold Conditions (RCS temperature -2000F);EALs in this category are applicable only inone or more cold operating modes.Category C EALs are directly associated with cold shutdown or refueling system safetyfunctions. Given the variability of plant configurations (e.g., systems out-of-service formaintenance, containment open, reduced AC power redundancy, time since shutdown)during these periods, the consequences of any given initiating event can vary greatly. Forexample, a loss of decay heat removal capability that occurs at the end of an extendedoutage has less significance than a similar loss occurring during the first week aftershutdown. Compounding these events is the likelihood that instrumentation necessary forassessment may also be inoperable. The cold shutdown and refueling system malfunctionEALs are based on performance capability to the extent possible with consideration givento RCS integrity, containment closure, and fuel clad integrity for the applicable operatingmodes (4 -Cold Shutdown, 5 -Refuel, D -Defueled).The events of this category pertain to the following subcategories:1. Loss of AC PowerLoss of emergency plant electrical power can compromise plant safety systemoperability including decay heat removal and emergency core cooling systems whichmay be necessary to ensure fission product barrier integrity. This category includesloss of onsite and offsite power sources for the 4.16 KV emergency buses.2. Loss of DC PowerLoss of emergency plant electrical power can compromise plant safety systemoperability including decay heat removal and emergency core cooling systems whichmay be necessary to ensure fission product barrier integrity. This category includesloss of power to the 125 VDC buses.Page 109 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases3. RPV LevelRPV water level is a measure of inventory available to ensure adequate core coolingand, therefore, maintain fuel clad integrity. The RPV provides a volume for the coolantthat covers the reactor core. The RPV and associated pressure piping (reactor coolantsystem) together provide a barrier to limit the release of radioactive material should thereactor fuel clad integrity fail.4. RCS TemperatureUncontrolled or inadvertent temperature or pressure increases are indicative of apotential loss of safety functions.5. Inadvertent CriticalityInadvertent criticalities pose potential personnel safety hazards as well being indicativeof losses of reactivity control.6. CommunicationsCertain events that degrade plant operator ability to effectively communicate withessential personnel within or external to the plant warrant emergency classification.Page 110EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 1 -Loss of AC PowerInitiating Condition: Loss of all offsite and all onsite AC power to 4.16 KV emergencybuses for _ 15 min.EAL:CA1.1 AlertLoss of all offsite and all onsite AC power, Table C-1, to 4.16 KV emergency buses2ENS*SWG101 and 2ENS*SWG103 for _ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.I Table C-1 AC Power Sources I* 2EGS*EG1Q"o

• 2EGS*EG30* Reserve Transformer AO Reserve Transformer B0 0 Aux Boiler TransformerMode Applicability:4 -Cold Shutdown, 5 -Refuel, D -DefueledBasis:Plant-Specific2ENS*SWG101, 2ENS*SWG102, and 2ENS*SWG103 are the 4.16 KV emergency buses.Bus 2ENS*SWG101 is dedicated to Division I of the On-site Emergency AC ElectricalDistribution System, bus 2ENS*SWG102 is dedicated to Division III (HPCS), and bus2ENS*SWG103 is dedicated to Division I1. Buses 2ENS*SWG101 and *SWG103 feed allStation redundant safety-related loads, except the HPCS system loads. The HPCS systemloads are fed by bus 2ENS*SWG1 02 (ref. 1, 2).All three divisions are normally energized by the On-site Normal AC ElectricalDistribution System via the off-site power sources through the reserve stationservice transformers 2RTX-XSRIA and 2RTX-XSR1 B.Page 111 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Baseso 2ENS*SWG102 from transformer 2RTX-XSRIAo 2ENS*SWG103 from transformer 2RTX-XSR1B." Buses 2ENS*SWG101 and 2ENS*SWG103 each have a backup source, theAuxiliary Boiler Transformer 2ABS-Xl. Also, 2ENS*SWG101 and *SWG103 eachhave a feeder to a normal AC (stub) bus, NNS-SWG014 and NNS-SWG015respectively." Bus 2ENS*SWG102 has a backup connection to the Reserve Station ServiceTransformer 2RTX-XSR1 B, if required., Each of the three 4.16 KV emergency buses has a standby diesel generator(2EGS*EG1, 2EGS*EG3, 2EGS*EG2) to carry its loads in case of a LOOP or incase of a sustained degraded voltage condition on the offsite source (ref. 3, 4).Consideration should be given to operable loads necessary to remove decay heat orprovide RPV makeup capability when evaluating loss of all AC power to vital buses. Eventhough an essential bus may be energized, if necessary loads (i.e., loads that if lost wouldinhibit decay heat removal capability or RPV makeup capability) are not operable on theenergized bus then the bus should not be considered operable.The fifteen-minute interval was selected as a threshold to exclude transient power losses.This EAL is the cold condition equivalent of the hot condition loss of all AC power EALSS1.1.GenericLoss of all AC power compromises all plant safety systems requiring electric power including RHR,ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink.The event can be classified as an Alert when in cold shutdown, refuel, or defueled mode becauseof the significantly reduced decay heat and lower temperature and pressure, increasing the time torestore one of the emergency buses, relative to that specified for the Site Area Emergency EAL.Escalating to Site Area Emergency, if appropriate, is by EALs in Category R.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.NMP2 Basis Reference(s):1. USAR Section 8.22. USAR Section 8.3Page 112 EPMP-EPP-0102Rev 00 (Draft A)

Attachment I -Emergency Action Level Technical Bases3. N2-SOP-03 Loss of AC Power4. N2-SOP-01 Station Blackout5. NEI 99-01 IC CA3Page 113EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction1 -Loss of AC PowerAC power capability to 4.16 KV emergency buses reduced to asingle power source for __ 15 min. such that ANY additional singlefailure would result in a complete loss of all 4.16 KV emergencybus powerEAL:CUI.1 Unusual EventAC power capability to 4.16 KV emergency buses 2ENS*SWG101 and 2ENS*SWG103reduced to a single power source, Table C-1, for >_ 15 min. (Note 4)ANDANY additional single power source failure will result in a loss of all power to 4.16 KVemergency buses 2ENS*SWG101 and 2ENS*SWG103Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-1 AC Power Sources0 2EGS*EG1")

• 2EGS*EG30* Reserve Transformer A0 Reserve Transformer Bo 0 Aux Boiler TransformerMode Applicability:4 -Cold Shutdown, 5 -Refuel, D -DefueledBasis:Plant-Specific2ENS*SWG101, 2ENS*SWG102, and 2ENS*SWG103 are the 4.16 KV emergency buses.Bus 2ENS*SWG1 01 is dedicated to Division I of the On-site Emergency AC ElectricalDistribution System, bus 2ENS*SWG102 is dedicated to Division III (HPCS), and bus2ENS*SWG103 is dedicated to Division I1. Buses 2ENS*SWG101 and 2ENS*SWG103feed all Station redundant safety-related loads, except the HPCS system loads. The HPCSPage 114EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical Basessystem loads are fed by bus 2ENS*SWG1 02 (ref. 1, 2).* All three divisions are normally energized by the On-site Normal AC ElectricalDistribution System via the off-site power sources through the reserve stationservice transformers 2RTX-XSRIA and 2RTX-XSR1 B.o 2ENS*SWG102 from transformer 2RTX-XSRIAo 2ENS*SWG103 from transformer 2RTX-XSR1B." Buses 2ENS*SWG101 and 2ENS*SWG103 each have a backup source, theAuxiliary Boiler Transformer 2ABS-X1. Also, 2ENS*SWG1 01 and 2ENS*SWG1 03each have a feeder to a normal AC (stub) bus, NNS-SWG014 and NNS-SWG015respectively." Bus 2ENS*SWG102 has a backup connection to the Reserve Station ServiceTransformer 2RTX-XSRI B, if required.* Each of the three 4.16 KV emergency buses has a standby diesel generator(2EGS*EG1, 2EGS*EG3, 2EGS*EG2) to carry its loads in case of a LOOP or incase of a sustained degraded voltage condition on the offsite source (ref. 3, 4).The fifteen-minute interval was selected as a threshold to exclude transient power losses.If multiple sources fail to energize the unit safety-related buses within 15 minutes, anUnusual Event is declared under this EAL. The subsequent loss of the single remainingpower source escalates the event to an Alert under EAL CA1.1.GenericThe condition indicated by this EAL is the degradation of the off-site and on-site AC power systemssuch that any additional single failure would result in a complete loss of 4.16 KV emergency busAC power to one or both units. This condition could occur due to a loss of off-site power with aconcurrent failure of all but one emergency generator to supply power to its emergency bus. Thesubsequent loss of this single power source would escalate the event to an Alert in accordancewith EAL CA1.1.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.NMP2 Basis Reference(s):1. USAR Section 8.22. USAR Section 8.33. N2-SOP-03 Loss of AC PowerPage 115 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases4. N2-SOP-01 Station Blackout5. NEI 99-01 IC CU3Page 116EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 2 -Loss of DC PowerInitiating Condition: Loss of required DC power for __ 15 min.EAL:CU2.1 Unusual Event< 105 VDC on required 125 VDC emergency buses for > 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Mode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificThe emergency 125 VDC power system includes three electrically independent andseparate switchgears (2BYS*SWG002A, 2BYS*SWG002B and 2CES*IPNL414). Division I((2BYS*SWG002A) and Division II (2BYS*SWG002B) feed the redundant emergency DCloads associated with Divisions I and II of the emergency onsite AC system, respectively.Division III (2CES*PNP414) feeds the emergency DC loads associated with Division III(HPCS system).Each emergency 125 VDC distribution system has a battery and a battery charger that arenormally connected to the bus such that these two sources of power are operating inparallel. The charger is normally supplying system electrical loads with the battery on afloat charge. Should both battery chargers for any particular battery be out of service atany point in the DC load cycle, the battery is capable of starting and operating itsassociated loads for 2 hr according to a precalculated load profile without the batteryterminal voltage falling below minimum acceptable level, 105 VDC. (ref. 1, 2, 3)In Cold Shutdown mode and Refuel mode, requirements on emergency 125 VDC powerare relaxed. The term "required" in this EAL signifies the minimum Technical Specificationsrequirements for shutdown conditions (ref. 2):* One Division I or Division II DC electrical power subsystem; andPage 117 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases, Division III DC electrical power subsystem when the HPCS system is required to beoperable.This EAL is the cold condition equivalent of the hot condition loss of DC powerEAL SS2.1.GenericThe purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor andcontrol the removal of decay heat during Cold Shutdown or Refueling operations.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.NMP2 Basis Reference(s):1. USAR Section 8.3.2.1.22. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.8.53. N2-SOP-04 Loss of DC Power4. NEI 99-01 IC CU7Page 118EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:C -Cold Shutdown / Refueling System Malfunction3 -RPV LevelLoss of RPV inventory affecting fuel clad integrity withContainment challengedEAL:CG3.1 General EmergencyRPV level < -14 in. for_ 30 min. (Note 4)ANDANY Containment Challenge Indication, Table C-3Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-3 Containment Challenge Indications* CONTAINMENT CLOSURE not established* Explosive mixture exists inside PrimaryContainment (H2 > 6% and 02 > 5%)* UNPLANNED rise in Primary Containmentpressure* RB area radiation > 8.OOE+3 mR/hrMode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificWhen RPV level drops the top of active fuel (an indicated RPV level of -14 in.), coreuncovery starts to occur (ref. 1, 2).Four conditions are associated with a challenge to Primary Containment integrity:CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteriagoverning Primary and Secondary Containment operability. If the TechnicalPage 119EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesSpecification criteria are met, therefore, CONTAINMENT CLOSURE has beenestablished. (ref. 3, 4, 5)Explosive (deflagration) mixtures in the Primary Containment are assumed to beelevated concentrations of hydrogen and oxygen. BWR industry evaluation ofhydrogen generation for development of EOPs/SAGs indicates that any hydrogenconcentration above minimum detectable is not to be expected within the shortterm. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowlyevolving, long-term condition. Hydrogen concentrations that rapidly develop aremost likely caused by metal-water reaction. A metal-water reaction is indicative ofan accident more severe than accidents considered in the plant design basis andwould be indicative, therefore, of a potential threat to Primary Containment integrity.Hydrogen concentration of approximately 6% is considered the global deflagrationconcentration limit.The specified values for this threshold are the minimum global deflagrationconcentration limits (6% hydrogen and 5% oxygen), and readily recognizablebecause 6% hydrogen is well above the EOP flowchart entry condition. Theminimum global deflagration hydrogen/oxygen concentrations (6%/5%, respectively)require intentional Primary Containment venting, which is defined to be a loss of thePrimary Containment barrier. (ref. 6, 7)The USAR requires the H2/02 analyzers to be able to provide and recordcombustible gas concentration in the Primary Containment within 90 minutesfollowing a LOCA with safety system injection. The H2/02 analyzers are normally instandby and require a 30 minute warm-up/self-test period before they startproviding data. (ref. 6)If the hydrogen or oxygen monitor is unavailable, sampling and analysis maydetermine gas concentrations. The validity of sample results must be judged basedupon plant conditions, since drawing and analyzing samples may take some time. Ifsample results cannot be relied upon and hydrogen concentrations cannot bedetermined by any other means, the concentrations must be considered "unknown."The monitors should not be considered "unavailable" until an attempt has beenPage 120 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesmade to place them in service. (ref. 2)" Any UNPLANNED rise in Primary Containment pressure in the Cold Shutdown orRefuel mode indicates CONTAINMENT CLOSURE cannot be assured and thePrimary Containment cannot be relied upon as a barrier to fission product release.* RB (Reactor Building) area radiation monitors should provide indication of increasedrelease that may be indicative of a challenge to CONTAINMENT CLOSURE. TheEOP Maximum Safe Operating level is 8.OOE+3 mR/hr and is indicative of problemsin the secondary containment that are spreading. The locations into which theprimary system discharge is of concern correspond to the areas addressed in DetailS of N2-EOP-SC (ref. 7).If RPV level is restored and maintained above the top of active fuel before a ContainmentChallenge condition occurs and subsequently a Containment Challenge condition isreached, this EAL is not met.GenericThis EAL represents the inability to restore and maintain RPV water level to above the top of activefuel with containment challenged. Fuel damage is probable if RPV water level cannot be restored,as available decay heat will cause boiling, further reducing the RPV water level. With theContainment breached or challenged then the potential for unmonitored fission product release tothe environment is high. This represents a direct path for radioactive inventory to be released tothe environment. This is consistent with the definition of a GE. The GE is declared on theoccurrence of the loss or IMMINENT loss of function of all three barriers.A number of variables can have a significant impact on heat removal capability challenging the fuelclad barrier. Examples include: mid-loop, reduced level/flange level, head in place, cavity flooded,RCS venting strategy, decay heat removal system design, vortexing pre-disposition, steamgenerator U-tube drainingAnalysis indicates that core damage may occur within an hour following continued core uncoverytherefore, 30 minutes was conservatively chosen.If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute core uncoverytime limit then escalation to General Emergency would not occur.NMP2 Basis Reference(s):1. N2-EOP-RPV RPV Control2. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document3. NIP-OUT-01 Shutdown Safety4. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.1.15. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.4.1Page 121 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases6. N2-EOP-PCH Hydrogen Control7. N2-EOP-SC Secondary Containment Control8. NEI 99-01 IC CG1Page 122EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV LevelInitiating Condition: Loss of RPV inventory affecting fuel clad integrity withContainment challengedEAL:CG3.2 General EmergencyRPV water level cannot be monitored with core uncovery indicated by ANY of thefollowing for _ 30 min. (Note 4):" ANY UNPLANNED RPV leakage indication, Table C-2* Erratic Source Range Monitor indicationANDANY Containment Challenge Indication, Table C-3Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-2 RPV Leakage Indications* Drywell equipment drain sump level rise* Drywell floor drain sump level rise* Reactor building equipment sump level rise* Reactor Building floor drain sump level rise* Suppression Pool level rise* UNPLANNED rise in RPV make-up rate* Observation of UNISOLABLE RCS leakageTable C-3 Containment Challenge Indications* CONTAINMENT CLOSURE not established* Explosive mixture exists inside PrimaryContainment (H2 > 6% and 02 > 5%)* UNPLANNED rise in Primary Containmentpressure* RB area radiation > 8.OOE+3 mR/hrPage 123EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesMode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificIf RPV water level monitoring capability is unavailable, all RPV water level indication wouldbe unavailable and, the RPV inventory loss must be detected by Table C-2, RPV LeakageIndications. Level increases must be evaluated against other potential sources of leakagesuch as cooling water sources inside the drywell to ensure they are indicative of RPVleakage. Drywell equipment and floor drain sump level rise is the normal method ofmonitoring and calculating leakage from the RPV. A Reactor Building equipment or floordrain sump level rise may also be indicative of RPV inventory losses external to thePrimary Containment from systems connected to the RPV. With RHR System operating inthe Shutdown Cooling mode, an UNPLANNED rise in suppression pool level could beindicative of RHR valve misalignment or leakage. If the make-up rate to the RPVunexplainably rises above the pre-established rate, a loss of RPV inventory may beoccurring even if the source of the leakage cannot be immediately identified. Visualobservation of leakage from systems connected to the RCS in areas outside the PrimaryContainment that cannot be isolated could be indicative of a loss of RPV inventory. (ref. 1,2,3)Four channels of log count rate meters are available in the Control Room to detect erraticsource range monitor indications (ref. 4):" SRM A & C on 2CEC*PNL606* SRM B & D on 2CEC*PNL633Post-TMI studies indicated that the installed nuclear instrumentation will operate erraticallywhen the core is uncovered and that source range monitors can be used as a tool formaking such determinations. Figure C-2 shows the response of the source range monitorduring the first few hours of the TMI-2 accident. The instrument reported an increasingsignal about 30 minutes into the accident. At this time, the reactor coolant pumps wererunning and the core was adequately cooled as indicated by the core outletthermocouples. Hence, the increasing signal was the result of an increasing two-phasePage 124 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesvoid fraction in the reactor core and vessel downcomer and the reduced shielding that thetwo-phase mixture provides to the source range monitor.Four conditions are associated with a challenge to Primary Containment integrity:CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteriagoverning Primary and Secondary Containment operability. If the TechnicalSpecification criteria are met, therefore, CONTAINMENT CLOSURE has beenestablished. (ref. 5, 9, 10)* Explosive (deflagration) mixtures in the Primary Containment are assumed to beelevated concentrations of hydrogen and oxygen. BWR industry evaluation ofhydrogen generation for development of EOPs/SAGs indicates that any hydrogenconcentration above minimum detectable is not to be expected within the shortterm. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowlyevolving, long-term condition. Hydrogen concentrations that rapidly develop aremost likely caused by metal-water reaction. A metal-water reaction is indicative ofan accident more severe than accidents considered in the plant design basis andwould be indicative, therefore, of a potential threat to Primary Containment integrity.Hydrogen concentration of approximately 6% is considered the global deflagrationconcentration limit.The specified values for this threshold are the minimum global deflagrationconcentration limits (6% hydrogen and 5% oxygen), and readily recognizablebecause 6% hydrogen is well above the EOP flowchart entry condition. Theminimum global deflagration hydrogen/oxygen concentrations (6%/5%, respectively)require intentional Primary Containment venting, which is defined to be a loss of thePrimary Containment barrier. (ref. 6, 7)The USAR requires the H2/02 analyzers to be able to provide and recordcombustible gas concentration in the Primary Containment within 90 minutesfollowing a LOCA with safety system injection. The H2/02 analyzers are normally inPage 125 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesstandby and require a 30 minute warrn-up/self-test period before they startproviding data. (ref. 6)If the hydrogen or oxygen monitor is unavailable, sampling and analysis maydetermine gas concentrations. The validity of sample results must be judged basedupon plant conditions, since drawing and analyzing samples may take some time. Ifsample results cannot be relied upon and hydrogen concentrations cannot bedetermined by any other means, the concentrations must be considered "unknown."The monitors should not be considered "unavailable" until an attempt has beenmade to place them in service. (ref. 7)* Any UNPLANNED rise in Primary Containment pressure in the Cold Shutdown orRefuel mode indicates CONTAINMENT CLOSURE cannot be assured and thePrimary Containment cannot be relied upon as a barrier to fission product release.* RB (Reactor Building) area radiation monitors should provide indication of increasedrelease that may be indicative of a challenge to CONTAINMENT CLOSURE. TheEOP Maximum Safe Operating level is 8.OOE+3 mR/hr and is indicative of problemsin the secondary containment that are spreading. The locations into which theprimary system discharge is of concern correspond to the areas addressed in DetailS of N2-EOP-SC (ref. 8).If RPV level is restored and maintained above the top of active fuel before a ContainmentChallenge condition occurs and subsequently a Containment Challenge condition isreached, this EAL is not met.GenericThis EAL represents the inability to restore and maintain RPV water level to above the top of activefuel with containment challenged. Fuel damage is probable if RPV water level cannot be restored,as available decay heat will cause boiling, further reducing the RPV water level. With theContainment breached or challenged then the potential for unmonitored fission product release tothe environment is high. This represents a direct path for radioactive inventory to be released tothe environment. This is consistent with the definition of a GE. The GE is declared on theoccurrence of the loss or IMMINENT loss of function of all three barriers.A number of variables can have a significant impact on heat removal capability challenging the fuelclad barrier. Examples include: initial RPV water level, shutdown heat removal system design.Analysis indicates that core damage may occur within an hour following continued core uncoverytherefore, 30 minutes was conservatively chosen.Page 126 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesIf CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute core uncoverytime limit then escalation to General Emergency would not occur.Sump and tank level increases must be evaluated against other potential sources of leakage suchas cooling water sources inside the containment to ensure they are indicative of RCS leakage.As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due tothis core shine should result in site specific monitor indication and possible alarm.Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically whenthe core is uncovered and that this should be used as a tool for making such determinations.NMP2 Basis Reference(s):1. USAR Section 5.2.52. USAR Section 7.6.1.33. N2-EOP-PC Primary Containment Control4. N2-OP-92 Neutron Monitoring5. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.4.16. N2-EOP-PCH Hydrogen Control7. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document8. N2-EOP-SC Secondary Containment Control9. NIP-OUT-01 Shutdown Safety10. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.1.111. NEI 99-01 IC CG1Page 127EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesFigure C-2: Response of the TMI-2 Source Range MeasurementDuring the First Six Hours of the AccidentVCq004JCIE)CDE10-.0 -CM~ :3CDLt I(sapeoap 6ol) puooaS ja sjunooPage 128EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3- RPV LevelInitiating Condition: Loss of RPV inventory affecting core decay heat removal capabilityEAL:CS3.1 Site Area EmergencyWith CONTAINMENT CLOSURE not established, RPV water level < 11.8 in.Mode Applicability:4- Cold Shutdown, 5 -RefuelBasis:Plant-SpecificWhen RPV water level decreases to 11.8 in., water level is six inches below the low-low-low ECCS actuation setpoint (ref. 1).The inability to restore and maintain level after reaching this setpoint infers a failure of theRCS barrier and Potential Loss of the Fuel Clad barrier.CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteria governingPrimary and Secondary Containment operability. If the Technical Specification criteria aremet, therefore, CONTAINMENT CLOSURE has been established. (ref. 2, 3, 4)GenericUnder the conditions specified by this EAL, continued decrease in RPV level is indicative of a lossof inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, orcontinued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.Escalation to a General Emergency is via EAL CG3.1, EAL CG3.2, RG1.1, RG1.2 or RG1.3.NMP2 Basis Reference(s):1. N2-OP-33 High Pressure Core Spray2. NIP-OUT-01 Shutdown Safety3. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.1.14. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.4.15. NEI 99-01 IC CS1Page 129 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV LevelInitiating Condition: Loss of RPV inventory affecting core decay heat removal capabilityEAL:CS3.2 Site Area EmergencyWith CONTAINMENT CLOSURE established, RPV water level < -14 in.Mode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificWhen RPV level drops the top of active fuel (an indicated RPV level of -14 in.), coreuncovery starts to occur (ref. 1, 2).CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteria governingPrimary and Secondary Containment operability. If the Technical Specification criteria aremet, therefore, CONTAINMENT CLOSURE has been established. (ref. 3, 4, 5)GenericUnder the conditions specified by this EAL, continued decrease in RPV level is indicative of a lossof inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, orcontinued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.Escalation to a General Emergency is via EAL CG3.1, EAL CG3.2, RG1.1, RG1.2 or RG1.3.NMP2 Basis Reference(s):1. N2-EOP-RPV RPV Control2. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document3. NIP-OUT-01 Shutdown Safety4. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.1.15. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.4.16. NEI 99-01 IC CS1Page 130 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV LevelInitiating Condition: Loss of RPV inventory affecting core decay heat removal capabilityEAL:CS3.3 Site Area EmergencyRPV water level cannot be monitored for >_ 30 min. (Note 4) with a loss of RPV inventoryas indicated by ANY of the following:* ANY UNPLANNED RPV leakage indication, Table C-2" Erratic Source Range Monitor indicationNote 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-2 RPV Leakage Indications* Drywell equipment drain sump level rise* Drywell floor drain sump level rise* Reactor building equipment sump level rise* Reactor Building floor drain sump level rise* Suppression Pool level rise* UNPLANNED rise in RPV make-up rate* Observation of UNISOLABLE RCS leakageMode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificIf RPV water level monitoring capability is unavailable, all RPV water level indication wouldbe unavailable and, the RPV inventory loss must be detected by Table C-2, RPV LeakageIndications. Level increases must be evaluated against other potential sources of leakagesuch as cooling water sources inside the drywell to ensure they are indicative of RPVleakage. Drywell equipment and floor drain sump level rise is the normal method ofmonitoring and calculating leakage from the RPV. A Reactor Building equipment or floordrain sump level rise may also be indicative of RPV inventory losses external to thePage 131 EPMP-EPP-0102.Rev 00 (Draft A) -Emergency Action Level Technical BasesPrimary Containment from systems connected to the RPV. With RHR System operating inthe Shutdown Cooling mode, an UNPLANNED rise in suppression pool level could beindicative of RHR valve misalignment or leakage. If the make-up rate to the RPVunexplainably rises above the pre-established rate, a loss of RPV inventory may beoccurring even if the source of the leakage cannot be immediately identified. Visualobservation of leakage from systems connected to the RCS in areas outside the PrimaryContainment that cannot be isolated could be indicative of a loss of RPV inventory. (ref. 1,2,3)Four channels of log count rate meters are available in the Control Room to detect erraticsource range monitor indications (ref. 4):" SRM A & C on 2CEC*PNL606" SRM B & D on 2CEC*PNL633Post-TMI studies indicated that the installed nuclear instrumentation will operate erraticallywhen the core is uncovered and that source range monitors can be used as a tool formaking such determinations. Figure C-2 shows the response of the source range monitorduring the first few hours of the TMI-2 accident. The instrument reported an increasingsignal about 30 minutes into the accident. At this time, the reactor coolant pumps wererunning and the core was adequately cooled as indicated by the core outletthermocouples. Hence, the increasing signal was the result of an increasing two-phasevoid fraction in the reactor core and vessel downcomer and the reduced shielding that thetwo-phase mixture provides to the source range monitor.GenericUnder the conditions specified by this EAL, continued decrease in RPV level is indicative of a lossof inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, orcontinued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.Escalation to a General Emergency is via EAL CG3.1, EAL CG3.2, RG1.1, RG1.2 or RG1.3.The 30-minute duration allows sufficient time for actions to be performed to recover inventorycontrol equipment.As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due tothis core shine should result in site specific monitor indication and possible alarm.NMP2 Basis Reference(s):Page 132 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases1. USAR Section 5.2.52. USAR Section 7.6.1.33. N2-EOP-PC Primary Containment Control4. N2-OP-92 Neutron Monitoring5. NEI 99-01 IC CS1Page 133EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesFigure C-2: Response of the TMI-2 Source Range MeasurementDuring the First Six Hours of the AccidentC'5.IV0E.9_cJi--004ECa0 (D<~0 ~0o -rCo U~) C(Sapuoap 601) puO09S jad s~unooPage 134EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3- RPV LevelInitiating Condition: Loss of RPV inventoryEAL:CA3.1 AlertRPV water level < 17.8 in.ORRPV water level cannot be monitored for _ 15 min. with ANY UNPLANNED RPV leakageindication, Table C-2 (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table C-2 RPV Leakage Indications* Drywell equipment drain sump level rise* Drywell floor drain sump level rise* Reactor building equipment sump level rise* Reactor Building floor drain sump level rise* Suppression Pool level rise* UNPLANNED rise in RPV make-up rate* Observation of UNISOLABLE RCS leakageMode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificThe threshold RPV water level of 17.8 in. is the low-low-low ECCS actuation setpoint (ref.1).Figure C-1 illustrates the RPV water level instrument ranges (ref. 2, 3).In Cold Shutdown mode, the RCS will normally be INTACT and standard RPV water levelmonitoring means are available. In the Refuel mode, the RCS is not INTACT and RPVwater level may be monitored by different means, including the ability to monitor levelvisually.Page 135 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesIn the second condition of this EAL, all RPV water level indication would be unavailableand, the RPV inventory loss must be detected by Table C-2, RPV Leakage Indications.Level increases must be evaluated against other potential sources of leakage such ascooling water sources inside the drywell to ensure they are indicative of RPV leakage.Drywell equipment and floor drain sump level rise is the normal method of monitoring andcalculating leakage from the RPV. A Reactor Building equipment or floor drain sump levelrise may also be indicative of RPV inventory losses external to the Primary Containmentfrom systems connected to the RPV. With RHR System operating in the Shutdown Coolingmode, an UNPLANNED rise in suppression pool level could be indicative of RHR valvemisalignment or leakage. If the make-up rate to the RPV unexplainably rises above thepre-established rate, a loss of RPV inventory may be occurring even if the source of theleakage cannot be immediately identified. Visual observation of leakage from systemsconnected to the RCS in areas outside the Primary Containment that cannot be isolatedcould be indicative of a loss of RPV inventory. (ref. 4, 5, 6)Depending on the configuration of the reactor cavity and Spent Fuel Pool (gates installedor removed) and the status of refueling operations (all spent fuel seated in storageracks/RPV or a bundle raised on the fuel grapple), a loss of inventory may reduce watershielding above irradiated components or spent fuel. EALs in Subcategory R.2 should beassessed for emergency classification due to the radiological consequences of suchevents.GenericThis EAL serves as a precursor to a loss of ability to adequately cool the fuel. The magnitude ofthis loss of water indicates that makeup systems have not been effective and may not be capableof preventing further RPV water level decrease and potential core uncovery. This condition willresult in a minimum emergency classification level of an Alert.The inability to restore and maintain level after reaching this setpoint would be indicative of afailure of the RCS barrier.If RPV water level continues to lower then escalation to Site Area Emergency will be via EALCS3.1, EAL CS3.2 or EAL CS3.3.NMP2 Basis Reference(s):1. N2-OP-33 High Pressure Core Spray2. N2-EOP-RPV RPV Control3. N2-OP-34 Nuclear Boiler, Automatic Depressurization, and Safety Relief ValvesPage 136 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases4. USAR Section 5.2.55. USAR Section 7.6.1.36. N2-EOP-PC Primary Containment Control7. NEI 99-01 IC CA1Page 137EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesFigure C-1 RPV Water Level Instrumentation Ranges (ref. 2, 3)I.I.325',HIGH LEVEL TRIP.HIGH LEVEL ALARMNORMAL WATER LEVELLOW LEVEL ALARM.LOW LEVEL TRIP202.3187.3 --178, 3159.30w(Dw0wILI35'DOUBLE LOW LEVEL TRIP 108.8TRIPE LOW LEVEL TRIP 17.8INSTRUMENT ZERO 0TOP OFACTIVE FUEL -14.01INOTE:ALL LEVELS ARE- REFERENCEDTO INSTRUMENT ZERO(380,69" ABOVE VESSEL ZERO)JET PUMP INSTR,,ACTIVE RANGE--INACTIVE RANGEPage 138EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: RCS leakageEAL:CU3.1 Unusual EventRCS leakage results in the inability to maintain or restore RPV water level > 159.3 in.for 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeMode Applicability:4 -Cold ShutdownBasis:Plant-SpecificFigure C-1 illustrates the RPV water level instrument ranges (ref. 1, 2).159.3 in. is the RPV low water level scram setpoint (ref. 1).RPV water level is monitored from -165 in. to +545 in. to ensure adequate coverage forexpected and postulated conditions of RPV water level. RPV water level measurement isderived by the differential pressure that exists between a reference leg and variable leg. Alllevel instruments are referenced to an "instrument zero", which is 380.69 inches above"vessel zero". The instrument zero is the top of the reactor vessel upper grid (top guide).RPV water level monitoring is subdivided into five ranges identified as:* Narrow provides indication and control signals for normal plant operation andprotection system actuation.* Wide provides indication and control signals for transient conditions below thenormal operating band and emergency equipment actuation.* Upset provides indication for transient conditions above normal operating band." Shutdown provides indication for vessel flood up and activities.* Fuel Zone provides indication for long term accident conditions where reactor levelcannot be restored.Page 139 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesThe shutdown range level indication is utilized during cold reactor startup and vessel floodup for refueling. The shutdown range instrument uses a single level transmitter(21SC*LT1 05) to provide an input to a level indicator on 2CES*PNL851 (Computer PointA486). (ref. 3)This Cold Shutdown EAL represents the hot condition EAL SU8.1, in which RCS leakageis associated with Technical Specification limits. In Cold Shutdown, these limits are notapplicable; hence, the use of RPV level as the parameter of concern in this EAL (ref.).GenericThis EAL is considered to be a potential degradation of the level of safety of the plant. The inabilityto maintain or restore level is indicative of loss of RCS inventory.Relief valve normal operation should be excluded from this EAL. However, a relief valve thatoperates and fails to close per design should be considered applicable to this EAL if the relief valvecannot be isolated.Prolonged loss of RCS inventory may result in escalation to the Alert emergency classification levelvia either EAL CA2.1 or EAL CA3.1.NMP2 Basis Reference(s):1. N2-EOP-RPV RPV Control2. N2-OP-34 Nuclear Boiler, Automatic Depressurization, and Safety Relief Valves3. NIP-OUT-01 Shutdown Safety4. NEI 99-01 IC CU1Page 140 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesFigure C-1 RPV Water Level Instrumentation Ranges (ref. 1, 2)325,HIGH LEVEL TRIPHIGH LEVEL ALARMN 0 RMAL WATE R. LEVELLOW LEVEL ALARMLOW. LEVEL TRIP202.3 j187.3 --178.3159.3w(Dz0w(Dwz0wDtILIII.5.5II351DOUBLE LOWLEVEL TRIP 108.8TRIPE LOW LEVEL TRIP 17.8INSTRUMENT ZERO 0TOP OF ACTIVE FUEL -14.04I-165dINOTE:ALL LEVELS ARE. REFERENCEDTO INSTRUMENT ZERO(380.69" ABOVE VESSEL ZERO)JET PUMP INSTR.ACTIVE RANGE--INACTIVE RANGEPage 141EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water Level.Initiating Condition: RCS LeakageEAL:CU3.2 Unusual EventUNPLANNED RPV water level drop below EITHER of the following for > 15 min. (Note 4):* 364 in. (RPV flange)0 RPV water level band (when the RPV water level band is established below theRPV flange)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeMode Applicability:5 -RefuelBasis:Plant-SpecificThe RPV flange level is at 364 in. or 330 ft 10 in. el (ref. 1).Figure C-1 illustrates the RPV water level instrument ranges (ref. 2, 3).RPV water level is monitored from -165 in. to +545 in. to ensure adequate coverage forexpected and postulated conditions of RPV water level. RPV water level measurement isderived by the differential pressure that exists between a reference leg and variable leg. Alllevel instruments are referenced to an "instrument zero", which is 380.69 inches above"vessel zero". The instrument zero is the top of the reactor vessel upper grid (top guide).RPV water level monitoring is subdivided into five ranges identified as:" Narrow provides indication and control signals for normal plant operation andprotection system actuation." Wide provides indication and control signals for transient conditions below thenormal operating band and emergency equipment actuation." Upset provides indication for transient conditions above normal operating band." Shutdown provides indication for vessel flood up and activities." Fuel Zone provides indication for long term accident conditions where reactor levelPage 142 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basescannot be restored.The shutdown range level indication is utilized during cold reactor startup and vessel floodup for refueling. The shutdown range instrument uses a single level transmitter(21SC*LT1 05) to provide an input to a level indicator on 2CES*PNL851 (Computer PointA486). (ref. 4)This Cold Shutdown EAL represents the hot condition EAL SU8.1, in which RCS leakageis associated with Technical Specification limits. In Cold Shutdown, these limits are notapplicable; hence, the use of RPV water level as the parameter of concern in this EAL (ref.5).GenericThis EAL is a precursor of more serious conditions and considered to be a potential degradation ofthe level of safety of the plant.Refueling evolutions that decrease RPV water level below the RPV flange are carefully plannedand procedurally controlled. An UNPLANNED event that results in water level decreasing belowthe RPV flange, or below the planned RPV water level for the given evolution (if the planned RPVwater level is already below the RPV flange), warrants declaration of a UE due to the reduced RCSinventory that is available to keep the core covered.The allowance of 15 minutes was chosen because it is reasonable to assume that level can berestored within this time frame using one or more of the redundant means of refill that should beavailable. If level cannot be restored in this time frame then it may indicate a more seriouscondition exists.Continued loss of RCS Inventory will result in escalation to the Alert emergency classification levelvia either EAL CA2.1 or EAL CA3.1.This EAL involves a decrease in RCS level below the top of the RPV flange that continues for 15minutes due to an UNPLANNED event. This EAL is not applicable to decreases in flooded reactorcavity level, which is addressed by EAL RU2.1, until such time as the level decreases to the levelof the vessel flange.NMP2 Basis Reference(s):1. N2-SOP-31 R Refueling Operations Alternate Shutdown Cooling2. N2-EOP-RPV RPV Control3. N2-OP-34 Nuclear Boiler, Automatic Depressurization, and Safety Relief Valves4. NIP-OUT-01 Shutdown Safety5. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.4.76. NEI 99-01 IC CU2Page 143 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesFigure C-1 RPV Water Level Instrumentation Ranges (ref. 2, 3)IL(D!wIw0 LU0ý Q wo LLHIGH LEVEL TRIP,HIGH LEVEL ALARMNORMAL WATER LEVELLOW LEVEL ALARM.LOW LEVEL TRIP202.3 j187.3 --178.3159.3-IIIDOUBLE LOW LEVEL TRIP 108.8TRIPE LOW LEVEL TRIP 17.8INSTRUMENT ZERO 0TOP OF ACTIVE FUEL -140I35,-165iINOTE: ET PUMP INSTR,ALL LEVELS ARE REFERENCED JET P ACTIVE RANGETO INSTRUMENT ZERO(380,69".ABOVE VESSEL ZERO) ..TNACTIVE RANGEPage 144EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 3 -RPV Water LevelInitiating Condition: RCS LeakageEAL:CU3.3 Unusual EventRPV water level cannot be monitored with a loss of RPV inventory as indicated by ANYUNPLANNED RPV leakage indication, Table C-2Table C-2 RPV Leakage Indications* Drywell equipment drain sump level rise* Drywell floor drain sump level rise* Reactor building equipment sump level rise* Reactor Building floor drain sump level rise* Suppression Pool level rise* UNPLANNED rise in RPV make-up rate* Observation of UNiSOLABLE RCS leakageMode Applicability:5 -RefuelBasis:Plant-SpecificIn this EAL, all RPV water level indication would be unavailable and, the RPV inventoryloss must be detected by Table C-2, RPV Leakage Indications. Level increases must beevaluated against other potential sources of leakage such as cooling water sources insidethe drywell to ensure they are indicative of RPV leakage. Drywell equipment and floordrain sump level rise is the normal method of monitoring and calculating leakage from theRPV. A Reactor Building equipment or floor drain sump level rise may also be indicative ofRPV inventory losses external to the Primary Containment from systems connected to theRPV. With RHR System operating in the Shutdown Cooling mode, an UNPLANNED rise insuppression pool level could be indicative of RHR valve misalignment or leakage. If themake-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPVPage 145 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesinventory may be occurring even if the source of the leakage cannot be immediatelyidentified. Visual observation of leakage from systems connected to the RCS in areasoutside the Primary Containment that cannot be isolated could be indicative of a loss ofRPV inventory. (ref. 1, 2, 3)Depending on the configuration of the reactor cavity and Spent Fuel Pool (gates installedor removed) and the status of refueling operations (all spent fuel seated in storageracks/RPV or a bundle raised on the fuel grapple), a loss of inventory may reduce watershielding above irradiated components or spent fuel. EALs in Subcategory R.2 should beassessed for emergency classification due to the radiological consequences of suchevents.GenericThis EAL is a precursor of more serious conditions and considered to be a potential degradation ofthe level of safety of the plant.Refueling evolutions that decrease RPV water level below the RPV flange are carefully plannedand procedurally controlled. An UNPLANNED event that results in water level decreasing belowthe RPV flange, or below the planned RPV water level for the given evolution (if the planned RPVwater level is already below the RPV flange), warrants declaration of a UE due to the reduced RPVinventory that is available to keep the core covered.Continued loss of RCS Inventory will result in escalation to the Alert emergency classification levelvia either EAL CA3.1 or EAL CA4.1. ,This EAL addresses conditions in the refueling mode when normal means of core temperatureindication and RCS level indication may not be available. Redundant means of RPV water levelindication will normally be installed (including the ability to monitor level visually) to assure that theability to monitor level will not be interrupted. However, if all level indication were to be lost during aloss of RPV inventory event, the operators would need to determine that RPV inventory loss wasoccurring by observing sump and tank level changes. Sump and tank level increases must beevaluated against other potential sources of leakage such as cooling water sources inside thecontainment to ensure they are indicative of RCS leakage.NMP2 Basis Reference(s):1. USAR Section 5.2.52. USAR Section 7.6.1.33. N2-EOP-PC Primary Containment Control4. NEI 99-01 IC CU2Page 146 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:EAL:C -Cold Shutdown / Refueling System Malfunction4 -RCS TemperatureInability to maintain plant in cold shutdownCA4.1 AlertAn UNPLANNED event results in EITHER:RCS temperature > 2001F for > Table C-4 durationORRPV pressure increase > 10 psi due to an UNPLANNED loss of decay heat removalcapabilityTable C-4 RCS Reheat Duration ThresholdsCONTAINMENTRCS Status COSURE DurationCLOSURE StatusINTACT N/A 60 min.*Established 20 min.*Not INTACTNot established 0 min.If an RCS heat removal system is in operation within this timeframe and RCS temperature is being reduced, the EAL is notapplicable.Mode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificSeveral instruments are capable of providing indication of RCS temperature with respect tothe Technical Specification cold shutdown temperature limit (2000F). These include (ref. 2):* Recirc operating -Temperature Recorder B35-R650 at P602:o Loop A: Channel 1, RCS LOOP A SUCTIONo Loop B: Channel 6, RCS LOOP B SUCTION" Shutdown cooling operating -Temperature Recorder El 2-R601 at P601Page 147EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Baseso Loop A: Point 1, RHR INLET TO HX Ao Loop B: Point 2, RHR INLET TO HX BIf Rx Recirc or Shutdown Cooling pumps are not in operation and reactor coolanttemperature is greater than or equal to 2120F, RCS temperature can be obtained byconverting the RPV pressure to temperature using the saturated steam tables.If RCS temperature exceeds 2000F, an operating mode change occurs. Although the eventmay have originated in cold conditions, the emergency classification shall be based on theoperating mode that existed at the time the event occurred (prior to any protective systemor operator action initiated in response to the condition). For events that occur in ColdShutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for modeapplicability, even if Hot Shutdown (or a higher mode) is entered during any subsequentheat-up. In particular, the fission product barrier EALs are applicable only to events thatinitiate in Hot Shutdown or higher.The RCS should be considered INTACT when the RCS pressure boundary is in its normalcondition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).CONTAINMENT CLOSURE is the procedurally defined actions taken to securecontainment (primary or secondary) and its associated structures, systems, andcomponents as a functional barrier to fission product release under existing plantconditions. This definition is less restrictive than Technical Specification criteria governingPrimary and Secondary Containment operability. If the Technical Specification criteria aremet, therefore, CONTAINMENT CLOSURE has been established. (ref. 3, 4, 5)The pressure rise of greater than 10 psig infers an RCS temperature in excess of theTechnical Specification cold shutdown limit (2000F) for which this EAL would otherwisepermit up to sixty minutes to restore RCS cooling before declaration of an Alert (RCSINTACT). This EAL therefore covers situations in which it is determined that, due to highdecay heat loads, the time provided to reestablish temperature control should be less thansixty minutes (as indicated by significant RCS re-pressurization).Wide range pressure indication (0-1200 psig) is capable of measuring pressure changes of10 psig (ref. 6).Page 148 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesIf RCS temperature exceeds 200'F, an operating mode change occurs. Although the eventmay have originated in cold conditions, the emergency classification shall be based on theoperating mode that existed at the time the event occurred (prior to any protective systemor operator action initiated in response to the condition). For events that occur in ColdShutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for modeapplicability, even if Hot Shutdown (or a higher mode) is entered during any subsequentheat-up. In particular, the fission product barrier EALs are applicable only to events thatinitiate in Hot Shutdown or higher.Escalation to a Site Area Emergency would be under EAL CS3.1 should boiling result insignificant RPV water level loss leading to core uncovery.GenericThe RCS Reheat Duration Thresholds table addresses complete loss of functions required for corecooling for greater than 60 minutes during refuel and cold shutdown modes when RCS integrity isestablished. The 60 minute time frame should allow sufficient time to restore cooling without therebeing a substantial degradation in plant safety.The RCS Reheat Duration Thresholds table also addresses the complete loss of functions requiredfor core cooling for greater than 20 minutes during Refuel and cold shutdown modes whenCONTAINMENT CLOSURE is established but RCS integrity is not established. The allowed 20minute time frame was included to allow operator action to restore the heat removal function, ifpossible.Finally, complete loss of functions required for core cooling during Refuel and cold shutdownmodes when neither CONTAINMENT CLOSURE nor RCS integrity are established is addressed.No delay time is allowed because the evaporated reactor coolant that may be released into theContainment during this heatup condition could also be directly released to the environment.The note (*) indicates that this EAL is not applicable if actions are successful in restoring an RCSheat removal system to operation and RCS temperature is being reduced within the specified timeframe.The 10 psig pressure increase addresses situations where, due to high decay heat loads, the timeprovided to restore temperature control, should be less than 60 minutes. The RPV pressuresetpoint was chosen because it is the lowest pressure that the site can read on installed ControlBoard instrumentation that is equal to or greater than 10 psig.Escalation to Site Area Emergency would be via EAL CS3.1 should boiling result in significant RPVlevel loss leading to core uncovery.A loss of Technical Specification components alone is not intended to constitute an Alert. Thesame is true of a momentary UNPLANNED excursion above the Technical Specification coldshutdown temperature limit when the heat removal function is available.The Emergency Director must remain alert to events or conditions that lead to the conclusion thatPage 149 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesexceeding the EAL is IMMINENT. If, in the judgment of the Emergency Director, an IMMINENTsituation is at hand, the classification should be made as if the threshold has been exceeded.NMP2 Basis Reference(s):1. Technical Specifications Table 1.1-12. N2-OSP-RCS-@001 RCS Pressure/Temperature Verification3. NIP-OUT-01 Shutdown Safety4. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.1.15. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.6.4.16. N2-OP-34 Nuclear Boiler, Automatic Depressurization and Safety Relief Valves,Attachment 17. NEI 99-01 IC CA4Page 150EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 4- RCS TemperatureInitiating Condition: UNPLANNED loss of decay heat removal capabilityEAL:CU4.1 Unusual EventUNPLANNED event results in RCS temperature > 200OFMode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificSeveral instruments are capable of providing indication of RCS temperature with respect tothe Technical Specification cold shutdown temperature limit (2001F). These include (ref. 2):" Recirc operating -Temperature Recorder B35-R650 at P602:o Loop A: Channel 1, RCS LOOP A SUCTIONo Loop B: Channel 6, RCS LOOP B SUCTION" Shutdown cooling operating -Temperature Recorder E12-R601 at P601o Loop A: Point 1, RHR INLET TO HX Ao Loop B: Point 2, RHR INLET TO HX BIf Rx Recirc or Shutdown Cooling pumps are not in operation and reactor coolanttemperature is greater than or equal to 212'F, RCS temperature can be obtained byconverting the RPV pressure to temperature using the saturated steam tables.If RCS temperature exceeds 200'F, an operating mode change occurs. Although the eventmay have originated in cold conditions, the emergency classification shall be based on theoperating mode that existed at the time the event occurred (prior to any protective systemor operator action initiated in response to the condition). For events that occur in ColdShutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for modeapplicability, even if Hot Shutdown (or a higher mode) is entered during any subsequentheat-up. In particular, the fission product barrier EALs are applicable only to events thatPage 151 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesinitiate in Hot Shutdown or higher.GenericThis EAL is a precursor of more serious conditions and, as a result, is considered to be a potentialdegradation of the level of safety of the plant. In cold shutdown the ability to remove decay heatrelies primarily on forced cooling flow. Operation of the systems that provide this forced coolingmay be jeopardized due to the unlikely loss of electrical power or RCS inventory. Since the RCSusually remains INTACT in the cold shutdown mode a large inventory of water is available to keepthe core covered.During refueling the level in the RPV will normally be maintained above the RPV flange. Refuelingevolutions that decrease water level below the RPV flange are carefully planned and procedurallycontrolled. Loss of forced decay heat removal at reduced inventory may result in more rapidincreases in RCS/RPV temperatures depending on the time since shutdown.Normal means of core temperature indication and RPV water level indication may not be availablein the Refuel mode. Redundant means of RPV water level indication are therefore procedurallyinstalled to assure that the ability to monitor level will not be interrupted. Escalation to Alert wouldbe via EAL CA3.1 based on an inventory loss or EAL CA4.1 based on exceeding its temperatureduration or pressure criteria.NMP2 Basis Reference(s):1. Technical Specifications Table 1.1-12. N2-OSP-RCS-@001 RCS Pressure/Temperature Verification3. NEI 99-01 IC CU4Page 152EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 4 -RCS TemperatureInitiating Condition: UNPLANNED loss of decay heat removal capabilityEAL:CU4.2 Unusual EventLoss of all RCS temperature and RPV water level indication for >_ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeMode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificSeveral instruments are capable of providing indication of RCS temperature with respect tothe Technical Specification cold shutdown temperature limit (2000F). These include (ref. 2):* Recirc operating -Temperature Recorder B35-R650 at P602:o Loop A: Channel 1, RCS LOOP A SUCTIONo Loop B: Channel 6, RCS LOOP B SUCTION* Shutdown cooling operating -Temperature Recorder E12-R601 at P601o Loop A: Point 1, RHR INLET TO HX Ao Loop B: Point 2, RHR INLET TO HX BIf Rx Recirc or Shutdown Cooling pumps are not in operation and reactor coolanttemperature is greater than or equal to 2120F, RCS temperature can be obtained byconverting the RPV pressure to temperature using the saturated steam tables.RPV water level is monitored from -165 in. to +545 in. to ensure adequate coverage forexpected and postulated conditions of RPV water level. RPV water level measurement isderived by the differential pressure that exists between a reference leg and variable leg. Alllevel instruments are referenced to an "instrument zero", which is 380.69 inches above"vessel zero". The instrument zero is the top of the reactor vessel upper grid (top guide).Page 153 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesRPV water level monitoring is subdivided into five ranges identified as:* Narrow provides indication and control signals for normal plant operation andprotection system actuation." Wide provides indication and control signals for transient conditions below thenormal operating band and emergency equipment actuation.* Upset provides indication for transient conditions above normal operating band.* Shutdown provides indication for vessel flood up and activities.* Fuel Zone provides indication for long term accident conditions where reactor levelcannot be restored.The shutdown range level indication is utilized during cold reactor startup and vessel floodup for refueling. The shutdown range instrument uses a single level transmitter(21SC*LT105) to provide an input to a level indicator on 2CES*PNL851 (Computer PointA486). (ref. 3)Although the event may have originated in cold conditions, the emergency classificationshall be based on the operating mode that existed at the time the event occurred (prior toany protective system or operator action initiated in response to the condition). For eventsthat occur in Cold Shutdown or Refuel, escalation is via EALs that have Cold Shutdown orRefuel for mode applicability, even if Hot Shutdown (or a higher mode) is entered duringany subsequent heat-up. In particular, the fission product barrier EALs are applicable onlyto events that initiate in Hot Shutdown or higher.GenericThis EAL is a precursor of more serious conditions and, as a result, is considered to be a potentialdegradation of the level of safety of the plant. In cold shutdown the ability to remove decay heatrelies primarily on forced cooling flow. Operation of the systems that provide this forced coolingmay be jeopardized due to the unlikely loss of electrical power or RPV inventory. Since the RCSusually remains INTACT in the cold shutdown mode a large inventory of water is available to keepthe core covered.During refueling the level in the RPV will normally be maintained above the RPV flange. Refuelingevolutions that decrease water level below the RPV flange are carefully planned and procedurallycontrolled. Loss of forced decay heat removal at reduced inventory may result in more rapidincreases in RPV temperatures depending on the time since shutdown.Normal means of core temperature indication and RPV water level indication may not be availablePage 154 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesin the Refuel mode. Redundant means of RPV water level indication are therefore procedurallyinstalled to assure that the ability to monitor level will not be interrupted. However, if all level andtemperature indication were to be lost in either the cold shutdown of refueling modes, this EALwould result in declaration of a UE if both temperature and level indication cannot be restoredwithin 15 minutes from the loss of both means of indication. Escalation to Alert would be via EALCA3.1 based on an inventory loss or EAL CA4.1 based on exceeding its temperature criteria.NMP2 Basis Reference(s):1. Technical Specifications Table 1.1-12. N2-OSP-RCS-@001 RCS Pressure/Temperature Verification3. NIP-OUT-01 Shutdown Safety4. NEI 99-01 IC CU4Page 155EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: C -Cold Shutdown / Refueling System MalfunctionSubcategory: 5 -Inadvertent CriticalityInitiating Condition: Inadvertent criticalityEAL:CU5.1 Unusual EventAn UNPLANNED sustained positive period observed on nuclear instrumentationMode Applicability:4 -Cold Shutdown, 5 -RefuelBasis:Plant-SpecificThe term "sustained" is used to allow exclusion of expected short-term positive periodsfrom planned fuel bundle or control rod movements during core alteration. These short-term positive periods are the result of the rise in neutron population due to subcriticalmultiplication.GenericThis EAL addresses criticality events that occur in Cold Shutdown or Refueling modes such as fuelmis-loading events and inadvertent dilution events. This EAL indicates a potential degradation ofthe level of safety of the plant, warranting a UE classification.Escalation would be by Emergency Director judgment.NMP2 Basis Reference(s):1. NEI 99-01 IC CU8Page 156 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesC -Cold Shutdown / Refueling System MalfunctionCategory:Subcategory: 6 -CommunicationsInitiating Condition: Loss of all onsite or offsite communications capabilitiesEAL:CU6.1 Unusual EventLoss of all Table C-5 onsite (internal) communication methods affecting the ability toperform routine operationsORLoss of all Table C-5 offsite (external) communication methods affecting the ability toperform offsite notificationsTable C-5 Communications SystemsSystem Onsite Offsite(internal) (external)PBX (normal dial telephones)GaitronicsStation radio (portable)Control Room installed satellite phones (non portable)ENSRECSXXXXXXXMode Applicability:4 -Cold Shutdown, 5 -Refuel, D -DefueledBasis:Plant-SpecificOnsite/offsite communications systems are listed in Table C-2 (ref., 1, 2, 3).This EAL is the cold condition equivalent of the hot condition EAL SU6.1.GenericThe purpose of this EAL is to recognize a loss of communications capability that either defeats theplant operations staff ability to perform routine tasks necessary for plant operations or the ability toPage 157EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basescommunicate issues with off-site authorities. The loss of off-site communications ability is expectedto be significantly more comprehensive than the condition addressed by 10 CFR 50.72.The availability of one method of ordinary off-site communications is sufficient to inform federal,state, and local authorities of plant issues. This EAL is intended to be used only whenextraordinary means (e.g., relaying of information from radio transmissions, individuals being sentto off-site locations, etc.) are being utilized to make communications possible.NMP2 Basis Reference(s):1. USAR Section 9.5.22. Nine Mile Point Nuclear Station Site Emergency Plan, Section 7.23. N2-OP-76 Plant Communications4. NEI 99-01 IC CU6Page 158 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory S -System MalfunctionEAL Group: Hot Conditions (RCS temperature > 2000F);EALs in this category are applicable only inone or more hot operating modes.Numerous system-related equipment failure events that warrant emergency classificationhave been identified in this category. They may pose actual or potential threats to plantsafety.The events of this category pertain to the following subcategories:1. Loss of AC PowerLoss of emergency plant electrical power can compromise plant safety systemoperability including decay heat removal and emergency core cooling systems whichmay be necessary to ensure fission product barrier integrity. This category includesloss of onsite and offsite power sources for the 4.16KV emergency buses.2. Loss of DC PowerLoss of emergency plant electrical power can compromise plant safety systemoperability including decay heat removal and emergency core cooling systems whichmay be necessary to ensure fission product barrier integrity. This category includesloss of power to the 125 VDC buses.3. Criticality & RPS FailureInadvertent criticalities pose potential personnel safety hazards as well being indicativeof losses of reactivity control.Events related to failure of the Reactor Protection System (RPS) to initiate andcomplete reactor scrams. In the plant licensing basis, postulated failures of the RPS tocomplete a reactor scram comprise a specific set of analyzed events referred to asAnticipated Transient Without Scram (ATWS) events. For EAL classification however,ATWS is intended to mean any scram failure event that does not achieve reactorshutdown. If RPS actuation fails to assure reactor shutdown, positive control ofreactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.4. Inability to Reach or Maintain Shutdown ConditionsPage 159 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesSystem malfunctions may lead to failure of the plant to be brought to the required plantoperating condition required by technical specifications if a limiting condition foroperation (LCO) is not met.5. InstrumentationCertain events that degrade plant operator ability to effectively assess plant conditionswithin the plant warrant emergency classification. Losses of annunciators are in thissubcategory.6. CommunicationsCertain events that degrade plant operator ability to effectively communicate withessential personnel within or external to the plant warrant emergency classification.7. Fuel Clad DegradationDuring normal operation, reactor coolant fission product activity is very low. Smallconcentrations of fission products in the coolant are primarily from the fission of trampuranium in the fuel clad or minor perforations in the clad itself. Any significant increasefrom these base-line levels (-5% clad failures) is indicative of fuel failures and iscovered under Category F, Fission Product Barrier Degradation. However, lesseramounts of clad damage may result in coolant activity exceeding TechnicalSpecification limits. These fission products will be circulated with the reactor coolantand can be detected by coolant sampling and/or the Letdown radiation monitor.8. RCS LeakageThe RPV provides a volume for the coolant that covers the reactor core. The RPV andassociated pressure piping (reactor coolant system) together provide a barrier to limitthe release of radioactive material should the reactor fuel clad integrity fail.Excessive RCS leakage greater than Technical Specification limits are utilized toindicate potential pipe cracks that may propagate to an extent threatening fuel clad,RCS and containment integrity.Page 160 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 1 -Loss of PowerInitiating Condition: Prolonged loss of all offsite and all onsite AC power to 4.16 KVemergency busesEAL:SGI.1 General EmergencyLoss of all offsite and all onsite AC power, Table S-1, to 4.16 KV emergency buses2ENS*SWG101 and 2ENS *SWG103AND EITHER:Restoration of 4.16 KV emergency bus 2ENS*SWG101 or 2ENS *SWG103 within 4hours is not likelyORRPV water level cannot be restored and maintained above -14 in. or RPV waterlevel cannot be determinedTable S-1 AC Power Sources0 2EGS*EG1W

• 2EGS*EG3U/).r_ 0 2EGS*EG2 (with 2ENS*SWG102crosstied to 2ENS*SWG101 or2ENS*SWG103)0 Reserve Transformer A0 Reserve Transformer B0* Aux Boiler TransformerMode Applicability:1 -Power Operation,2 -Startup, 3 -Hot ShutdownBasis:Plant-Specific2ENS*SWG101, *SWG102, and *SWG103 are the 4.16 KV emergency buses. Bus2ENS*SWG101 is dedicated to Division I of the On-site Emergency AC ElectricalDistribution System, bus 2ENS*SWG102 is dedicated to Division III (HPCS), and bus2ENS*SWG103 is dedicated to Division II. Buses 2ENS*SWG101 and *SWG103 feed allPage 161EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesStation redundant safety-related loads, except the HPCS system loads. The HPCS systemloads are fed by bus 2ENS*SWG1O2 (ref. 1, 2)." All three divisions are normally energized by the On-site Normal AC ElectricalDistribution System via the off-site power sources through the reserve stationservice transformers 2RTX-XSRIA and 2RTX-XSR1B.o 2ENS*SWG102 from transformer 2RTX-XSR1Ao 2ENS*SWG103 from transformer 2RTX-XSR1B." Buses 2ENS*SWG101 and *SWG103 each have a backup source, the AuxiliaryBoiler Transformer 2ABS-X1. Also, 2ENS*SWG101 and *SWG103 each have afeeder to a normal AC (stub) bus, NNS-SWG014 and NNS-SWG015 respectively." Bus 2ENS*SWG102 has a backup connection to the Reserve Station ServiceTransformer 2RTX-XSR1 B, if required." Each of the three 4.16 KV emergency buses has a standby diesel generator(2EGS*EG1, 2EGS*EG3, 2EGS*EG2) to carry its loads in case of a LOOP or incase of a sustained degraded voltage condition on the offsite source (ref. 3, 4).2EGS*EG2 (Division Ill) is capable of powering either the Division I or Division II4.16 KV emergency bus through manual breaker alignments. The availability of2EGS*EG2 as an onsite AC power source in Table S-1 only applies if 2EGS*EG2 isaligned to energize 2ENS*SWG101 or 2ENS*SWG103.Consideration should be given to operable loads necessary to remove decay heat orprovide RPV makeup capability when evaluating loss of all AC power to vital buses. Eventhough an essential bus may be energized, if necessary loads (i.e., loads that if lost wouldinhibit decay heat removal capability or RPV makeup capability) are not operable on theenergized bus then the bus should not be considered operable.Four hours is the station blackout coping period (ref. 4, 5).An RPV water level instrument reading of -14 in. indicates RPV water level is at the top ofactive fuel. When RPV water level is at or above the top of active fuel, the core iscompletely submerged. Core submergence is the most desirable means of core cooling.When RPV water level is below the top of active fuel, the uncovered portion of the corePage 162 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesmust be cooled by less reliable means (i.e., steam cooling or spray cooling). If coreuncovery is threatened, the EOPs specify alternate, more extreme, RPV water level controlmeasures in order to restore and maintain adequate core cooling (ref. 6). Since coreuncovery begins if RPV water level drops to -14 in., the level is indicative of a challenge tocore cooling and the Fuel Clad barrier.Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that RPV water level cannot be restored and maintained above the top ofactive fuel may be made at, before, or after RPV water level actually decreases to thispoint. (ref. 6)When RPV water level cannot be determined, EOPs require RPV flooding strategies. RPVwater level indication provides the primary means of knowing if adequate core cooling isbeing maintained. When all means of determining RPV water level are unavailable, thefuel clad barrier is threatened and reliance on alternate means of assuring adequate corecooling must be attempted. The instructions in EOP-C4 specify these means, whichinclude emergency depressurization of the RPV and injection into the RPV at a rateneeded to flood to the elevation of the main steam lines or hold RPV pressure above theMinimum Steam Cooling Pressure (in ATWS events). (ref. 7) If RPV water level cannot bedetermined with respect to the top of active fuel, a potential loss of the Fuel Clad barrierexists.Note that EOP-C5 may require intentional uncovery of the core and control of RPV waterlevel between -14 in. and -39 in., the Minimum Steam Cooling RPV Water Level(MSCRWL) (ref. 8). Under these conditions, a high-power ATWS event exists and requiresat least a Site Area Emergency classification in accordance with the ATWS/CriticalityEALs.GenericLoss of all AC power to emergency busses compromises all plant safety systems requiring electricpower including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolongedloss of all AC power to emergency buses will lead to loss of fuel clad, RCS, and containment, thuswarranting declaration of a General Emergency.This EAL is specified to assure that-in the unlikely event of a prolonged loss of all AC power to 4.16KV emergency buses, timely recognition of the seriousness of the event occurs and thatdeclaration of a General Emergency occurs as early as is appropriate, based on a reasonablePage 163 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesassessment of the event trajectory.The likelihood of restoring at least one emergency bus should be based on a realistic appraisal ofthe situation since a delay in an upgrade decision based on only a chance of mitigating the eventcould result in a loss of valuable time in preparing and implementing public protective actions.In addition, under these conditions, fission product barrier monitoring capability may be degraded.NMP2 Basis Reference(s):1. USAR Section 8.22. USAR Section 8.33. N2-SOP-03 Loss of AC Power4. N2-SOP-01 Station Blackout5. USAR Section 8.3.1.5.26. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document7. N2-EOP-C4 RPV Flooding8. N2-EOP-C5 Failure to Scram9. NEI 99-01 IC SG1Page 164 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:S -System Malfunction1 -Loss of AC PowerLoss of all offsite and all onsite AC power to 4.16 KV emergencybuses for 15 min.EAL:SSI.1 Site Area EmergencyLoss of all offsite and all onsite AC power, Table S-1, to 4.16 KV emergency buses2ENS*SWG101 and 2ENS*SWG103 for _ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table S-1 AC Power Sources0 2EGS*EG10 2EGS*EG3" 2EGS*EG2 (with 2ENS*SWG1020 crosstied to 2ENS*SWG101 or2ENS*SWG103)0 Reserve TransformerA" Reserve Transformer B* Aux Boiler TransformerMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-Specific2ENS*SWG101, *SWG102, and *SWG103 are the 4.16 KV emergency buses. Bus2ENS*SWG1 01 is dedicated to Division I of the On-site Emergency AC ElectricalDistribution System, bus 2ENS*SWG102 is dedicated to Division III (HPCS), and bus2ENS*SWG103 is dedicated to Division I1. Buses 2ENS*SWG101 and *SWG103 feed allStation redundant safety-related loads, except the HPCS system loads. The HPCS systemloads are fed by bus 2ENS*SWG102 (ref. 1, 2).* All three divisions are normally energized by the On-site Normal AC ElectricalPage 165 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesDistribution System via the off-site power sources through the reserve stationservice transformers 2RTX-XSRIA and 2RTX-XSR1 B.o 2ENS*SWG102 from transformer 2RTX-XSRIAo 2ENS*SWG103 from transformer 2RTX-XSR1B." Buses 2ENS*SWG101 and *SWG103 each have a backup source, the AuxiliaryBoiler Transformer 2ABS-Xl. Also, 2ENS*SWG101 and *SWG103 each have afeeder to a normal AC (stub) bus, NNS-SWG014 and NNS-SWG015 respectively.* Bus 2ENS*SWG102 has a backup connection to the Reserve Station ServiceTransformer 2RTX-XSR1 B, if required." Each of the three 4.16 KV emergency buses has a standby diesel generator(2EGS*EG1, 2EGS*EG3, 2EGS*EG2) to carry its loads in case of a LOOP or incase of a sustained degraded voltage condition on the offsite source (ref. 3, 4).2EGS*EG2 (Division Ill) is capable of powering either the Division I or Division II4.16 KV emergency bus through manual breaker alignments. It is unlikely that theseactions could be performed within the fifteen-minute interval of this EAL. Theavailability of 2EGS*EG2 as an onsite AC power source in Table S-1 only applies if2EGS*EG2 is aligned to energize 2ENS*SWG101 or 2ENS*SWG103.Consideration should be given to operable loads necessary to remove decay heat orprovide RPV makeup capability when evaluating loss of all AC power to vital buses. Eventhough an essential bus may be energized, if necessary loads (i.e., loads that if lost wouldinhibit decay heat removal capability or RPV makeup capability) are not operable on theenergized bus then the bus should not be considered operable.The fifteen-minute interval was selected as a threshold to exclude transient power losses.GenericLoss of all AC power to emergency busses compromises all plant safety systems requiring electricpower including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolongedloss of all AC power to 4.16 KV emergency buses will lead to loss of Fuel Clad, RCS, andContainment, thus this event can escalate to a General Emergency.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-sitepower.Escalation to General Emergency is via EALs in Category F or EAL SG1.1.Page 166 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesNMP2 Basis Reference(s):1. USAR Section 8.22. USAR Section 8.33. N2-SOP-03 Loss of AC Power4. N2-SOP-01 Station Blackout5. NEI 99-01 IC SS1Page 167EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 1 -Loss of AC PowerInitiating Condition: AC power capability to 4.16 KV emergency buses reduced to asingle power source for __15 min. such that ANY additional singlefailure would result in a complete loss of all 4.16 KV emergencybus powerEAL:SAI.1 AlertAC power capability to 4.16 KV emergency buses 2ENS*SWG101 and 2ENS*SWG103reduced to a single power source, Table S-1, for >_ 15 min. (Note 4)ANDANY additional single power source failure will result in a loss of all power to 4.16 KVemergency buses 2ENS*SWG101 and 2ENS*SWG103Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Table S-1 AC Power Sources* 2EGS*EG1* 2EGS*EG3* 2EGS*EG2 (with 2ENS*SWG1020 crosstied to 2ENS*SWG101 or2ENS*SWG103)* Reserve Transformer A* Reserve Transformer B* Aux Boiler TransformerMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-Specific2ENS*SWG101, *SWG102, and *SWG103 are the 4.16 KV emergency buses. Bus2ENS*SWG101 is dedicated to Division I of the On-site Emergency AC ElectricalDistribution System, bus 2ENS*SWG102 is dedicated to Division III (HPCS), and busPage 168 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases2ENS*SWG102 is dedicated to Division II. Buses 2ENS*SWGi01 and *SWG103 feed allStation redundant safety-related loads, except the HPCS system loads. The HPCS systemloads are fed by bus 2ENS*SWG102 (ref. 1, 2)." All three divisions are normally energized by the On-site Normal AC ElectricalDistribution System via the off-site power sources through the reserve stationservice transformers 2RTX-XSR1A and 2RTX-XSRIB.o 2ENS*SWG102 from transformer 2RTX-XSR1Ao 2ENS*SWG103 from transformer 2RTX-XSRIB." Buses 2ENS*SWG101 and *SWG103 each have a backup source, the AuxiliaryBoiler Transformer 2ABS-Xl. Also, 2ENS*SWG101 and *SWG103 each have afeeder to a normal AC (stub) bus, NNS-SWG01 4 and NNS-SWG01 5 respectively.* Bus 2ENS*SWG102 has a backup connection to the Reserve Station ServiceTransformer 2RTX-XSR1 B, if required.* Each of the three 4.16 KV emergency buses has a standby diesel generator(2EGS*EG1, 2EGS*EG3, 2EGS*EG2) to carry its loads in case of a LOOP or incase of a sustained degraded voltage condition on the offsite source (ref. 3, 4).2EGS*EG2 (Division Ill) is capable of powering either the Division I or Division II4.16 KV emergency bus through manual breaker alignments. It is unlikely that theseactions could be performed within the fifteen-minute interval of this EAL. Theavailability of 2EGS*EG2 as an onsite AC power source in Table S-1 only applies if2EGS*EG2 is aligned to energize 2ENS*SWG101 or 2ENS*SWG103.The fifteen-minute interval was selected as a threshold to exclude transient power losses.If the capability for multiple sources to energize the unit vital buses within 15 minutes is notrestored, an Alert is declared under this EAL. The subsequent loss of the single remainingpower source escalates the event to a Site Area Emergency'under EAL SS1.1.GenericThe condition indicated by this EAL is the degradation of the off-site and on-site AC power systemssuch that any additional single failure would result in a complete loss of 4.16 KV emergency busAC power to one or both units. This condition could occur due to a loss of off-site power with aconcurrent failure of all but one emergency generator to supply power to its emergency buses.Page 169 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesAnother related condition could be the loss of all off-site power and loss of on-site emergencygenerators with only one train of 4.16 KV emergency buses being backfed from the unit maingenerator, or the loss of on-site emergency generators with only one train of 4.16 KV emergencybuses being backfed from off-site power. The subsequent loss of this single power source wouldescalate the event to a Site Area Emergency in accordance with EAL SS1.1.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.NMP2 Basis Reference(s):1. USAR Section 8.22. USAR Section 8.33. N2-SOP-03 Loss of AC Power4. N2-SOP-01 Station Blackout5. NEI 99-01 IC SA5Page 170 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 1 -Loss of AC PowerInitiating Condition: Loss of all offsite AC power to 4.16KV vital buses for __ 15 min.EAL:SUI.I Unusual EventLoss of all offsite AC power, Table S-1, to 4.16 KV emergency buses 2ENS*SWG101 and2ENS*SWG103Table S-1 AC Power Sources0 2EGS*EG1* 2EGS*EG3O 2EGS*EG2 (with 2ENS*SWG1020 crosstied to 2ENS*SWG101 or2ENS*SWG103)* Reserve Transformer A"O e Reserve Transformer B0 0 Aux Boiler TransformerMode Applicability:1 -Power Operation, 2 -Basis:Plant-SpecificStartup, 3 -Hot Shutdown2ENS*SWG101, *SWG102, and 2ENS*SWG103 are the 4.16 KV emergency buses. Bus2ENS*SWG101 is dedicated to Division I of the On-site Emergency AC ElectricalDistribution System, bus 2ENS*SWG102 is dedicated to Division III (HPCS), and bus2ENS*SWG103 is dedicated to Division II. Buses 2ENS*SWG101 and *SWG103 feed allStation redundant safety-related loads, except the HPCS system loads. The HPCS systemloads are fed by bus 2ENS*SWG102 (ref. 1, 2).All three divisions are normally energized by the On-site Normal AC ElectricalDistribution System via the off-site power sources through the reserve stationservice transformers 2RTX-XSR1A and 2RTX-XSRIB.Page 171EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical Baseso 2ENS*SWG102 from transformer 2RTX-XSR1Ao 2ENS*SWG103 from transformer 2RTX-XSR1B." Buses 2ENS*SWG101 and *SWG103 each have a backup source, the AuxiliaryBoiler Transformer 2ABS-XI. Also, 2ENS*SWG101 and *SWG103 each have afeeder to a normal AC (stub) bus, NNS-SWG014 and NNS-SWG015 respectively." Bus 2ENS*SWG1 02 has a backup connection to the Reserve Station ServiceTransformer 2RTX-XSR1 B, if required." Each of the three 4.16 KV emergency buses has a standby diesel generator(2EGS*EG1, 2EGS*EG3, 2EGS*EG2) to carry its loads in case of a LOOP or incase of a sustained degraded voltage condition on the offsite source (ref. 3, 4).2EGS*EG2 (Division Ill) is capable of powering either the Division I or Division II4.16 KV emergency bus through manual breaker alignments. It is unlikely that theseactions could be performed within the fifteen-minute interval of this EAL. Theavailability of 2EGS*EG2 as an onsite AC power source in Table S-1 only applies if2EGS*EG2 is aligned to energize 2ENS*SWG101 or 2ENS*SWG103.The NMP2 electrical distribution configuration precludes restoration of offsite powersources within 15 minutes in all instances, once lost. Therefore no time component isallocated for this EAL threshold.GenericProlonged loss of off-site AC power reduces required redundancy and potentially degrades thelevel of safety of the plant by rendering the plant more vulnerable to a complete loss of AC powerto emergency busses.NMP2 Basis Reference(s):1. USAR Section 8.22. USAR Section 8.33. N2-SOP-03 Loss of AC Power4. N2-SOP-01 Station Blackout5. NEI 99-01 IC SU1Page 172 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 2 -Loss of DC PowerInitiating Condition: Loss of all emergency DC power for __ 15 min.EAL:SS2.1 Site Area Emergency< 105 VDC on both 2BYS*SWG002A and 2BYS*SWG002B for _ 15 min. (Note 4)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable time.Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificThe emergency 125 VDC power system includes three electrically independent andseparate switchgears (2BYS*SWG002A, 2BYS*SWG002B and 2CES*IPNL414). Division I((2BYS*SWG002A) and Division II (2BYS*SWG002B) feed the redundant emergency DCloads associated with Divisions I and II of the emergency onsite AC system, respectively.Division III (2CES*PNP414) feeds the emergency DC loads associated with Division III(HPCS system). 2CES*IPNL414 is not included in this EAL because it only supplies powerto HPCS loads.Each emergency 125 VDC distribution system has a battery and a battery charger that arenormally connected to the bus such that these two sources of power are operating inparallel. The charger is normally supplying system electrical loads with the battery on afloat charge. Should both battery chargers for any particular battery be out of service atany point in the DC load cycle, the battery is capable of starting and operating itsassociated loads for 2 hr according to a precalculated load profile without the batteryterminal voltage falling below minimum acceptable level, 105 VDC. (ref. 1, 2, 3)This EAL is the hot condition equivalent of the cold condition loss of DC powerEAL CU2.1.GenericLoss of all DC power compromises ability to monitor and control plant safety functions. ProlongedPage 173 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesloss of all DC power will cause core uncovering and loss of containment integrity when there issignificant decay heat and sensible heat in the reactor system.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Escalation to a General Emergency would occur by EALs in Category R and Category F.NMP2 Basis Reference(s):1. USAR Section 8.3.2.1.22. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.8.43. N2-SOP-04 Loss of DC Power4. NEI 99-01 IC SS3Page 174EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 3 -Criticality & RPS FailureInitiating Condition: Automatic scram and all manual actions fail to shut down thereactor and indication of an extreme challenge to the ability to coolthe core existsEAL:SG3.1 General EmergencyAn automatic scram fails to shut down the reactor as indicated by reactor power > 4%ANDAll manual actions fail to shut down the reactor as indicated by reactor power > 4%AND EITHER of the following exist or have occurred:RPV water level cannot be restored and maintained above -39 in. or RPV waterlevel cannot be determinedORSuppression pool temperature and RPV pressure cannot be maintained below theHeat Capacity Temperature Limit (N2-EOP-PC Figure M)Mode Applicability:1 -Power Operation, 2 -StartupBasis:Plant-SpecificThis EAL addresses the following:" Any automatic reactor scram signal followed by a manual scram that fails to shutdown the reactor to an extent the reactor is producing energy in excess of the heatload for which the safety systems were designed (EAL SS3.1), and.Indications that either core cooling is extremely challenged or heat removal isextremely challenged.Reactor shutdown achieved by use of the alternate control rod insertion methods of EOP-C5 is also credited as a successful manual scram provided reactor power can be reducedbelow the APRM downscale trip setpoint before indications of an extreme challenge toeither core cooling or heat removal exist (ref. 1, 2).The APRM downscale trip setpoint (4%) is a minimum reading on the power range scalePage 175 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesthat indicates power production (ref. 1, 2). It also approximates the decay heat which theshutdown systems were designed to remove and is indicative of a condition requiringimmediate response to prevent subsequent core damage. At or below the APRMdownscale trip setpoint, plant response will be similar to that observed during a normalshutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters(e.g., number of open SRVs, number of open main turbine bypass valves, main steamflow, RPV pressure and suppression pool temperature trend, etc.) can be used todetermine if reactor power is greater than 4% power (ref. 2).The combination of failure of both front line and backup protection systems to function inresponse to a plant transient, along with the continued production of heat, poses a directthreat to the Fuel Clad and RCS barriers.By definition, an operating mode change occurs when the Mode Switch is moved from thestartup/hot standby or run position to the shutdown position. The plant operating mode thatexisted at the time the event occurs (i.e., Power Operation or Startup), however, requiresemergency classification of at least an Alert. The operating mode change associated withmovement of the Mode Switch, by itself, does not justify failure to declare an emergencyfor ATWS events.Indication that core cooling is extremely challenged is manifested by:RPV level cannot be restored and maintained above -39 in. (ref. 1, 2). The MinimumSteam Cooling RPV Water Level (MSCRWL) is the lowest RPV water level at whichthe covered portion of the reactor core will generate sufficient steam to preclude anyclad temperature in the uncovered portion of the core from exceeding 1500'F.Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that RPV level cannot be restored and maintained above theMSCRWL may be made at, before, or after RPV level actually decreases to thispoint.When RPV water level cannot be determined, EOPs require RPV floodingstrategies. RPV water level indication provides the primary means of knowing ifadequate core cooling is being maintained. When all means of determining RPVwater level are unavailable, the fuel clad barrier is threatened and reliance onPage 176 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesalternate means of assuring adequate core cooling must be attempted. Theinstructions in N2-EOP-C4 specify these means, which include emergencydepressurization of the RPV and injection into the RPV at a rate needed to flood tothe elevation of the main steam lines or hold RPV pressure above the MinimumSteam Cooling Pressure (in ATWS events) (ref. 3).The HCTL is the highest wetwell temperature from which emergency RPVdepressurization will not raise:o Suppression chamber temperature above the design value (270'F), oro Suppression chamber pressure above Primary Containment Pressure Limitbefore the rate of energy transfer from the RPV to the containment is greaterthan the capacity of the containment vent.The HCTL is a function of RPV pressure and suppression pool water level. It isutilized to preclude failure of the containment and equipment in the containmentnecessary for the safe shutdown of the plant. Plant parameters in excess of theHCTL could be a precursor of primary containment failure. (ref. 2)The HCTL is given in N2-EOP-PC Figure M. This threshold is met when RPVBLOW DOWN is required in N2-EOP-PC, Step SPT-6 (ref. 4). This conditionaddresses loss of functions required for hot shutdown with the reactor at pressureand temperature.GenericUnder these conditions, the reactor is producing more heat than the maximum decay heat load forwhich the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful.The reactor should be considered shutdown when it producing less heat than the maximum decayheat load for which the safety systems are designed (4% power). In the event either of thesechallenges exists at a time that the reactor has not been brought below the power associated withthe safety system design a core melt sequence exists. In this situation, core degradation can occurrapidly. For this reason, the General Emergency declaration is intended to be anticipatory of thefission product barrier table declaration to permit maximum off-site intervention time.NMP2 Basis Reference(s):1. N2-EOP-C5 Failure to Scram2. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document3. N2-EOP-C4 RPV FloodingPage 177 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Bases4. N2-EOP-PC Primary Containment Control5. NEI 99-01 IC SG2Page 178EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 3 -Criticality & RPS FailureInitiating Condition: Automatic scram fails to shut down the reactor and manual actionstaken from the reactor control console are not successful inshutting down the reactorEAL:SS3.1 Site Area EmergencyAn automatic scram failed to shut down the reactor as indicated by reactor power > 4%ANDManual actions taken at the reactor control console (mode switch in shutdown, manualscram push buttons and ARI) failed to shut down the reactor as indicated by reactor power> 4%Mode Applicability:1 -Power Operation, 2 -StartupBasis:Plant-SpecificThis EAL addresses any automatic reactor scram signal followed by a manual scram thatfailed to shut down the reactor to an extent the reactor is producing energy in excess of theheat load for which the safety systems were designed.For the purposes of emergency classification at the Site Area Emergency level, successfulmanual scram actions are those which can be quickly performed from the reactor controlconsole (i.e., Mode Switch, manual scram pushbuttons and ARI actuation). Reactorshutdown achieved by use of the alternate control rod insertion methods of EOP-C5 doesnot constitute a successful manual scram (ref. 1, 2).The APRM downscale trip setpoint (4%) is a minimum reading on the power range scalethat indicates power production (ref. 1). It also approximates the decay heat which theshutdown systems were designed to remove and is indicative of a condition requiringimmediate response to prevent subsequent core damage. At or below the APRMdownscale trip setpoint, plant response will be similar to that observed during a normalshutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters(e.g., number of open SRVs, number of open main turbine bypass valves, main steamPage 179 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesflow, RPV pressure and wetwell temperature trend, etc.) can be used to determine ifreactor power is greater than 4% power.By definition, an operating mode change occurs when the Mode Switch is moved from thestartup/hot standby or run position to the shutdown position. The plant operating mode thatexisted at the time the event occurs (i.e., Power Operation or Startup), however, requiresemergency classification of at least an Alert. The operating mode change associated withmovement of the Mode Switch, by itself, does not justify failure to declare an emergencyfor ATWS events.Escalation of this event to a General Emergency would be under EAL SG3.1 orEmergency Director judgment.GenericUnder these conditions, the reactor is producing more heat than the maximum decay heat load forwhich the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful.A Site Area Emergency is warranted because conditions exist that lead to IMMINENT loss orpotential loss of both fuel clad and RCS.The reactor should be considered shutdown when it producing less heat than the maximum decayheat load for which the safety systems are designed (4% power).Manual scram actions taken at the reactor control console are any set of actions by the reactoroperator(s) at which causes or should cause control rods to be rapidly inserted into the core andshuts down the reactor.Manual scram actions are not considered successful if action away from the reactor controlconsole is required to scram the reactor. This EAL is still applicable even if actions taken awayfrom the reactor control console are successful in shutting the reactor down because the designlimits of the fuel may have been exceeded or because of the gross failure of the Reactor ProtectionSystem to shutdown the plant.Escalation of this event to a General Emergency would be due to a prolonged condition leading toan extreme challenge to either core-cooling or heat removal.NMP2 Basis Reference(s):1. N2-EOP-C5 Failure to Scram2. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document3. NEI 99-01 IC SS2Page 180 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 3 -Criticality & RPS FailureInitiating Condition: Automatic scram failed to shut down the reactor and the manualactions taken from the reactor control console are successful inshutting down the reactorEAL:SA3.1 AlertAn automatic scram failed to shut down the reactorANDManual actions taken at the reactor control console (mode switch in shutdown, manualscram push buttons or ARI) successfully shut down the reactor as indicated by reactorpower < 4%Mode Applicability:1 -Power Operation, 2 -StartupBasis:Plant-SpecificThe first condition of this EAL identifies the need to cease critical reactor operations byactuation of the automatic Reactor Protection System (RPS) scram function. A reactorscram is automatically initiated by the Reactor Protection System (RPS) when certaincontinuously monitored parameters exceed predetermined setpoints. A reactor scram maybe the result of manual or automatic action in response to various plant conditions (ref. 1):Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclearpower promptly drops to a fraction of the original power level and then decays to a levelseveral decades less with a negative period. The reactor power drop continues untilreactor power reaches the point at which the influence of source neutrons on reactorpower starts to be observable. A predictable post-scram response from an automaticreactor scram signal should therefore consist of a prompt drop in reactor power as sensedby the nuclear instrumentation and a lowering of power into the source range. A successfulscram has therefore occurred when there is sufficient rod insertion from the trip of RPS tobring the reactor power to or below the APRM downscale trip setpoint of 4%. For thepurposes of this EAL, a successful automatic initiation of ARI that reduces reactor power toor below 4% is a not a successful automatic scram. If automatic actuation of ARI hasPage 181 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical Basesoccurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI isa backup means of inserting control rods in the unlikely event that an automatic RPSscram signal exists but the reactor continues to generate significant power. (ref. 2, 3)For the purposes of emergency classification at the Alert level, successful manual scramactions are those which can be quickly performed from the reactor control console (i.e.,mode switch, manual scram pushbuttons, and manual ARI actuation). Reactor shutdownachieved by use of the alternate control rod insertion methods of EOP-C5 does notconstitute a successful manual scram (ref. 2).Following any automatic RPS scram signal EOPs prescribe insertion of redundant manualscram signals to back up the automatic RPS scram function and ensure reactor shutdownis achieved. Even if the first subsequent manual scram signal inserts all control rods to thefull-in position immediately after the initial failure of the automatic scram, the lowest level ofclassification that must be declared is an Alert.If the operator determines the reactor must be scrammed before one of the RPS setpointsis reached, procedures require that the Mode Switch first be placed in the shutdownposition. Although manipulation of the Mode Switch is a manual action, the RPS logictrains are actuated as with an automatic RPS-initiated scram. If reactor power remainsabove the APRM downscale trip setpoint after the Mode Switch is placed in shutdown,RPS has failed and, as a minimum, an Alert emergency declaration is required. Ifsubsequent actuation of the reactor scram pushbuttons and manual initiation of ARI do notreduce reactor power to or below the APRM downscale trip setpoint, a Site AreaEmergency declaration is required under EAL SS3.1.In the event that the operator identifies a reactor scram is IMMINENT and initiates asuccessful manual reactor scram before the automatic scram setpoint is reached, nodeclaration is required. The successful manual scram of the reactor before it reaches itsautomatic scram setpoint or reactor scram signals caused by instrumentation channelfailures do not lead to a potential fission product barrier loss. If manual reactor scramactions fail to reduce reactor power to or below 4%, the event escalates to the Site AreaEmergency under EAL SS3.1.By procedure, operator actions include the initiation of an immediate manual scramPage 182 EPMP-EPP-0102Rev 00 (Draft A)

Attachment I -Emergency Action Level Technical Basesfollowing receipt of an automatic scram signal. 'If there are no clear indications that theautomatic scram failed (such as a time delay following indications that a scram setpointwas exceeded), it may be difficult to determine if the reactor was shut down because ofautomatic scram or manual actions. If a subsequent review of the scram actuationindications reveals that the automatic scram did not cause the reactor to be shut down,consideration should be given to evaluating the fuel for potential damage and the reportingrequirements of 50.72 should be considered for the transient event.By definition, an operating mode change occurs when the Mode Switch is moved from thestartup/hot standby or run position to the shutdown position. The plant operating mode thatexisted at the time the event occurs (i.e., Power Operation or Startup), however, requiresemergency classification of at least an Alert. The operating mode change associated withmovement of the Mode Switch, by itself, does not justify failure to declare an emergencyfor ATWS events.GenericThe reactor should be considered shutdown when it producing less heat than the maximum decayheat load for which the safety systems are designed (4% power).Manual scram actions taken at the reactor control console are any set of actions by the reactoroperator(s) which causes or should cause control rods to be rapidly inserted into the core andshuts down the reactor.This condition indicates failure of the automatic protection system to scram the reactor. Thiscondition is more than a potential degradation of a safety system in that a front line automaticprotection system did not function in response to a scram signal. Thus the plant safety has beencompromised because of the failure of RPS to automatically shut down the plant. An Alert isindicated because conditions may exist that lead to potential loss of fuel clad barrier or RCS barrierand because of the failure of the Reactor Protection System to automatically shut down the plant.If manual actions taken at the reactor control console fail to shut down the reactor, the event wouldescalate to a Site Area Emergency.NMP2 Basis Reference(s):1. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, Table3.3.1.1-12. N2-EOP-C5 Failure to Scram3. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document4. NEI 99-01 IC SA2Page 183 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:S -System Malfunction3 -Criticality & RPS FailureInitiating Condition: Inadvertent criticalityEAL:SU3.1 Unusual EventAn UNPLANNED sustained positive period observed on nuclear instrumentationMode Applicability: Hot ShutdownBasis:Plant-SpecificThe term "sustained" is used to allow exclusion of expected short-term positive periodsfrom planned fuel bundle or control rod movements during core alteration. These short-term positive periods are the result of the rise in neutron population due to subcriticalmultiplication.GenericThis EAL addresses inadvertent criticality events. While the primary concern of this EAL iscriticality This EAL addresses inadvertent criticality events. This EAL indicates a potentialdegradation of the level of safety of the plant, warranting a UE classification. This EAL excludesinadvertent criticalities that occur during planned reactivity changes associated with, reactorstartups (e.g., criticality earlier than estimated).Escalation would be by EALs in Category F, as appropriate to the operating mode at the time ofthe event.NMP2 Basis Reference(s):1. NEI 99-01 IC SU8Page 184EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:EAL:S -System Malfunction4 -Inability to Reach or Maintain Shutdown ConditionsInability to reach required shutdown within Technical SpecificationlimitsSU4.1 Unusual EventPlant is not brought to required operating mode within Technical Specifications LCOrequired action completion timeMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificLimiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safeoperation of the unit. The actions associated with an LCO state conditions that typicallydescribe the ways in which the requirements of the LCO can fail to be met. Specified witheach stated condition are required action completion times. (ref. 1)GenericLimiting Conditions of Operation (LCOs) require the plant to be brought to a required operatingmode when the Technical Specification required configuration cannot be restored. Depending onthe circumstances, this may or may not be an emergency or precursor to a more severe condition.In any case, the initiation of plant shutdown required by the site Technical Specifications requires afour hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safetyenvelope when being shut down within the allowable required action completion time in theTechnical Specifications. An immediate UE is required when the plant is not brought to therequired operating mode within the allowable required action completion time in the TechnicalSpecifications. Declaration of a UE is based on the time at which the LCO-specified required actioncompletion time period elapses under the site Technical Specifications and is not related to howlong a condition may have existed.NMP2 Basis Reference(s):1. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 1.32. NEI 99-01 IC SU2Page 185EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:Initiating Condition:EAL:S -System Malfunction5 -InstrumentationInability to monitor a significant transient in progressSS5.1 Site Area EmergencyLoss of > approximately 75% of annunciation or indication on all of the following ControlRoom panels for _ 15 min. (Note 4):" 2CEC*PNL601* 2CEC*PNL602" 2CEC*PNL603" 2CEC*PNL851" 2CEC*PNL852ANDA significant transient is in progress, Table S-2ANDCompensatory indications are unavailable (Plant Process Computer, SPDS)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeTable S-2 Significant Transients" Automatic turbine runback > 25% thermal reactor power* Electric load rejection > 25% full electrical load* Reactor scram* ECCS injection* Thermal power oscillations > 10%Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificPlant Process Computer and SPDS are considered compensatory indication.Significant transients are listed in Table S-2.Page 186 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesGenericThis EAL is intended to recognize the threat to plant safety associated with the complete loss ofcapability of the control room staff to monitor plant response to a significant transient."Planned" and "UNPLANNED" actions are not differentiated since the loss of instrumentation of thismagnitude is of such significance during a transient that the cause of the loss is not anameliorating factor.Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety systemannunciators or indicators are lost, there is an increased risk that a degraded plant condition couldgo undetected. It is not intended that plant personnel perform a detailed count of theinstrumentation lost but use the value as a judgment threshold for determining the severity of theplant conditions. It is also not intended that the Shift Manager be tasked with making a judgmentdecision as to whether additional personnel are required to provide increased monitoring of systemoperation.It is further recognized that most plant designs provide redundant safety system indication poweredfrom separate uninterruptible power supplies. While failure of a large portion of annunciators ismore likely than a failure of a large portion of indications, the concern is included in this EAL due todifficulty associated with assessment of plant conditions. The loss of specific, or several, safetysystem indicators should remain a function of that specific system or component operability status.This will be addressed by the specific Technical Specification. The initiation of a TechnicalSpecification imposed plant shutdown related to the instrument loss will be reported via 10 CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the NOUE isbased on EAL SU4.1A Site Area Emergency is considered to exist if the control room staff cannot monitor safetyfunctions needed for protection of the public while a significant transient is in progress.Annunciators for this EAL are limited to include those identified in the Abnormal OperatingProcedures, in the Emergency Operating Procedures, and in other EALs (.g., area, process, and/oreffluent rad monitors, etc.)Indications needed to monitor safety functions necessary for protection of the public include controlroom indications, computer generated indications and dedicated annunciation capability."Compensatory indications" in this context includes computer based information such as PlantProcess Computer and SPDS.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.NMP2 Basis Reference(s):1. USAR Figure 1.2-152. N2-OP-91A Process Computer3. N2-OP-91 B Safety Parameter Display System (SPDS)4. SOP-78A EOP Key Parameter Alternate Instrumentation5. NEI 99-01 IC SS6Page 187 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 5 -InstrumentationInitiating Condition: UNPLANNED loss of safety system annunciation or indication inthe Control Room with either (1) a significant transient in progress,or (2) compensatory indicators are unavailableEAL:SA5.1 AlertUNPLANNED loss of > approximately 75% of annunciation or indication on all of thefollowing Control Room panels for >_ 15 min. (Note 4):" 2CEC*PNL601* 2CEC*PNL602* 2CEC*PNL603" 2CEC*PNL851" 2CEC*PNL852AND EITHER:A significant transient is in progress, Table S-2ORCompensatory indications are unavailable (Plant Process Computer, SPDS)Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeTable S-2 Significant Transients" Automatic turbine runback > 25% thermal reactor power* Electric load rejection > 25% full electrical load" Reactor scram" ECCS injection" Thermal power oscillations > 10%Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificPlant Process Computer and SPDS are considered compensatory indication.Page 188EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesSignificant transients are listed in Table S-2.GenericThis EAL is intended to recognize the difficulty associated with monitoring changing plantconditions without the use of a major portion of the annunciation or indication equipment during asignificant transient."Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety systemannunciators or indicators are lost, there is an increased risk that a degraded plant condition couldgo undetected. It is not intended that plant personnel perform a detailed count of theinstrumentation lost but use the value as a judgment threshold for determining the severity of theplant conditions. It is also not intended that the Shift Manager be tasked with making a judgmentdecision as to whether additional personnel are required to provide increased monitoring of systemoperation.It is further recognized that most plant designs provide redundant safety system indication poweredfrom separate uninterruptible power supplies. While failure of a large portion of annunciators ismore likely than a failure of a large portion of indications, the concern is included in this EAL due todifficulty associated with assessment of plant conditions. The loss of specific, or several, safetysystem indicators should remain a function of that specific system or component operability status.This will be addressed by the specific Technical Specification. The initiation of a TechnicalSpecification imposed plant shutdown related to the instrument loss will be reported via 10 CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the UE is basedon EAL SU4.1.Annunciators or indicators for this EAL include those identified in the Abnormal OperatingProcedures, in the Emergency Operating Procedures, .and in other EALs (e.g., area, process,and/or effluent rad monitors, etc.)."Compensatory indications" in this context includes computer based information such as PlantProcess Computer and SPDS.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor thetransient in progress due to a concurrent loss of compensatory indications with a significanttransient in progress during the loss of annunciation or indication.NMP2 Basis Reference(s):1. USAR Figure 1.2-152. N2-OP-91A Process Computer3. N2-OP-91 B Safety Parameter Display System (SPDS)4. SOP-78A EOP Key Parameter Alternate Instrumentation5. NEI 99-01 IC SA4Page 189 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:S -System Malfunction5 -InstrumentationInitiating Condition: UNPLANNED loss of safety system annunciation or indication inthe Control Room for __ 15 min.EAL:SU5.1 Unusual EventUNPLANNED loss of > approximately 75% of annunciation or indication on all of thefollowing Control Room panels for __ 15 min. (Note 4):" 2CEC*PNL601* 2CEC*PNL602* 2CEC*PNL603* 2CEC*PNL851* 2CEC*PNL852Note 4: The ED should not wait until the applicable time has elapsed, but should declare the event as soon as it isdetermined that the condition has exceeded, or will likely exceed, the applicable timeMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificNoneGenericThis EAL is intended to recognize the difficulty associated with monitoring changing plantconditions without the use of a major portion of the annunciation or indication equipment.Recognition of the availability of computer based indication equipment is considered."Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety systemannunciators or indicators are lost, there is an increased risk that a degraded plant condition couldgo undetected. It is not intended that plant personnel perform a detailed count of theinstrumentation lost but use the value as a judgment threshold for determining the severity of theplant conditions.It is further recognized that plant design provides redundant safety system indication powered fromseparate uninterruptible power supplies. While failure of a large portion of annunciators is morelikely than a failure of a large portion of indications, the concern is included in this EAL due toPage 190EPMP-EPP-01 02Rev 00 (Draft A) -Emergency Action Level Technical Basesdifficulty associated with assessment of plant conditions. The loss of specific, or several, safetysystem indicators should remain a function of that specific system or component operability status.This will be addressed by the specific Technical Specification. The initiation of a TechnicalSpecification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72 If the shutdown is not in compliance with the Technical Specification action, the UE is basedon EAL SU4.1.Annunciators or indicators for this EAL include those identified in the Abnormal OperatingProcedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, process,and/or effluent rad monitors, etc.).Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.This UE will be escalated to an Alert based on a concurrent loss of compensatory indications or if asignificant transient is in progress during the loss of annunciation or indication.NMP2 Basis Reference(s):1. USAR Figure 1.2-152. N2-OP-91A Process Computer3. N2-OP-91B Safety Parameter Display System (SPDS)4. SOP-78A EOP Key Parameter Alternate Instrumentation5. NEI 99-01 IC SU3Page 191EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:S -System Malfunction6 -CommunicationsInitiating Condition: Loss of all onsite or offsite communications capabilitiesEAL:SU6.1 Unusual EventLoss of all Table S-3 onsite (internal) communication methods affecting the ability toperform routine operationsORLoss of all Table S-3 offsite (external) communication methods affecting the ability toperform offsite notificationsTable S-3 Communications SystemsSystem Onsite Offsite(internal) (external)PBX (normal dial telephones)GaitronicsStation radio (portable)Control Room installed satellite phones (non portable)ENSRECSXXXXXXXMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificOnsite/offsite communications systems are listed in Table S-3 (ref. 1, 2, 3).This EAL is the hot condition equivalent of the cold condition EAL CU6.1.GenericThe purpose of this EAL is to recognize a loss of communications capability that either defeats theplant operations staff ability to perform routine tasks necessary for plant operations or the ability tocommunicate issues with off-site authorities.Page 192EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesThe loss of off-site communications ability is expected to be significantly more comprehensive thanthe condition addressed by 10 CFR 50.72.The availability of one method of ordinary off-site communications is sufficient to inform federal,state, and local authorities of plant problems. This EAL is intended to be used only whenextraordinary means (e.g., relaying of information from non-routine radio transmissions, individualsbeing sent to off-site locations, etc.) are being used to make communications possible.NMP2 Basis Reference(s):1. USAR Section 9.5.22. Nine Mile Point Nuclear Station Site Emergency Plan, Section 7.23. N2-OP-76 Plant Communications4. NEI 99-01 IC SU6Page 193 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 7 -Fuel Clad DegradationInitiating Condition: Fuel clad degradationEAL:SU7.1 Unusual EventReactor coolant activity > 4 pCi/gm 1-131 EquivalentMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificThis EAL addresses reactor coolant samples exceeding Technical Specification 3.4.8(ref. 1). A reactor coolant sample analysis with specific activity in excess of the TechnicalSpecification limit of 4 pCi/gm 1-131 Equivalent is indicative of a degradation of the fuelclad, and is a precursor of more serious problems. This activity level for which operation isallowed to continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to accommodate short duration Iodine spikesfollowing changes in thermal power.GenericThis EAL is included because it is a precursor of more serious conditions and, as result, isconsidered to be a potential degradation of the level of safety of the plant.Escalation of this EAL to the Alert level is via the EALs in Category F.This threshold addresses coolant samples exceeding coolant technical specifications for transientiodine spiking limits.NMP2 Basis Reference(s):1. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, 3.4.82. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2,3.4.8.A.13. NEI 99-01 IC SU4Page 194 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: S -System MalfunctionSubcategory: 7 -Fuel Clad DegradationInitiating Condition: Fuel clad degradationEAL:SU7.2 Unusual EventOffgas radiation DRMS high (red) alarm for > 15 min.Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificElevated offgas radiation activity represents a potential degradation in the level of safety ofthe plant and a potential precursor of more serious problems. The Technical Specificationallowable limit is an offgas level not to exceed 350,000 pCi/sec (ref. 1). The DRMS alarmsetpoint has been conservatively selected because it is operationally significant and isreadily recognizable by Control Room operating staff. 15 minutes is allotted for operatoraction to reduce the offgas radiation levels and exclude TRANSIENT conditions (ref. 2, 3,4). The high offgas radiation alarm is set using methodology outlined in the ODCM (ref. 5).GenericThis EAL is included because it is a precursor of more serious conditions and, as result, isconsidered to be a potential degradation of the level of safety of the plant.Escalation of this EAL to the Alert level is via the EALs in Category F.This threshold addresses radiation monitor readings that provide indication of a degradation of fuelclad integrity.NMP2 Basis Reference(s):1. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No.2, 3.7.42. N2-ARP-01 Annunciator Response Procedures for annunciator 8512533. N2-ARP-01 Annunciator Response Procedures for annunciator 8513264. N2-SOP-17 Fuel Failure or High Activity in Rx Coolant or Offgas5. Offsite Dose Calculation Manual 3.3.26. NEI 99-01 IC SU4Page 195 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory:Subcategory:S -System Malfunction8 -RCS LeakageInitiating Condition: RCS leakageEAL:SU8.1 Unusual EventUnidentified or reactor coolant pressure boundary leakage > 10 gpmORIdentified reactor coolant leakage > 25 gpmMode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificElevated RCS leakage may be detected by the following annunciators (ref. 1-4):* 873115 DRWL FLR DRN LEAK RATE HIGH (setpoint 4 gpm)* 873111 DRWL FLR DRN TANK 1 LEVELHI-HI* 873105 DRWL EQPT DRN TANK 1 LEVEL HI-HI0 873110 DRWL EQPT DRN DAILY LK RATE HIGHThe Plant Process Computer monitors unidentified and identified leakage over six minuteintervals (Computer Point DERXA01) as well as a twenty-four hour average (ComputerPoint 2DER-FI1 01). Leak rates can also be verified by alternate measurements accordingto N2-OSP-LOG-S001, Attachments 6 and 7 (ref. 5, 6).GenericThis EAL is included as a UE because it may be a precursor of more serious conditions and, asresult, is considered to be a potential degradation of the level of safety of the plant. The 10 gpmvalue for the unidentified or pressure boundary leakage was selected as it is observable withnormal control room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances).Relief valve normal operation should be excluded from this EAL. However, a relief valve thatoperates and fails to close per design should be considered applicable to this EAL if the relief valvecannot be isolated.Page 196EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesThe EAL for identified leakage is set at a higher value due to the lesser significance of identifiedleakage in comparison to unidentified or pressure boundary leakage. In either case, escalation ofthis EAL to the Alert level is via EALs in Category F.NMP2 Basis Reference(s):1. N2-ARP-01 Annunciator Response Procedures for annunciator 8731152. N2-ARP-01 Annunciator Response Procedures for annunciator 8731113. N2-ARP-01 Annunciator Response Procedures for annunciator 8731054. N2-ARP-01 Annunciator Response Procedures for annunciator 8731105. N2-OSP-LOG-SO01 Shift Checks -Mode 16. N2-OP-67 Drywell Equipment and Floor Drains System7. NEI 99-01 IC SU5Page 197 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory F -Fission Product Barrier DegradationEAL Group: Hot Conditions (RCS temperature > 2001F);EALs in this category are applicable only inone or more hot operating modes.EALs in this category represent threats to the defense in depth design concept thatprecludes the release of highly radioactive fission products to the environment. Thisconcept relies on multiple physical barriers any one of which, if maintained INTACT,precludes the release of significant amounts of radioactive fission products to theenvironment. The primary fission product barriers are:A. Fuel Clad (FC): Zirconium tubes which house the ceramic uranium oxide pelletsalong with the end plugs which are welded into each end of the fuel rods comprisethe FC barrier.B. Reactor Coolant System (RCS): The reactor vessel shell, vessel head, CRDhousings, vessel nozzles and penetrations, and all primary systems directlyconnected to the RPV up to the outermost Primary Containment isolation valvecomprise the RCS barrier.C. Containment (PC): The drywell, the suppression chamber/pool, their respectiveinterconnecting paths, and other connections up to and including the outermostcontainment isolation valves comprise the Primary Containment barrier.The EALs in this category require evaluation of the loss and potential loss thresholds listedin the fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "PotentialLoss" signify the relative damage and threat of damage to the barrier. "Loss" means thebarrier no longer assures containment of radioactive materials. "Potential Loss" meansintegrity of the barrier is threatened and could be lost if conditions continue to degrade.The number of barriers that are lost or potentially lost and the following criteria determinethe appropriate emergency classification level:Unusual Event:Any loss or any potential loss of ContainmentAlert:Any loss or any potential loss of either Fuel Clad or RCSSite Area Emergency:Loss or potential loss of any two barriersGeneral Emergency:Loss of any two barriers and loss or potential loss of the third barrierPage 198 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesThe logic used for Category F EALs reflects the following considerations:" The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than theContainment Barrier. UE EALs associated with RCS and Fuel Clad Barriers areaddressed under Category S.* At the Site Area Emergency level, there must be some ability to dynamically assesshow far present conditions are from the threshold for a General Emergency. Forexample, if Fuel Clad and RCS Barrier "Loss" thresholds existed, that, in addition tooff-site dose assessments, would require continual assessments of radioactiveinventory and containment integrity. Alternatively, if both Fuel Clad and RCS Barrier"Potential Loss" thresholds existed, the ED would have more assurance that therewas no immediate need to escalate to a General Emergency.* The ability to escalate to higher emergency classification levels as an eventdeteriorates must be maintained. For example, RCS leakage steadily increasingwould represent an increasing risk to public health and safety." The Containment Barrier should not be declared lost or potentially lost based onexceeding Technical Specification action statement criteria, unless there is an eventin progress requiring mitigation by the Containment barrier.Page 199EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: Loss of ANY two barriers and loss or potential loss of the thirdbarrierEAL:FGI.1 General EmergencyLoss of ANY two fission product barriersAND.Loss or potential loss of third fission product barrier (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificFuel Clad, RCS and Containment comprise the fission product barriers. Table F-1(Attachment 2) lists the fission product barrier thresholds, bases and references.At the General Emergency classification level each barrier is weighted equally. A GeneralEmergency is therefore appropriate for any combination of the following conditions:* Loss of Fuel Clad, RCS and Containment barriers" Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier* Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier" Loss of Fuel Clad and Containment barriers with potential loss of RCS barrierGenericNoneNMP2 Basis Reference(s):1. NEI 99-01 IC FG1Page 200 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: Loss or potential loss of ANY two barriersEAL:FSI.1 Site Area EmergencyLoss or potential loss of ANY two fission product barriers (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificFuel Clad, RCS and Containment comprise the fission product barriers. Table F-1(Attachment 2) lists the fission product barrier thresholds, bases and references.At the Site Area Emergency classification level, each barrier is weighted equally. A SiteArea Emergency is therefore appropriate for any combination of the following conditions:* One barrier loss and a second barrier loss (i.e., loss -loss)" One barrier loss and a second barrier potential loss (i.e., loss -potential loss)" One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss)At the Site Area Emergency classification level, the ability to dynamically assess theproximity of present conditions with respect to the threshold for a General Emergency isimportant. For example, the existence of Fuel Clad and RCS Barrier loss thresholds inaddition to offsite dose assessments would require continual assessments of radioactiveinventory and Containment integrity in anticipation of reaching a General Emergencyclassification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed,the Emergency Director would have greater assurance that escalation to a GeneralEmergency is less IMMINENT.GenericNoneNMP2 Basis Reference(s):1. NEI 99-01 IC FS1Page 201 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: ANY loss or ANY potential loss of EITHER Fuel Clad OR RCSEAL:FA1.1 AlertANY loss or ANY potential loss of EITHER Fuel Clad barrier OR RCS barrier (Table F-I)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificFuel Clad, RCS and Containment comprise the fission product barriers. Table F-1(Attachment 2) lists the fission product barrier thresholds, bases and references.At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavilythan the Containment barrier. Unlike the Containment barrier, loss or potential loss ofeither the Fuel Clad or RCS barrier may result in the relocation of radioactive materials ordegradation of core cooling capability. Note that the loss or potential loss of Containmentbarrier in combination with loss or potential loss of either Fuel Clad or RCS barrier resultsin declaration of a Site Area Emergency under EAL FS1.GenericNoneNMP2 Basis Reference(s):1. NEI 99-01 IC FA1Page 202 EPMP-EPP-0102Rev 00 (Draft A) -Emergency Action Level Technical BasesCategory: Fission Product Barrier DegradationSubcategory: N/AInitiating Condition: ANY loss or ANY potential loss of ContainmentEAL:FUI.1 Unusual EventANY loss or ANY potential loss of Containment barrier (Table F-i)Mode Applicability:1 -Power Operation, 2 -Startup, 3 -Hot ShutdownBasis:Plant-SpecificFuel Clad, RCS and Containment comprise the fission product barriers. Table F-1(Attachment 2) lists the fission product barrier thresholds, bases and references.Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier.Unlike the Fuel Clad and RCS barriers, the loss of either of which results in an Alert (EALFA1.1), loss of the Containment barrier in and of itself does not result in the relocation ofradioactive materials or the potential for degradation of core cooling capability. However,loss or potential loss of the Containment barrier in combination with the loss or potentialloss of either the Fuel Clad or RCS barrier results in declaration of a Site Area Emergencyunder EAL FS1.1.GenericNoneNMP2 Basis Reference(s):1. NEI 99-01 IC FUlPage 203 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisIntroductionTable F-1 lists the threshold conditions that define the Loss and Potential Loss of the threefission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The tableis structured so that each of the three barriers occupies adjacent columns. Each fissionproduct barrier column is further divided into two columns; one for Loss thresholds and onefor Potential Loss thresholds.The first column of the table (to the left of the Fuel Clad Loss column) lists the categories(types) of fission product barrier thresholds. The fission product barrier categories are:A. RPV LevelB. Primary Containment Pressure / TemperatureC. IsolationD. RadE. JudgmentEach category occupies a row in Table F-1 thus forming a matrix defined by thecategories. The intersection of each row with each Loss/Potential Loss column forms a cellin which one or more fission product barrier thresholds appear. If NEI 99-01 does notdefine a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.Thresholds are assigned sequential numbers within each Loss and Potential Loss columnbeginning with number one. In this manner, a threshold can be identified by its categorytitle and number. For example, the first Fuel Clad barrier Loss in Category A would beassigned "FC Loss A.1," the third Containment barrier Potential Loss would be assigned"PC P-Loss B.3," etc.If a cell in Table F-1 contains more than one numbered threshold, each of the numberedthresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessaryto exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.Subdivision of Table F-1 by category facilitates association of plant conditions to theapplicable fission product barrier Loss and Potential Loss thresholds. This structurepromotes a systematic approach to assessing the classification status of the fissionPage 204 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basisproduct barriers.When equipped with knowledge of plant conditions related to the fission product barriers,the EAL-user first scans down the category column of Table F-I, locates the likelycategory and then reads across the fission product barrier Loss and Potential Lossthresholds in that category to determine if a threshold has been exceeded. If a thresholdhas not been exceeded, the EAL-user proceeds to the next likely category and continuesreview of the thresholds in the new categoryIf the EAL-user determines that any threshold has been exceeded, by definition, the barrieris lost or potentially lost -even if multiple thresholds in the same barrier column areexceeded; only that one barrier is lost or potentially lost. The EAL-user must examine eachof the three fission product barriers to determine if other barrier thresholds in the categoryare lost or potentially lost. For example, if Primary Containment radiation is sufficientlyhigh, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Containmentbarrier can occur. Barrier Losses and Potential Losses are then applied to the algorithmsgiven in EALs FG1.1, FS1.1, FA1.1 and FUI.1 to determine the appropriate emergencyclassification.In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first,followed by the RCS barrier and finally the Containment barrier threshold bases. In eachbarrier, the bases are given according to category Loss followed by category PotentialLoss beginning with Category A, then B,..., E.Page 205 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisTable F-1 Fission Product Barrier MatrixFuel Clad Barrier Reactor Coolant System Barrier Containment BarrierCategory Loss Potential Loss Loss Potential Loss Loss Potential Loss1. RPV water level cannot berestored and maintained 1, RPV water level cannot beA 1 Primary Containment above -14 in. following restored and maintained None None 1. Primary Containment Flooding isRPM Level Flooding is required depressurization of the RPV or above -14 in. or RPV water requiredRPV water level cannot be level cannot be determineddetermined1. Primary Containment pressure 2. Primary Containment pressure >rise followed by a rapid 45 psig and risingB UNPLANNED drop in Primary 3. Explosive mixture exists insidePrimary Containment pressure Primary ContainmentContainm None None 2, Primary Containment pressure None Primary Containment pressure (a 6% H2 and 2 5% 02)ent > 1.68 psig due to RCS leakage response not consistent with 4. Suppression pool temperature andLOCA conditions RPV pressure cannot bePreasure Imaintained below the HeatTemp. Capacity Temperature Limit(N2-EOP-PC Figure M)3. Failure of all PrimaryContainment isolation valves inANY one line to close followingauto or manual initiation3. Release pathway exists outside ANDPrimary Containment resulting UNISOLABLE primary system Direct downstream pathwayfrom isolation failure in ANY of leakage outside Primary outside Primary Containment andthe following (excluding normal Containment as indicated by to the environment existsprocess system flowpaths from exceeding EITHER: 4. Intentional Primary ContainmentC None None an UNISOLABLE system): RB area temperature above an venting per EOPs NoneotMain steam line isolation selpoint 5. UNISOLABLE primary systemI RCIC steam line OR leakage outside PrimaryContainment as indicated byRWCU RB area radiation above an exceeding EITHER:* Feedwater alarm setpoint RB area maximum safe4. RPV blowdown is required temperature value(N2-EOP-SC Detail S)ORRB area radiation> 8.OOE+3 mR/hr2. Drywell area radiationD 3100 R/hr (3.1 6 mRem/hr) None 5. Drywell area radiation None None 5. Drywall area radiation_> 41 R/hr (4.1 E4 mRem/hr) :6.0 E4 R/hr (6.0 E7 mRem/hr)Red 3. Reactor coolant activity> 300 pCi/gm 1-131 Equivalent4. ANY condition in the opinion of 2. ANY condition in the opinion of 6. ANY condition in the opinion of 2. ANY condition in the opinion of the 6. ANY condition in the opinion of 6. ANY condition in the opinion of theE the Emergency Director that the Emergency Director that the Emergency Director that Emergency Director that indicates .the Emergency Director that Emergency Director that indicatesindicates loss of the Fuel Clad indicates potential loss of the indicates loss of the Reactor potential loss of the Reactor indicates loss of the Containment potential loss of the ContainmentJudgment barrier Fuel Clad barrier Coolant System barrier Coolant System barrier barrier barrierPage 206EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: A. RPV Water LevelDegradation Threat: LossThreshold:1. Primary Containment Flooding is requiredBasis:Plant-SpecificRequirements for Primary Containment Flooding are established in EOP-RPV Step L-16;EOP-C5 Steps L-8, L-10 and L-18; and EOP-C4 Override 1. These EOPs provideinstructions to ensure adequate core cooling by maintaining RPV water level aboveprescribed limits or operating sufficient RPV injection sources when level cannot bedetermined. SAP entry is required when (ref. 1):RPV water level cannot be restored and maintained above -39 in. with insufficientCore Spray Cooling: The Minimum Steam Cooling RPV Water Level (MSCRWL) isthe lowest RPV water level at which the covered portion of the reactor core willgenerate sufficient steam to preclude any clad temperature in the uncovered portionof the core from exceeding 15000F. Core Spray Cooling is insufficient if RPV waterlevel cannot be restored and maintained at or above -62 in. with at least 6350 gpmcore spray loop flow. Consistent with the EOP definition of "cannot be restored andmaintained," the determination that the parameter cannot be restored andmaintained above the limit may be made at, before, or after the parameter actuallydecreases to this point.RPV water level cannot be determined and it is determined that core damage isoccurring: When RPV water level cannot be determined, EOPs require RPVflooding strategies. RPV water level indication provides the primary means ofknowing if adequate core cooling is being maintained. When all means ofdetermining RPV water level are unavailable, reliance on alternate means ofassuring adequate core cooling must be attempted. The instructions in EOP-C4specify these means, which include emergency depressurization of the RPV andPage 207 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss I Potential Loss Matrix and Basisinjection into the RPV at a rate needed to flood to the elevation of the main steamlines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWSevents)This threshold is also a Potential Loss of the Containment barrier (PC P-Loss A.1). SinceSAP entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RCSLoss A.1). Primary Containment Flooding (SAP entry), therefore, represents a Loss of twobarriers and a Potential Loss of a third, which requires a General Emergency classification.GenericThis site specific value corresponds to the level used in EOPs to indicate challenge of core cooling.This is the minimum value to assure core cooling without further degradation of the clad.NMP2 Basis Reference(s):1. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document2. N2-EOP-C4 RPV Flooding3. NEI 99-01 FC Loss 2Page 208EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: LossThreshold:NonePage 209EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: C. IsolationDegradation Threat: LossThreshold:NonePage 210EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: D. RadDegradation Threat: LossThreshold:2. Drywell area radiation > 3100 R/hr (3.1 E6 mRem/hr)Basis:Plant-SpecificIt is important to recognize that the radiation monitor may be sensitive to shine from theRPV or RCS piping (caused by lower than normal RPV water level for example). TheDrywell High Range Radiation Monitors are the following (ref. 1):" 2CEC*PNL88OD: DRMS 2RMS*RE1 B/DRMS*RUZ1ARMS*RUZ1 B* 2CEC*PNL88OB: DRMS 2RMS*RE1A/CRMS*RUZ1 CRMS*RUZ1 DFigure F-1 illustrates the location of the following four detectors inside the drywell (ref. 1):* 2RMS*RE1A P.C. 268 170EAZ* 2RMS*RE1C P.C. 267 024EAZ* 2RMS*RE1B P.C. 268 245EAZ* 2RMS*RE1D P.C. 268 353EAZThe threshold value was calculated assuming the instantaneous release and dispersal ofthe reactor coolant noble gas and iodine inventory associated with a concentration of 300pCi/gm 1-131 Equivalent (or approximately 5% clad failure) into the drywell atmosphere(ref. 2).Page 211 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisGenericThe 3100 R/hr (3.1 E6 mRem/hr) reading is a value which indicates the release of reactor coolant,with elevated activity indicative of fuel damage, into the drywell.Reactor coolant concentrations of this magnitude are several times larger than the maximumconcentrations (including iodine spiking) allowed within technical specifications and are thereforeindicative of fuel damage.This value is higher than that specified for RCS barrier Loss threshold D.5. Thus, this thresholdindicates a loss of both Fuel Clad barrier and RCS barrier that appropriately escalates theemergency classification level to a Site Area Emergency.There is no Potential Loss threshold associated with this item.NMP2 Basis Reference(s):1. N2-RSP-RMS-R106 Channel Calibration Test of the Drywell High Range AreaRadiation Monitors2. Calculation PR-C-24-03. NEI 99-01 FC Loss 4Page 212 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisFigure F-I: Drywell High Range Radiation Monitor Detector Locations (ref. 1)Drywell 261DrywelI 261Irs//AZ 24_Dag___'N/ 'N/N /N AZ1~'tJ OO~c~ýe~qrsrjneI &Equ i pme nthrt--t/0~~~t"tJ/I'--N-N---,If\ \NFWS'.E , N'I-NN-.,-- ----*A~~~Nf<"IN- ~C ~ 2~5 ~~reese~c~ape-~ -------HRA ILCC-,e~ \.zN~..-'CAhatchNPage 213EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:Fuel CladCategory:D. RadDegradation Threat: LossThreshold:3. Reactor coolant activity > 300 pCi/gm 1-131 EquivalentBasis:Plant-SpecificNoneGenericThe site specific value corresponds to 300 pCi/gm 1-131 Equivalent. Assessment by the EAL TaskForce indicates that 300 pCi/gm 1-131 Equivalent coolant activity is well above that expected foriodine spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivityindicates significant clad damage and thus the Fuel Clad Barrier is considered lost.There is no Potential Loss threshold associated with this item.NMP2 Basis Reference(s):1. General Electric NEDO-22215, Procedures for the Determination of the Extent of CoreDamage Under Accident Conditions2. NEI 99-01 FC Loss 1Page 214EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: E. JudgmentDegradation Threat: LossThreshold:4. ANY condition in the opinion of the Emergency Director that indicates loss of the FuelClad barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the Fuel Clad barrier is lost. Such a determination should include IMMINENTbarrier degradation, barrier monitoring capability and dominant accident sequences." IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to recognition of the inability to reach safety acceptance criteriabefore completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the Fuel Clad barrier is lost. In addition, the inability to monitor the barriershould also be incorporated in this threshold as a factor in Emergency Director judgment that thebarrier may be considered lost.Page 215 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP2 Basis Reference(s):1. NEI 99-01 FC Loss 6Page 216 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: A. RPV LevelDegradation Threat: Potential LossThreshold:1. RPV water level cannot be restored and maintained above -14 in. followingdepressurization of the RPV or cannot be determinedBasis:Plant-SpecificAn RPV water level instrument reading of -14 in. indicates RPV water level is at the top ofactive fuel. When RPV water level is at or above the top of active fuel, the core iscompletely submerged. Core submergence is the most desirable means of core cooling.When RPV water level is below the top of active fuel following depressurization of the RPV(automatically, manually or by failure of the RCS barrier), the uncovered portion of the coremust be cooled by less reliable means (i.e., spray cooling). If core uncovery is threatened,the EOPs specify alternate, more extreme, RPV water level control measures in order torestore and maintain adequate core cooling (ref. 1).Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that RPV water level cannot be restored and maintained above the top ofactive fuel may be made at, before, or after RPV water level actually decreases to thispoint. (ref. 1)When RPV water level cannot be determined, EOPs require RPV flooding strategies. RPVwater level indication provides the primary means of knowing if adequate core cooling isbeing maintained. When all means of determining RPV water level are unavailable, thefuel clad barrier is threatened and reliance on alternate means of assuring adequate corecooling must be attempted. The instructions in EOP-C4 specify these means, whichinclude emergency depressurization of the RPV and injection into the RPV at a rateneeded to flood to the elevation of the main steam lines or hold RPV pressure above theMinimum Steam Cooling Pressure (in ATWS events). (ref. 2) If RPV water level cannot bedetermined with respect to the top of active fuel, a potential loss of the fuel clad barrierPage 217 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basisexists.Note that EOP-C5 may require intentional uncovery of the core and control of RPV waterlevel between -14 in. and -39 in., the Minimum Steam Cooling RPV Water Level(MSCRWL) (ref. 3). Under these conditions, a high-power ATWS event exists and requiresat least a Site Area Emergency classification in accordance with the ATWS/CriticalityEALs.GenericThe site specific RPV water level threshold is the same as the RCS barrier Loss threshold A.1 andcorresponds to the RPV water level at the top of the active fuel. Thus, this threshold indicates aPotential Loss of the Fuel Clad barrier and a Loss of RCS barrier that appropriately escalates theemergency classification level to a Site Area Emergency. This threshold is considered to beexceeded when, as specified in the site specific EOPs, that RPV water cannot be restored andmaintained above the specified level following depressurization of the RPV (either manually,automatically or by failure of the RCS barrier).NMP2 Basis Reference(s):1. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document2. N2-EOP-C4 RPV Flooding3. N2-EOFF-C5 Failure to Scram4. NEI 99-01 FC Potential Loss 2Page 218EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: Potential LossThreshold:NonePage 219EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: C. IsolationDegradation Threat: Potential LossThreshold:NonePage 220EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: D. RadDegradation Threat: Potential LossThreshold:NonePage 221EPMP-EPP-01 02Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Fuel CladCategory: E. JudgmentDegradation Threat: Potential LossThreshold:2. ANY condition in the opinion of the Emergency Director that indicates potential loss ofthe Fuel Clad barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the Fuel Clad barrier is potentially lost. Such a determination should includeIMMINENT barrier degradation, barrier monitoring capability and dominant accidentsequences." IMMINENT barrier deqradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to recognition of the inability to reach safety acceptance criteriabefore completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the Fuel Clad barrier is potentially lost. In addition, the inability to monitor thebarrier should also be incorporated in this threshold as a factor in Emergency Director judgmentthat the barrier may be considered potentially lost.Page 222 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP2 Basis Reference(s):1. NEI 99-01 FC Potential Loss 6Page 223 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: A. RPV LevelDegradation Threat: LossThreshold:1. RPV water level cannot be restored and maintained above -14 in. or cannot bedeterminedBasis:Plant-SpecificAn RPV water level instrument reading of -14 in. indicates RPV water level is at the top ofactive fuel (ref. 1). The top of the active fuel is significantly lower than the normal operatingRPV water level control band. To reach this level, RPV inventory loss would havepreviously required isolation of the RCS and Containment (PC) barriers, and initiation of allECOS. If RPV water level cannot be maintained above the top of active fuel, ECCS andother sources of RPV injection have been ineffective or incapable of reversing thedecreasing level trend. The cause of the loss of RPV inventory is therefore assumed to bea Loss of Coolant Accident (LOCA). By definition, a LOCA event is a Loss of the RCSbarrier.Consistent with the EOP definition of "cannot be restored and maintained," thedetermination that RPV water level cannot be restored and maintained above the top ofactive fuel may be made at, before, or after RPV water level actually decreases to thispoint. (ref. 1)When RPV level cannot be determined, EOPs require RPV flooding strategies. The RPVflooding instructions in EOP-C4 first specify emergency depressurization of the RPV (ref.2), which is defined to be a Loss of the RCS barrier (RCS Loss C.4).Note that EOP-C5 may require intentional uncovery of the core and control of RPV waterlevel between -14 in. and -39 in., the Minimum Steam Cooling RPV Water Level(MSCRWL) (ref. 3). Under these conditions, a high-power ATWS event exists and requiresat least a Site Area Emergency classification in accordance with the ATWS/CriticalityPage 224 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisEALs.GenericThe Loss threshold RPV water level of 161 in. corresponds to the level that is used in EOPs toindicate challenge of core cooling.This threshold is the same as Fuel Clad Barrier Potential Loss threshold A.1 and corresponds to achallenge to core cooling. Thus, this threshold indicates a Loss of RCS barrier and Potential Lossof Fuel Clad barrier that appropriately escalates the emergency classification level to a Site AreaEmergency.Unlike the Fuel Clad barrier RPV water level Potential Loss threshold (top of the active fuel), theadditional requirement that the RPV be depressurized is not associated with the RCS barrierPotential Loss. The significant loss of inventory that must occur to determine that RPV water levelcannot be restored and maintained above the threshold is, by itself, a very strong indication thatthe RCS barrier is no longer capable of retaining sufficient inventory to keep the core submerged,and thus represents a Loss of the RCS Barrier.There is no Potential Loss threshold associated with this item.NMP2 Basis Reference(s):1. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document2. N2-EOP-C4 RPV Flooding3. N2-EOP-C5 Failure to Scram4. NEI 99-01 RCS Loss 2Page 225EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: LossThreshold:2. Primary Containment pressure > 1.68 psig due to RCS leakageBasis:Plant-SpecificThe drywell high pressure scram setpoint is an entry condition to the EOP flowcharts:EOP-RPV, RPV Control, and EOP-PC, Primary Containment Control (ref. 1, 2). NormalPrimary Containment (PC) pressure control functions such as operation of drywell coolingand venting through GTS are specified in EOP-PC in advance of less desirable but moreeffective functions such as operation of drywell or suppression chamber sprays.In the NMP2 design basis, Primary Containment pressures above the drywell highpressure scram setpoint are assumed to be the result of a high-energy release into thecontainment for which normal pressure control systems are inadequate or incapable ofreversing the increasing pressure trend. Pressures of this magnitude, however, can becaused by non-LOCA events such as a loss of drywell cooling or inability to controlPrimary Containment vent/purge (ref. 3, 4).The threshold phrase "...due to RCS leakage" focuses the barrier failure on the RCSinstead of the non-LOCA malfunctions that may adversely affect Primary Containmentpressure. Primary Containment pressure greater than 1.68 psig with corollary indications(e.g., elevated drywell temperature, indications of loss of RCS inventory) should, therefore,be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressuregreater than 1.68 psig should not be considered an RCS barrier loss.GenericThe Primary Containment pressure of 1.68 psig is based on the drywell high pressure set pointwhich indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.There is no Potential Loss threshold associated with this item.Page 226 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP2 Basis Reference(s):1. N2-EOP RPV RPV Control2. N2-EOP-PC Primary Containment Control3. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document4. USAR Section 6.25. NEI 99-01 RCS Loss 1Page 227EPMP-EPP-01 02Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: C. IsolationDegradation Threat: LossThreshold:3. Release pathway exists outside Primary Containment resulting from isolation failure inANY of the following systems (excluding normal process system flowpaths from anUNISOLABLE system):" Main steam line" RCIC steam line* RWCU" FeedwaterBasis:Plant-SpecificThe conditions of this threshold include required containment isolation failures allowing aflow path to the environment. A release pathway outside Primary Containment exists whenflow is not prevented by downstream isolations. Emergency declaration under thisthreshold would not be required in the case of a failure of both isolation valves to close butno downstream flowpath exists. Similarly, if the emergency response requires the normalprocess flow of a system outside Primary Containment (e.g., EOP requirement to bypassMSIV low RPV water level interlocks and maintain the main condenser as a heat sinkusing main turbine bypass valves), the threshold is not met. The combination of thesethreshold conditions represent the loss of both the RCS and Containment (see PC LossC.3) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or PotentialLoss of any two barriers). (ref. 1-4)Even though RWCU and Feedwater systems do not contain steam, they are included inthe list because an UNISOLABLE break could result in the high-pressure discharge of fluidthat is flashed to steam from relatively large volume systems directly connected to theRCS.GenericPage 228 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisAn UNISOLABLE MSL break is a breach of the RCS barrier. Thus, this threshold is included forconsistency with the Alert emergency classification level.Other large high-energy line breaks such as Feedwater, RWCU, or RCIC that are UNISOLABLEalso represent a significant loss of the RCS barrier and should be considered as MSL breaks forpurposes of classification.NMP2 Basis Reference(s):1. USAR Section 5.4.52. USAR Section 5.4.63. USAR Section 5.4.84. USAR Section 5.4.95. NEI 99-01 RCS Loss 3APage 229EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: C. IsolationDegradation Threat: LossThreshold:4. RPV blowdown is requiredBasis:Plant-SpecificRPV blowdown (Emergency RPV Depressurization) is specified in the EOP flowchartswhen symbols containing the phrase "BLOW DOWN" are reached. The requirements foremergency RPV depressurization appear in the following EOPs (ref. 1-7):* EOP-RPV RPV Control" EOP-PC Primary Containment Control" EOP-SC Secondary Containment Control" EOP-RR Radioactivity Release Control" EOP-PCH Hydrogen Control" EOP-C3 Steam Cooling" EOP-C5 Failure to ScramRPV blowdown (Emergency RPV Depressurization) is also performed upon entry to EOP-C4 (ref. 8).GenericPlant symptoms requiring Emergency RPV Depressurization (RPV blowdown) per the EOPflowcharts are indicative of a loss of the RCS barrier. If Emergency RPV depressurization isrequired, the plant operators are directed to open safety relief valves (SRVs) and keep them open.Even though the RCS is being vented into the suppression pool, a loss of the RCS should beconsidered to exist due to the diminished effectiveness of the RCS pressure barrier to a release offission products beyond its boundary.Page 230 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP2 Basis Reference(s):1. N2-EOP-RPV RPV Control2. N2-EOP-PC primary Containment Control3. N2-EOP-SC Secondary Containment Control4. N2-EOP-RR Radioactivity Release Control5. N2-EOP-PCH Hydrogen Control6. N2-EOP-C3 Steam Cooling7. N2-EOP-C5 Failure to Scram8. N2-EOP-C4 RPV Flooding9. NEI 99-01 RCS Loss 3Page 231EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: D. RadDegradation Threat: LossThreshold:5. Drywell area radiation 41 R/hr (4.1 E4 mRem/hr)Basis:Plant-SpecificIt is important to recognize that the radiation monitor may be sensitive to shine from theRPV or RCS piping (caused by lower than normal RPV water level for example). TheDrywell High Range Radiation Monitors are the following (ref. 1):" 2CEC*PNL88OD: DRMS 2RMS*RE1 B/DRMS*RUZIARMS*RUZ1 B* 2CEC*PNL88OB: DRMS 2RMS*RE1A/CRMS*RUZ1 CRMS*RUZ1 DFigure F-1 illustrates the location of the following four detectors inside the drywell (ref. 1):" 2RMS*RE1A P.C. 268 170EAZ" 2RMS*RE1C P.C. 267 024EAZ* 2RMS*RE1B P.C. 268 245EAZ* 2RMS*RE1D P.C. 268 353EAZThe threshold value was calculated assuming the instantaneous release and dispersal ofthe reactor coolant noble gas and iodine inventory associated with normal operatingconcentrations (i.e., within Technical Specifications) into the drywell atmosphere (ref. 2).The reading is less than that specified for the Fuel Clad Loss because no damage to thePage 232 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basisfuel clad is assumed in this RCS Loss. Only leakage from the RCS is assumed in thisEAL.GenericThe 41 RPhr reading is a value which indicates the release of reactor coolant to the PrimaryContainment.This reading will be less than that specified for Fuel Clad barrier Loss threshold D.2. Thus, thisthreshold would be indicative of a RCS leak only. If the radiation monitor reading increased to thatvalue specified by Fuel Clad Barrier threshold, fuel damage would also be indicated.There is no Potential Loss threshold associated with this item.NMP2 Basis Reference(s):1. N2-RSP-RMS-R106 Channel Calibration Test of the Drywell High Range AreaRadiation Monitors2. Calculation PR-C-24-03. NEI 99-01 RCS Loss 4Page 233EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisFigure F-I: Drywell High Range Radiation Monitor Detector Locations (ref. 1)Drywell 261Drywell 26177r/N-.;" .,A7 A24Dec:reeKmN 6'/-\Peuo r-e! &.- Equipmenthatzh..Li'- )i,- -"//I,/ \____ )N'_ N'~0 ________Q *)7I, )//'N /N /'N,-( '~N 'NN', ~"N.__/'NA-;;M2 -REInL:! -; A n Le --SVCAescap~ehatchPage 234EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: E. JudgmentDegradation Threat: LossThreshold:6. ANY condition in the opinion of the Emergency Director that indicates loss of the RCSbarrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the RCS barrier is lost. Such a determination should include IMMINENTbarrier degradation, barrier monitoring capability and dominant accident sequences." IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to the recognition of the inability to reach safety acceptancecriteria before completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the RCS barrier is lost. In addition, the inability to monitor the barrier shouldalso be incorporated in this threshold as a factor in Emergency Director judgment that the barriermay be considered lost.Page 235 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP2 Basis Reference(s):1. NEI 99-01 RCS Loss 6Page 236EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: A. RPV LevelDegradation Threat: Potential LossThreshold:NonePage 237EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:Category:Degradation Threat:Threshold:Reactor Coolant SystemB. Primary Containment Pressure / TemperaturePotential LossNonePage 238EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: C. IsolationDegradation Threat: Potential LossThreshold:1. UNISOLABLE primary system leakage outside Primary Containment as indicated byexceeding EITHER:RB area temperature above an isolation setpointORRB area radiation above an alarm setpointBasis:Plant-SpecificThe presence of elevated general area temperatures or radiation levels in the ReactorBuilding (RB) may be indicative of UNISOLABLE primary system leakage outside thePrimary Containment. When parameters reach the threshold level, equipment failure ormisoperation may be occurring. Elevated parameters may also adversely affect the abilityto gain access to or operate equipment within the affected area. (ref. 1, 2)In general, multiple indications should be used to determine if a primary system isdischarging outside Primary Containment. For example, a high area radiation conditiondoes not necessarily indicate that a primary system is discharging into the secondarycontainment since this may be caused by radiation shine from nearby steam lines or themovement of radioactive materials. Conversely, a high area radiation condition inconjunction with other indications (e.g. room flooding, high area temperatures, reports ofsteam in the secondary containment, an unexpected rise in feedwater flowrate, orunexpected main turbine control valve closure) may indicate that a primary system isdischarging into the secondary containment.GenericEOP-SC temperature isolation setpoints or area radiation alarm setpoints in the areas of the mainsteam line tunnel, main turbine generator, RCIC, etc., indicate a direct path from the RCS to areasoutside Primary Containment.Page 239 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisThe indicators reaching the threshold barriers and confirmed to be caused by RCS leakagewarrant an Alert classification. An UNISOLABLE leak which is indicated by a high alarm setpointescalates to a Site Area Emergency when combined with Containment Barrier Loss threshold C.5(after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is alsoexceeded.NMP2 Basis Reference(s):1. N2-EOP-SC Secondary Containment Control2. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document3. NEI 99-01 RCS Potential Loss 3BPage 240EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: D. RadDegradation Threat: Potential LossThreshold:NonePage 241EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: Reactor Coolant SystemCategory: E. JudgmentDegradation Threat: Potential LossThreshold:2. ANY condition in the opinion of the Emergency Director that indicates potential loss ofthe RCS barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the RCS barrier is potentially lost. Such a determination should includeIMMINENT barrier degradation, barrier monitoring capability and dominant accidentsequences." IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to the inability to reach final safety acceptance criteria beforecompleting all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.GenericThis, threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the RCS barrier is potentially lost. In addition, the inability to monitor thebarrier should also be incorporated in this threshold as a'factor in Emergency Director judgmentthat the barrier may be considered potentially lost.Page 242 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisNMP2 Basis Reference(s):1. NEI 99-01 RCS Potential Loss 6Page 243 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:Category:Degradation Threat:Threshold:ContainmentA. RPV LevelLossNonePage 244EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:ContainmentCategory:B. Primary Containment Pressure / TemperatureDegradation Threat: LossThreshold:1. Primary Containment pressure rise followed by a rapid UNPLANNED drop in PrimaryContainment pressureBasis:Plant-SpecificNoneGenericRapid UNPLANNED loss of pressure (i.e., not attributable to drywell spray or condensation effects)following an initial pressure increase from a high energy line break indicates a loss of containmentintegrity. Primary Containment pressure should increase as a result of mass and energy releaseinto containment from a LOCA. Thus, Primary Containment pressure not increasing under theseconditions indicates a loss of containment integrity.This indicator relies on operator recognition of an unexpected response for the condition andtherefore does not have a specific value associated with it. The unexpected response is importantbecause it is the indicator for a containment bypass condition.NMP2 Basis Reference(s):1. NEI 99-01 CMT Loss 1APage 245 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: LossThreshold:2. Primary Containment pressure response not consistent with LOCA conditionsBasis:Plant-SpecificUSAR Section 6.2.1 provides a summary of Primary Containment pressure response forseveral postulated accident conditions resulting in the release of RCS inventory to thecontainment. These accidents include:" Rupture of a recirculation line* Rupture of a main steam line* Intermediate size liquid line rupture* Small size steam line ruptureThe containment response to the main steam line, intermediate liquid line and small sizesteam line breaks were bounded by the recirculation line break. (ref. 1)USAR Figures 6.2-2 and 6.2-3 illustrate the containment pressure response due to arecirculation line break (ref. 2, 3). The maximum calculated drywell pressure is 39.75 psigand is well below the design allowable pressure of 45 psig. (ref. 4, 5)Due to conservatisms in LOCA analyses, actual pressure response is expected to be lessthan the analyzed response. For example, blowdown mass flowrate may be only 60-80%of the analyzed rate, initial containment pressure may be less than 0.75 psig, etc.LOCA conditions are manifested on Control Room instrumentation by drywell pressurerising with suppression chamber pressure following in a manner similar to that shown inUSAR Figures 6.2-2 and 6.2-3. A broken SRV tailpipe could infer this threshold ifsuppression chamber pressure is higher than drywell pressure; however, if the SRV isPage 246 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basisclosed, the condition would no longer exist.GenericRapid UNPLANNED loss of pressure (i.e., not attributable to drywell spray or condensation effects)following an initial pressure increase from a high energy line break indicates a loss of containmentintegrity. Primary Containment pressure should increase as a result of mass and energy releaseinto containment from a LOCA. Thus, Primary Containment pressure not increasing under theseconditions indicates a loss of containment integrity.This indicator relies on operator recognition of an unexpected response for the condition andtherefore does not have a specific value associated with it. The unexpected response is importantbecause it is the indicator for a containment bypass condition.NMP2 Basis Reference(s):1. USAR Section 6.2.12. USAR Figure 6.2-23. USAR Figure 6.2-34. USAR Table 6.2-185. USAR Section 6.2.1.1.26. NEI 99-01 CMT Loss 1BPage 247EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: C. IsolationDegradation Threat: LossThreshold:3. Failure of all Primary Containment isolation valves in ANY one line to close followingauto or manual initiationANDDirect downstream pathway outside Primary Containment and to the environmentexistsBasis:Plant-SpecificThis threshold addresses failure of open isolation devices which should close upon receiptof a manual or automatic Primary Containment isolation signal resulting in a significantradiological release pathway directly to the environment. The concern is the UNISOLABLEopen pathway to the environment. A failure of the ability to isolate any one line indicates abreach of Primary Containment integrity. Technical Specifications Table 3.6.1.3-1 providesa list of applicable isolation valves (ref. 1).As stated above, the adjective "Direct" modifies "pathway" to discriminate against releasepaths through interfacing liquid systems. Leakage into a closed system is to be consideredonly if the closed, system is breached and thereby creates a significant pathway to theenvironment. Examples include UNISOLABLE Main steam line or RCIC steam line breaks,UNISOLABLE RWCU system breaks, and unisloable Primary Containment atmospherevent paths. If the main condenser is available with an UNISOLABLE main steam line,there may be releases through the steam jet air ejectors and gland seal exhausters. Thesepathways are monitored, however, and do not meet the intent of a nonisolable releasepath to the environment. These minor releases are assessed using the Category R EALs.The existence of an in-line charcoal filter (GTS) does not make a release path indirectsince the filter is not effective at removing fission noble gases. Typical filters have anefficiency of 95-99% removal of iodine. Given the magnitude of the core inventory ofiodine, significant releases could still occur. In addition, since the fission product releasePage 248 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basiswould be driven by boiling in the reactor vessel, the high humidity in the release streamcan be expected to render the filters ineffective in a short period.The threshold is met if the breach is not isolable from the Control Room or an attempt forisolation from the Control Room has been made and was unsuccessful. An attempt forisolation from the Control Room should be made prior to the emergency classification. Ifoperator actions from the Control Room are successful, this threshold is not applicable.Credit is not given for operator actions taken in-plant (outside the Control Room) to isolatethe breach.N2-EOP-PC, Primary Containment Control may specify Primary Containment venting andintentional bypassing of the containment isolation valve logic even if offsite radioactivityrelease rate limits are exceeded (ref. 2). Under these conditions with a VALID containmentisolation signal, the Containment barrier should be considered lost.GenericThese thresholds address incomplete containment isolation that allows direct release to theenvironment.The use of the modifier "direct" in defining the release path discriminates against release pathsthrough interfacing liquid systems. The existence of an in-line charcoal filter does not make arelease path indirect since the filter is not effective at removing fission product noble gases. Typicalfilters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory ofiodine, significant releases could still occur. In addition, since the fission product release would bedriven by boiling in the reactor vessel, the high humidity in the release stream can be expected torender the filters ineffective in a short period.NMP2 Basis Reference(s):1. Improved Technical Specifications Nine Mile Point Nuclear Station, Unit No. 2, Table3.6.1.3-12. N2-EOP-PC Primary Containment Control3. NEI 99-01 CMT Loss 3APage 249 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: C. IsolationDegradation Threat: LossThreshold:4. Intentional Primary Containment venting per EOPsBasis:Plant-SpecificN2-EOP-PC, Primary Containment Control, and N2-EOP-PCH, Hydrogen Control, mayspecify Primary Containment venting and intentional bypassing of the containmentisolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1, 2).The threshold is met when the operator begins venting the Primary Containment inaccordance with EOP-6, Support Procedures (Attachment 21 or 25), not when actions aretaken to bypass interlocks prior to opening the vent valves (ref. 3). Purge and vent actionsspecified in N2-EOP-PC Step PCP-1 to control Primary Containment pressure below thedrywell high pressure scram setpoint or EOP-PCH Step 31 or 34 to lower hydrogenconcentration does not meet this threshold because such action is only permitted if offsiteradioactivity release rates will remain below the ODCM limits (ref. 1, 2).GenericThese thresholds address incomplete containment isolation that allows direct release to theenvironment.Site specific EOPs may direct containment isolation valve logic(s) to be intentionally bypassed,regardless of radioactivity release rates. Under these conditions with a VALID containmentisolation signal, the containment should also be considered lost if containment venting is actuallyperformed.Intentional venting of Primary Containment for Primary Containment pressure or combustible gascontrol per EOPs to the secondary containment and/or the environment is considered a loss ofcontainment. Containment venting for pressure when not in an accident situation should not beconsidered.NMP2 Basis Reference(s):1. N2-EOP-PC Primary Containment Control2. N2-EOP-PCH Hydrogen ControlPage 250 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basis3. EOP-6 NMP2 EOP Support Procedure4. NEI 99-01 CMT Loss 3BPage 251EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: C. IsolationDegradation Threat: LossThreshold:5. UNISOLABLE primary system leakage outside Primary Containment as indicated byexceeding EITHER:RB area maximum safe temperature value (N2-EOP-SC Detail S)ORRB area radiation > 8.OOE+3 mR/hrBasis:Plant-SpecificThe presence of elevated general area temperatures or radiation levels in the ReactorBuilding (RB) may be indicative of UNISOLABLE primary system leakage outside thePrimary Containment. The EOP maximum safe values define this Containment barrierthreshold because they are indicative of problems in the secondary containment that arespreading and pose a threat to achieving a safe plant shutdown. This threshold addressesproblematic discharges outside Primary Containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concerncorrespond to the areas addressed in N2-EOP-SC Detail S (ref. 1). See Figure F-2.A "Maximum Safe Value" is the highest value at which equipment necessary for the safeshutdown of the plant will operate and personnel can perform any actions necessary forthe safe shutdown of the plant.The maximum safe value for temperature is dependent on whether access is needed toareas within the reactor building to perform actions required by other EOP steps. Onlyareas in which the actions must be taken (and there is no other alternative) qualify as"areas" when determining the number of affected areas. (ref. 2)The maximum safe value for radiation is 8.OOE+3 mR/hr.In general, multiple indications should be used to determine if a primary system isdischarging outside Primary Containment. For example, a high area radiation conditionPage 252 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basisdoes not necessarily indicate that a primary system is discharging into the secondarycontainment since this may be caused by radiation shine from nearby steam lines or themovement of radioactive materials. Conversely, a high area radiation condition inconjunction with other indications (e.g. room flooding, high area temperatures, reports ofsteam in the secondary containment, an unexpected rise in feedwater flowrate, orunexpected main turbine control valve closure) may indicate that a primary system isdischarging into the secondary containment.GenericThis threshold addresses incomplete containment isolation that allows direct release to theenvironment.In addition, The presence of area radiation or temperature Maximum Safe Values indicatingUNISOLABLE primary system leakage outside the Primary Containment are addressed after acontainment isolation. The indicators should be confirmed to be caused by RCS leakage.There is no Potential Loss threshold associated with this item.NMP2 Reference(s):1. N2-EOP-SC Secondary Containment Control2. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document3. NEI 99-01 CMT Loss 3CPage 253 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisFigure F-2: N2-EOP-SC Detail SSI Maximum Safe ValuesParameter Location Maximum Safe ValueArea Temperature All areas 212°F(EOP-6 Att 28)Areas when access 135°Fis required for support ofEOP actions.Area Radiation All areas .8.OOE+3 mR/hrArea Water Level All areas- Flooding alarmPage 254EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: D. RadDegradation Threat: LossThreshold:NonePage 255EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: E. JudgmentDegradation Threat: LossThreshold:6. ANY condition in the opinion of the Emergency Director that indicates loss of theContainment barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the Containment barrier is lost. Such a determination should includeIMMINENT barrier degradation, barrier monitoring capability and dominant accidentsequences." IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to recognition of the inability to reach safety acceptance criteriabefore completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.Page 256 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisGenericThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the Containment barrier is lost. In addition, the inability to monitor the barriershould also be incorporated in this threshold as a factor in Emergency Director judgment that thebarrier may be considered lost.The Containment barrier should not be declared lost based on exceeding Technical Specificationaction statement criteria, unless there is an event in progress requiring mitigation by theContainment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Cladand/or RCS) the Containment barrier status is addressed by Technical Specifications.NMP2 Basis Reference(s):1. NEI 99-01 CMT Loss 6Page 257 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: A. RPV LevelDegradation Threat: Potential LossThreshold:1. Primary Containment Flooding is requiredBasis:Plant-SpecificRequirements for Primary Containment Flooding are established in EOP-RPV Step L-1 6;EOP-C5 Steps L-8, L-10 and L-18; and EOP-C4 Override 1. These EOPs provideinstructions to ensure adequate core cooling by maintaining RPV water level aboveprescribed limits or operating sufficient RPV injection sources when level cannot bedetermined. SAP entry is required when (ref. 1):RPV water level cannot be restored and maintained above -39 in. with insufficientCore Spray Cooling: The Minimum Steam Cooling RPV Water Level (MSCRWL) isthe lowest RPV water level at which the covered portion of the reactor core willgenerate sufficient steam to preclude any clad temperature in the uncovered portionof the core from exceeding 1500'F. Core Spray Cooling is insufficient if RPV waterlevel cannot be restored and maintained at or above -62 in. with at least 6350 gpmcore spray loop flow. Consistent with the EOP definition of "cannot be restored andmaintained," the determination that the parameter cannot be restored andmaintained above the limit may be made at, before, or after the parameter actuallydecreases to this point.RPV water level cannot be determined and it is determined that core damage isoccurring: When RPV water level cannot be determined, EOPs require RPVflooding strategies. RPV water level indication provides the primary means ofknowing if adequate core cooling is being maintained. When all means ofdetermining RPV water level are unavailable, reliance on alternate means ofassuring adequate core cooling must be attempted. The instructions in EOP-C4specify these means, which include emergency depressurization of the RPV andPage 258 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basisinjection into the RPV at a rate needed to flood to the elevation of the main steamlines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWSevents)This threshold is also a Loss of the Fuel Clad barrier (FC Loss A.1). Since PrimaryContainment Flooding occurs after core uncovery has occurred a Loss of the RCS barrierexists (RCS Loss A.1). Primary Containment Flooding (SAP entry), therefore, represents aLoss of two barriers and a Potential Loss of a third, which requires a General Emergencyclassification.GenericThere is no Loss threshold associated with this item.The potential loss requirement for drywell flooding indicates adequate core cooling cannot beestablished and maintained and that core melt is possible. Entry into Primary ContainmentFlooding procedures (SAPs) is a logical escalation in response to the inability to maintain adequatecore cooling.The condition in this potential loss threshold represents a potential core melt sequence which, ifnot corrected, could lead to vessel failure and increased potential for containment failure. Inconjunction with Reactor Vessel water level "Loss" thresholds in the Fuel Clad and RCS barriercolumns, this threshold will result in the declaration of a General Emergency -- loss of two barriersand the potential loss of a third.NMP2 Basis Reference(s):1. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document2. N2-EOP-C4 RPV Flooding3. NEI 99-01 CMT Potential Loss 2Page 259EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: Potential LossThreshold:2. Primary Containment pressure > 45 psig and risingBasis:Plant-SpecificIf this threshold is exceeded, a challenge to the Primary Containment structure hasoccurred because assumptions used in the accident analysis are no longer VALID and anunanalyzed condition exists (ref. 1). This constitutes a Potential Loss of the Containmentbarrier even if a containment breach has not occurred.GenericThe Primary Containment pressure of 45 psig is based on the Primary Containment designpressure.NMP2 Basis Reference(s):1. USAR Section 6.2.1.1.22. NEI 99-01 CMT Potential Loss 1APage 260EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: B. Primary Containment Pressure / TemperatureDegradation Threat: Potential LossThreshold:3. Explosive mixture exists inside Primary Containment (> 6% H2 and 2 5% 02)Basis:Plant-SpecificExplosive (deflagration) mixtures in the Primary Containment are assumed to be elevatedconcentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generationfor development of EOPs/SAPs indicates that any hydrogen concentration above minimumdetectable is not to be expected within the short term. Post-LOCA hydrogen generationprimarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogenconcentrations that rapidly develop are most likely caused by metal-water reaction. Ametal-water reaction is indicative of an accident more severe than accidents considered inthe plant design basis and would be indicative, therefore, of a potential threat to PrimaryContainment integrity. Hydrogen concentration of approximately 6% is considered theglobal deflagration concentration limit (ref. 1).Except for brief periods during plant startup and shutdown, oxygen concentration in thePrimary Containment is maintained at insignificant levels by nitrogen inertion. Thespecified values for this Potential Loss threshold are the minimum global deflagrationconcentration limits (6% hydrogen and 5% oxygen, ref. 1) and readily recognizablebecause 6% hydrogen is well above the N2-EOP-PCH entry condition (ref. 2). Theminimum global deflagration hydrogen/oxygen concentrations (6%/5%, respectively)require intentional Primary Containment venting, which is defined to be a Loss ofContainment (PC Loss C.4).The USAR requires the H2/02 analyzers to be able to provide and record combustible gasconcentration in the Primary Containment within 90 minutes following a LOCA with safetysystem injection. The H2/02 analyzers are normally in standby and require a 30 minutePage 261 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basiswarrn-up/self-test period before they start providing data. (ref. 1)If the hydrogen or oxygen monitor is unavailable, sampling and analysis may determinegas concentrations. The validity of sample results must be judged based upon plantconditions, since drawing and analyzing samples may take some time. If sample resultscannot be relied upon and hydrogen concentrations cannot be determined by any othermeans, the concentrations must be considered "unknown." The monitors should not beconsidered "unavailable" until an attempt has been made to place them in service. (ref. 1)GenericBWRs specifically define the limits associated with explosive mixtures in terms of deflagrationconcentrations of hydrogen and oxygen.NMP2 Basis Reference(s):1. NER-2M-039, NMP2 Emergency Operating Procedures (EOP) Basis Document2. N2-EOP-PCH Hydrogen Control3. NEI 99-01 CMT Potential Loss 1BPage 262EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier:Category:ContainmentB. Primary Containment Pressure / TemperatureDegradation Threat: Potential LossThreshold:4. Suppression pool temperature and RPV pressure cannot be maintained below theHeat Capacity Temperature Limit (N2-EOP-PC Figure M)Basis:Plant-SpecificThe Heat Capacity Temperature Limit (HCTL) is given in EOP Figure M. This threshold ismet when N2-EOP-PC Step SPT-6 is reached (ref. 1).GenericThe Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature fromwhich Emergency RPV Depressurization will not raise:" Suppression chamber temperature above the design value (270°F),OR" Suppression chamber pressure above Primary Containment Pressure Limit, before the rateof energy transfer from the RPV to the containment is greater than the capacity of thecontainment vent.The HCTL is a function of RPV pressure and suppression pool water level. It is utilized to precludefailure of the containment and equipment in the containment necessary for the safe shutdown ofthe plant and therefore, the inability to maintain plant parameters below the limit constitutes apotential loss of Containment.NMP2 Basis Reference(s):1. N2-EOP-PC Primary Containment Control2. NEI 99-01 CMT Potential Loss 1CPage 263EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: D. RadDegradation Threat: Potential LossThreshold:5. Drywell area radiation __ 6.0 E4 R/hr (6.0 E7 mRem/hr)Basis:Plant-SpecificIt is important to recognize that the radiation monitor may be sensitive to shine from theRPV or RCS piping (caused by lower than normal RPV water level for example). TheDrywell High Range Radiation Monitors are the following (ref. 1):* 2CEC*PNL88OD: DRMS 2RMS*RE1B/DRMS*RUZ1ARMS*RUZ1 B" 2CEC*PNL88OB: DRMS 2RMS*RE1A/CRMS*RUZ1CRMS*RUZ1 DFigure F-1 illustrates the location of the following four detectors inside the drywell (ref. 1):" 2RMS*RE1A P.C. 268 170EAZ* 2RMS*RE1C P.C. 267 024EAZ" 2RMS*RE1B P.C. 268 245EAZ" 2RMS*RE1D P.C. 268 353EAZThe threshold value was calculated assuming the instantaneous release and dispersal ofthe reactor coolant noble gas and iodine inventory associated with 20% fuel clad damageinto the drywell atmosphere (ref. 2, 3). The referenced calculation yields a value of 5.6 E4R/hr. This has been rounded to 6.0 E4 R/hr because it is observable on existingPage 264 EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and Basisinstrumentation.GenericThe 6.0 E4 R/hr reading is a value that indicates significant fuel damage well in excess of thatrequired for loss of RCS and Fuel Clad.Regardless of whether containment is challenged, this amount of activity in containment, ifreleased, could have such severe consequences that it is prudent to treat this as a potential loss ofcontainment, such that a General Emergency declaration is warranted.There is no Loss threshold associated with this item.NMP2 Basis Reference(s):1. N2-RSP-RMS-R106 Channel Calibration Test of the Drywell High Range AreaRadiation Monitors2. Calculation PR-C-24-03. CCN No. 009718 Calculation of Drywell Radiation General Emergency EAL4. NEI 99-01 CMT Potential Loss 4Page 265EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisFigure F-I: Drywell High Range Radiation Monitor Detector Locations (ref. 1)Drywell 261Drywell 261\N .7i'iA5Z 24 DereesI7 \\A .170 D,\ / ,"./"\7_...),Per-ronne! &Equ ipmenth q tzhj())352 D*{Er10-NEr =qesap/R L*N~ýJ['-Nh. a ItcPage 266EPMP-EPP-0102Rev 00 (Draft A) -Fission Product Barrier Loss / Potential Loss Matrix and BasisBarrier: ContainmentCategory: E. JudgmentDegradation Threat: Potential LossThreshold:6. ANY condition in the opinion of the Emergency Director that indicates potential loss ofthe Containment barrierBasis:Plant-SpecificThe Emergency Director judgment threshold addresses any other factors relevant todetermining if the Containment barrier is potentially lost. Such a determination shouldinclude IMMINENT barrier degradation, barrier monitoring capability and dominantaccident sequences." IMMINENT barrier degradation exists if the degradation will likely occur within twohours based on a projection of current safety system performance. The term"IMMINENT" refers to recognition of the inability to reach safety acceptance criteriabefore completion of all checks.* Barrier monitoring capability is decreased if there is a loss or lack of reliableindicators. This assessment should include instrumentation operability concerns,readings from portable instrumentation and consideration of offsite monitoringresults.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Director should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergencyclassification declarations.Page 267 EPMP-EPP-0102Rev 00 (Draft A) Fission Product Barrier Loss / Potential Loss Matrix and BasisGenericThis threshold addresses any other factors that are to be used by the Emergency Director indetermining whether the Containment barrier is potentially lost. in addition, the inability to monitorthe barrier should also be incorporated in this threshold as a factor in Emergency Directorjudgment that the barrier may be considered potentially lost.The Containment barrier should not be declared potentially lost based on exceeding TechnicalSpecification action statement criteria, unless there is an event in progress requiring mitigation bythe Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Cladand/or RCS) the Containment barrier status is addressed by Technical Specifications.NMP2 Basis Reference(s):1. NEI 99-01 CMT Potential Loss 6Page 268EPMP-EPP-01 02Rev 00 (Draft A)